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| l THE CONTINUING PROBLEM OF RADIOACTIVE METAL SCRAP James G. Yusko, CHP and Joel O. Lubenau, CHP 1
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| PA Department of Environment. Resources !
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| and US Nuclear Regulatory Commission !
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| l I
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| Presented at the 27th Annual Conference on Radiation Control San Antonio, TX May 9, 1995 i i This paper will be published as part of the Proceedings of the 27th ananal Conference on Radiation Control available from the Conference of Radiation Control Program Directors, Inc. 205 Capital Avenue, Frankfort, KY 40601.
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| 9610300231 960830 t
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| THE CONTINUING PROBLEM OF RADIOACTIVE METAL SCRAP James G. Yusko, CHP and Joel O. Lubenau , CHP 1
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| PA Department of Environmental Resources and 2
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| US Nuclear Regulatory Commission l
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| ABSTRACT Metal scrap found to contain radioactive materials continues ;
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| to challenge regulatory agencies as discoveries of this unwanted constituent increase. And while efforts are made to prevent the exposure of personnel at metal manufacturing mills and scrap yard. 1 when radioactivity is discovered in a shipment of metal scrap, this '
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| has not stemmed the number of discoveries. Sources and devices continue to be found, leading to difficulties in the disposal of the radioactive materials, especially with the closure of licensed '
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| LLRW facilities to non-compact state members. Naturally-occurring !
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| radioactive materials continue to be found, principally as surface contaminants of metals for recycling. And although NORM contamination does not generally pose a threat to the health and safety of personnel at metal mills and scrap yards, there is no consensus about the disposition of NORM-contaminated metal. The changing of trade barriers (such as the North American Free Trade Agreement) also factors into the problem, as materials cross < l international boundaries and enter the recycling stream. The efforts of entities such as Conference committees, federal regulatory agencies (e.g., NRC, EPA, DOT), state radiation control agencies and the affected industries will be presented and discussed.
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| INTRODUCTION A scrap dealer in Illinois, after borrowing the survey meters and expertise of a nearby national laboratory which occasionally sold metal scrap to him, decided to purchase a radiation survey meter. He initially found a radiation level above what he was led to believe was a normal ambient background. He verified that the batteries were installed properly in the instrument, and proceeded to conduct a more complete survey of his property. He found radiation levels up to 1 millirem per hour. He called for help.
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| Eventually, a source of cesium-137 containing 14 CBq (370 mci) was found to have been buried on the property. No one knows where the source came from, and it was on his property for at least three years before it was discovered in December, 1994. Obviously, there was the potential for radiation exposure of scrap yard workers.
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| 1 In January,1994, a scrap dealer in Estonia detected radiation in a load of scrap delivered by a truck. A cesium-137 source was found and was transferred to a hazardous waste disposal facility, but it was then stolen in October, 1994, when three brothers removed the metal box containing the source. When the source fell out of the box, one of the men picked up the source and put it in his pocket. The source was found by Estonian authorities after the man and the family dog died, and a stepson was hospitalized because of radiation injuries. Radiation levels in the man's kitchen were above one gray (100 rad) per hour. Later, a second source was found along a roadside by authorities who happened to have their i detector turned on as they drove by.
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| i These recent episodes illustrate dramatically the continuing I problem of radioactive materials in metal scrap. It is also a ,
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| problem whose true scope is yet to be measured. In the April, ~ j 1995, issue of Health Physics, we reported on 315 events involving radioactive materials in metal scrap for the period ending in 1993.
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| We also reported that we were workir3 with the Steel Manufacturers i Association (SMA) to add similar data collected by its members.
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| This effort has been successful and we can now report on the ,
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| combined data updated through December, 1994. Canadian data is i also included.
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| RECENT EXPERIENCE Through the end of December, 1994, the authors are aware of 1101 cases where radioactivity has been discovered in shipments of i metal scrap. Of this total, there have been 24 smeltings in the United States, principally af fecting the steel industry [ figure 1] .
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| Both sets of data have a significant, common feature: there has l been a clear, upward trend in the discoveries of radioactive material in metal scrap.
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| As can be seen from the accompanying slide [ figure 2], what are generally referred to as portal monitors have by far detected the greatest number of abnormal radiation finds, although the use of hand-held, portable survey meters also has found about 2% of the total, and even visual sightings have resulted in sources being
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| ?
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| found. One operator became suspicious when he saw an object stencilled " property of the U. S. government" on it and tritium gauges were found. As you can see, only a few of the discoveries have been by sight, although an interesting, more recent sighting i will be discussed later. The portal monitors come in many configurations, some of which have detectors to cover four sides of a vehicle (two sides, top and even bottom) ; some of which are coupled to computer microprocessors, which update background radiation levels continuously and can be set to alarm if variations as little as a few percent above background are detected. Others may be either " simple" sodium iodide or plastic scintillators coupled to alarming rate meters.
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| This chart [ figure 3] shows the distribution of radioactive materials found. About 10% of the finds have involved Atomic Energy Act materials, but the vast majority, about 70%, have involved either radium (as discrete finds or sources) or naturally-occurring radioactive materials --
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| NORM. However, sources or devices have accounted for about one eighth of the finds. As can I be seen, when sources or devices are found, Atomic Energy Act 1 materials comprise the majority of finds. j As noted earlier, the number of discoveries has generally been 1 increasing every year (figure 4]. One possible explanation for l this is that there are more sources finding their way to the scrap !
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| yard, but the data we have on discovered sources shows a less '
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| gradual increase (figure 5] . Another explanation is the lii l installation and use of portal monitors by the metal manufacturing mills and their (feeder) scrap yards. This reminds one author of i
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| , something he learned in an epidemiology class a few decades ago: j the monitors may be seeing prevalence, not incidence. It may be ';
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| that the problem -- especially of NORM-contaminated items -- has j always been there, but we're only recently seeing how widespread l this is, especially as the instrumentation grows more l sophisticated. {!
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| l World-wide, we are aware of 36 smeltings of radioactive materials in metals (figure 6]. Most of these (25) have occurred in steel mills and foundries, although there have also been five smeltings in aluminum mills, two each in copper and gold, and one each in lead and zine mills. Although most of the events have !
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| i occurred in the United States, this aspect may be due to better I
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| : domestic information flow. This may also be due to our reliance upon technology, too, in that these portal monitors do an excellent (although by no means fool-proof) job at detection. An example of this was the detection early this year of contaminated steel plate, which originated in Bulgaria, and which was detected first in Mississippi. Other foreign smeltings which resulted in the exportation of contaminated product (s) were detected similarly.
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| Counting this Bulgaria smelting, there' have been seven cases where contaminated product was exported to the United States. The
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| - radioactive materials involved in these smeltings are also shown.
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| The Steel Manufacturers Association is an organization >
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| : representing 55 steel manufacturing companies wita 107 mills in North America. Their members report their discoveries of abnormal radiation found in (incoming) shipments of metal scrap feed. This is done on a quarterly basis, and the information is added to the incident database. Information they provided proved to be the largest volume of detections in 1994. The Radiation in Resource Recovery (E-23) Committee has been active in this field since it was chartered in 1991. Besides the SMA, we are seeking to have other industry trade organizations adapt a similar reporting scheme. Although the steel industry has been affected the most by smeltings, no metal manufacturing process is immune from the ;
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| problem. What our efforts have shown, however, is how diverse the trade organizations are. As an example, the most recent smelting
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| in the U.S. occurred in a foundry which is not a member of organizations such as the SMA, the Institute of Scrap Recycling Industries, Inc. (ISRI), the American Iron and Steel Institute (AISI), the Specialty Steel Institute of the United States (SSIUS) or other organizations with whom E-23 committee members have (previously) interacted.
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| A good portion of the incidents are reported by the use of Department of Transportation exemption E-10656, for shipments of metal and scrap discovered in transit as containing or being contaminated with radioactive materials. To those of you who are involved with the investigations on these discoveries, we thank you for your efforts and assistance.
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| * The enactment of the North American Free Trade Agreement (NAFTA) is expected to result in an increase in the trade on scrap metal between the Agreement member companies. In this context, it is interesting to note that none of the three member companies in Mexico reported any discoveries, but this may also be due to those companies not having portal monitors. The database includes 21 discoveries by industries in Canada, although the data is not as specific as desired. This may be due to a lack of uniform reporting requirements, but then, no uniform reporting requirements exist in the U.S., either.
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| As many of you are aware, the Environmental Protection Agency (EPA) has been named as the lead federal agency on incidents where abnormal radioactivity has been discovered in the public sector.
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| This would include not only scrap metal radioactivity but also unusual or abnormal radioactivity in such places as landfills. The lead federal agency is responsible for providing technical j assistance, when called upon by the state, in handling these i situations. An early test of this system is underway. Early in l March, 1995, a scrap dealer in Pennsylvania notified the state l radiation control agency that he had found a radioactive device (slide] in a load of scrap from South Carolina. The label on the l device said that it contained 37 GBq (1 Ci) of hydrogen-3. ;
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| Further, the state determined that the device was a generally- !
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| licensed device distributed by a company in California. California representatives informed Pennsylvania that the company had gone out i of business in 1988, and that if any records of the distribution of I generally-licensed devices existed, they were archived, and thus not readily accessible -- to man or machine. South Carolina was also contacted, but their records showed no match for between the device and any of its licensees. NRC has been requested to provide assistance, to help identify the original licensee and to arrange for disposal of this orphan source. EPA, as lead federal agency, has been similarly requested to provide assistance. We would like to report there is a happy ending to this, but one does not (yet) exist. l l'
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| Last September, the E-23 Committee, along with representatives of both other industry trade organizations and several government agencies, met with the EPA about the problem of accidental 4
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| 1 1
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| smeltings of radioactive materials in the steel industry. The EPA is looking at the problem of smeltings and may sponsor workshops or l
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| other venues in their efforts at preventing pollution. One major l problem from the smelting of radiation sources, principally cesium- l 137, is the creation of a mixed waste, since the furnace dust now l contains a radioactive constituent beyond the usual heavy metals.
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| Details of this meeting were given in the November,1994, Newsbrief of the CRCPD.
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| 1 IMPACTS OF SMELTINGS AND DISCOVERIES '
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| In 1993, a scrap dealer in Florida who performs radiation monitoring discovered what turned out to be a radiation therapy treatment head. Fortunately, no source was inside, but the device could have contained a few terabecquerel (kilocuries) of cobalt-60.
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| Had this therapy head gone through a shredder at a scrap yard, the '
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| head would have been flailed into small pieces, and, had a source been inside, the source could have been breached, resulting in the ,
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| contamination of the facility beyond potentially lethal radiation levels from the bare source. In 1994, a gauge containing 12 GBq ;
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| l (330 mci) of cesium-137 was inadvertently disposed of with metal scrap and transferred to a scrap processor. The metal scrap was I 4
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| then shredded and this processing separated the source from the !
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| source housing. Fortunately, in this case, the shredding process '
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| did not result in breaching the source. [As a note on the shredded '
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| source: markings present on the source capsule enabled the regulatory agency to determine the owner of the device. The owner was fined $250 for improper disposal.] Recall that in the Mexican incident the source capsule was breached, and its thousands of millimeter-sized pellets contaminated the junk yard and were also tracked around the neighborhood. We have already noted the unfortunate incident in Estonia which resulted in injury and death.
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| Recall further the smelting of cobalt-60 in Juarez, Mexico of early 1984. A tractor trailer carrying reinforcing rods was discovered as being " radioactive" when it exited a gate at the Los Alamos National Laboratory. Some of the reinforcing rods were used in residential construction and had to be removed. What has come to light recently were stories from the Far East of the r construction of high-rise apartments built with contaminated reinforcing rods. This may have been a serendipitous discovery:
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| in Taiwan, a dentist was establishing an office in a building, and had called upon the local authorities to check the shielding. The authorities discovered high levels of radiation without the X-ray machine being activated! The authorities . traced this to the rcinforcing rods in the walls of the buildings. A Japanese television program showed there may have been several dozen - I apologize for not using an SI unit for this! -- of these buildings constructed. Estimates were that individuals living in these could have received doses on the order of one sievert (100 rem) or more before the exposures were discovered. It is possible that the extent of the usage of this contaminated product and consequently the radiation doses received by individuals residing in or using l
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| l
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| those buildings may never be known with any high degree of certainty.
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| Can this happen here? Yes. These incidents show that radioactive contamination and radiation injuries can result when radioactive materials become mixed with metal scrap. Thus, there is a strong health and safety incentive for developing ways to prevent radioactive sources from entering the metal scrap recycling stream, to detect those sources which do enter it, and to enable the safe storage and eventual disposal of sources which are found.
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| CONCLUSIONS AND RECOMMENDATIONS i The prottem of radioactive scrap is going to be with us for the foreseeable future. To counter the threat, there have been a number of initiatives taken, both to enhance awareness and to solve I the problem of what to do now that the alarm has gone off. I Many of you have seen the warning poster " Radioactive Scrap - i!
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| Beware" issued by the NRC (NUREG/BR-0108). With the EPA now being the lead federal agency, self-adhesive labels to update the notice, ,
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| background, and "where to get help" sections of the posters are '
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| available from the NRC. Note that, although the telephone numbers for (all) the state radiation control agencies were not printed on the poster, it recommends calling the State radiation control agency first. We strongly suggest that when you visit the metal mills and scrap yards, either to gather information or as part of an investigation, leave your calling cards, look at their i procedures and recommend that they contact you (or your agency) first.
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| l Beyond this, the Institute of Scrap Recycling Industries, Inc. !
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| (ISRI) has published its second issue of their Recommended Practice !
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| and Procedure " Radioactivity in the Scrap Recycling Process," which )
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| was also distributed to the Conference. ISRI also has produced a i videotape about this, and copies have been sent to ISRI members and I to the state radiation control agencies. A Spanish version of this I videotape is also being prepared. The Radioactivity in Resource Recovery (E-23) Committee participated in providing the technical -
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| expertise for the ISRI document and videotape. l, As stated before, the use of radiation monitors has been the ;I most effective measure in preventing accidental smeltings. These, ,.
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| however, cannot guarantee 100.00% protection. This can be :l demonstrated by a few mills having had the unpleasant experiences !'
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| of a second smelting. But monitoring of incoming scrap at a steel mill may be represent the least effective time and place for such monitoring. We recommend that you monitor early and often -- when it is received by the " mom and pop" scrap dealers, when it leaves them to go to a larger facility, at this facility, before it is bundled or shredded -- in fact, all along the line. We also recommend that users and suppliers of scrap consider specifying in their contracts that the scrap has been monitored and found to be
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| free of abnormal radioactivity. This, however, leads to another issue, that of the lack of consensus standards for monitoring equipment or for acceptable levels of NORM or even other radioactive materials. We need to work further with the affected industries and other federal agencies, most notably the EPA, in the development of such standards or levels. As an example of the concern about equipment, during April, 1995, the SMA sponsored a
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| : gauntlet test, in which manufacturers of scrap monitoring systems were invited to install their radiation monitors for tests of system sensitivity in detecting a 7.4 GBq (0.2 Ci) cesium-137 source buried in different configurations of shredded scrap. Eight companies participated in this test, which was held April 17-28.
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| I The DOT exemption for scrap metal radioactivity, which this Conference sought, has been in existence since 1991, and has undergone one extension (another will be sought, since these exemptions are for a duration of two years only) . This has been an effective means of providing assistance to the facilities where the scrap metal has been discovered as radioactive. The assistance this brings can prevent workers from unnecessary radiation exposure or even from breaching a discovered source of radiation at a facility.
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| The NRC has been requested to improve its control over, and licensees' accountability for, licensed devices. This would apply not only to specific licensees but would also involve looking at the concepts behind the general license philosophy as well. The presence of gages in scrap metal loads indicates a breakdown in the control of the devices. If a device makes its way into a steel mill and is smelted, it may cost that mill upwards of $10,000,000 (US$) to decontaminate the mill and dispose of the radioactive contamination, assuming that there is a place to send the contaminated material and that no mixed waste was generated.
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| Although NRC is looking at regulatory alternatives to improve the control and accountability of sources and devices by its licensees, the truth is that the number of radioactive sources that inadvertently enter the scrap stream can never be reduced to zero (assuming that radiation sources will continue to be used and produced). The industries involved with metal recycling must r maintain and enhance their vigilance. An estimate is that only two thirds of the steel mills and only half of the scrap metal dealers have any sort of radiation detection instruments, either fixed or portable, to screen incoming metal for radioactivity.
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| To help address this part of the problem, the E-23 Committee developed a position paper advocating universal radiation monitoring by all persons who handle metal scrap. The CRCPD Executive Board endorsed this position paper, and it is being circulated for comment, and, we hope, endorsement by industry trade organizations. Eventually, we seek to gain appropriate governmental agency endorsement as well. Initial reactions have been favorable; such a statement can provide needed leverage for mills to include provisions in scrap metal purchase contracts to
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| - - - .. - . - . . . . - . - - . . . . . . . . . . . . - . - . . . . = - . _ . . . .. - _
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| 1 require suppliers to perform radiation monitoring of the metal scrap at the processing or recycling facilities where the volumes and densities of the scrap metal are smaller, and thus less likely to shield a radiation source from detection.
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| If this tactic succeeds, regardless of government action to improve control over licensed sources, we should see a trend of fewer sources uncovered by the metal manufacturing facilities, which should ultimately result in fewer smeltings, we will still need to address the problem of providing for safe storage and ultimate disposal of found sources, but we will be moving in the right direction.
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| i N
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| l Smelting of radioactive materials
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| ; summary of U.S. occurrences i
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| !
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| * Since 1983, twenty-four (24) domestic instances of the accidental smelting of radioactive sources i
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| * 15 Cs-137; 1 Co-60; 3 Ra-226; 1 Th; 3 Acc.; 1 Am ;
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| 5 (FIGURE 1)
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| {;
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| * 16 occurred in steel mills, 3 Al; 2 Aui 1
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| * Cu,Pb,Zn j
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| * Events discovered by monitoring of slag / dross (4)
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| ! or of flue dust shipments (15). One discovery occurred at a highway weigh station.
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| i
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| * There were 12 foreign events, in addition to these:
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| l Mexico, Taiwan, Brazil, Italy (3), Ireland, India,
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| ! Russia, Estonia, Kazakhstan, Bulgaria. Others??
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| ) (revised January,1996) 1 .
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| Discoveries of radioactivity ,l l
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| in metal scrap (selected cases) !!
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| * reported / documented U.S., Canadian cases only '
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| * Time Span: January,1985 - December,1994 (FIGtlRE 2)
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| * aSeS $7 dSCopeS, M SmeWngS) l'
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| * How discovered:
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| - Stationary monitors -1053 (95.7%)
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| - hand surveys - 19 (1.7%) ;
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| - warning labels -
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| 9 (~1%)
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| - unknown - 19 (~1.7%)
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| - other -1 (~0.1%)
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| (revised AprH 21,1995) JGY f
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| Discoveries of radioactivity in metal scrap (continued) thru 12.31.94
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| ~
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| a wmmww- "w Radium gricium-2418
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| [""~' .dky Uranium 22
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| __ Cobalt-60 6 u 0 (FIGURE 3) NORM 706 ----3555'"? ,.
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| Cesium-137 4 H,Kr Sr 3 unknown 228
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| \
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| accel. prod. 4 (N=1101; revised April 21,1995 JGY Events, Smeltings of RAM
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| " selected" (= reported) cases
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| # / year 3g 276 -
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| 250 -
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| 226 --
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| 200 -
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| 175 --
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| 150 - -
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| 125 - -
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| (FIGURE 4) 100 - - - -
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| 75 - - - -
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| 60 - - - - --
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| , i i iI,I, , , , , , ,
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| 1983 1984 1986 1966 1987 1988 1989 1990 1991 1992 1993 1994 Year I i Events E US Smeltings as of December 31,1994 I
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| \
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| l 4
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| l
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| ! i
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| ! l l Sources, devices found in metal scrap, 1983 - 1994 40 35 -
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| f 30 -- -
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| 25 14 20 - --
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| M i --
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| (FIGURE 5) g , _
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| 16 - - - - -
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| 5 -- - -
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| 0 1983 1984 1985 1986 1987 1988 1989 1990 1991 1992 1993 1994 j C Ra-226 E Unknown C Cs-137 E Co-60 C Thorium C Uranium E other N = 141; 4.24.95 JGY l
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| World-wide Smeltings of Radioactive Sources ,!
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| {
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| l tron / Steel !
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| t Aluminum e
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| (FIGURE 6)
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| Gold -
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| j Copper i
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| Other (Pb, Zn) j O 2 4 6 8 10 12 14 16 18 (N = 36; Feb.1995)
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| .e # '
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| Q. ,
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| /
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| l l
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| IMPROPER TRANSFER / DISPOSAL SCENARIOS l i FOR GENERALLY LICENSED DEVICES Manuscript Completed: April 1987 1
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| Prepared by l Michael Stabin, Kermit Paulson, and Shelly Robinson /0RAU i
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| l Oak Ridge Associated Universities Oak Ridge, TN 37831-0117
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| ~
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| Prepared for Division of Fuel Cycle and Material Safety U.S. Nuclear Regulatory Commission
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| : Washington, DC 20555 NRC FIN B0299 l
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| m e
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| L ~4 1=1uceLP m , g. M P.
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| . , _ _ _ _ ., - . - - - - _ . ~ - . . - = _ _ _ . _ - _ - .._ _ -- .
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| l Ii Abstract l
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| A recent study by the Nuclear Regulatory Commiesion revealed areas l of safety concern related to the imptoser transfer or disposal of certain i generally licensed devices containing radioactive material.
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| provides an evaluation of the potential for inadvertent exposure of theThis report public to the radioactive material in these devices and gives some estimates of the radiation doses which could resultfrom such exposures.
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| i Devices listed in an NRC registry were organized into sixteen categories and scenarios were developed which predicted the probabilities of a ,
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| device in a category passing through a specific process, being contacted by a partic"lar Radiation s. ,e type of individual, and ending up in a given situation.
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| derived for twoestimates for of broad types external and internal irradiation were situations: ,
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| 1
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| ; . a device over dispersed is essentially a wide area. intact and one in which the activity has beenone in which th(
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| \
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| a 1
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| i d
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| i l
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| 4 d
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| i k
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| 1 l
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| 1
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| l j
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| i Exocutivo Summary i Many products i members containing radioactive material may safety programs of the general because public and industry without extensive radi j regulated under general the nature of the products allows them atoon be
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| : Regulatory Commi.sion licenses. A recent study by. the Nuclear i to the improper transfer r4'ealed'several or disposal . areas of safety concern related i unfamiliar with license of these products by persons I consequences of such requirements. This devices improper handling of report analyzes potential i containing sealed sources. sixteen classes of such j
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| resulting from the improper transfer or disposal derived. Probabilities of certain events vi of these'd 8
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| individuals to Radiation dose estimates were calculated for. e ces wereexposur generated throughradioactivity from these devices. Although the results this analysis were based on a small amount of a, dat some conclusions about public health concerns are evident .
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| ; The two major i internal irradiation. dosimetry concerns radiation field The former are external irradiation and j
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| }
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| the source. Internal from the radioactive source in a device or contact 1 the device, in which irradiation may occur from improper handling of intake of activity may be released from the source, or
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| }
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| s because of radioactivity which has been metals recycling, the device inadvertently being incinerateddispersed over a vida area or buried in a landfill, , passe through
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| ; radioactive material into public drinking water supplieswith' leaching o ,
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| i Exposure most of the devices, even under a worstosure 20 weeks case or ' assu for j
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| ~
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| were greater than 500 mrea (0.005 Sv) samma include all typ 1
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| (class g), X-ray fluorescence gauges
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| ; self-luminous devices containinganalyzers cases, Kr-85 (class (class E-1 ami E-2), and
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| { contact G-1). In almost all
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| { to localized radiation doses the radioactive source for three hours will produce
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| ; general and these public.
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| dose For many devices, the source is extremely e, i chromatographs, analytical equivalents would not be expected (e.g., gas instruments i sources). In with calibration or reference through a other built-in device mechanism, types,and however, significant theradiati source can be accessed if the device. is manipulated by persons on doses can occur hazards.
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| unaware of the radiation Internal radiation doses worst ' case situation were derived for several situations. A material by inhalationwas defined as intake of 30% of the radioactive or ingestion. More realistic numbers were generated 10*'. by assuming a much smaller fraction, usually between Estimates 10 6 were also generated for and materialwhich material if a device may have wereleached incinerated, from aand for ingestion of radioactivinh e situations, landfill site. For these' and activity level of the radionuclida concerned.
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| were y Sign derived for inhalation of An 241, Ra-226, Cm-244 mates
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| , Pm-147 Pu-239, i
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| and, to a lesser extent, Po-210 Pb-210, Fe-55, Ni-63, Sr-90, Co 60, Cs-137, and Cd 109. High dose equivalents were estimated for ingestion of Am 241, Car 244, Pm 147, and, to a lesser extent, Po-210, Pb-210, Fe 55, Sr-90, Co-60, Cs-137, and Cd-109. Dose equivalents from intake of material from the plume of an incinerator were highest for Am-241, 22 226, and Cm 244. Ingestion ,2 radios: ' material from landfills were significant for Ra 226 and Sr 90. The model used for transport from the landfill predicted very long holdup times in the landfill soil matrix for some nuclides, which resulted in the prediction that no Am-241, Cm-244, or Pm 147 would be ingested. These nuclides were all significant hazards if ingested, and would be of concern if the-migration rates were significantly faster than predicted by this model.
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| U3 Although more literature was available for consequences of metal recycling of radlonuclides than for any other category, only quantitative information was listed for Co 60. Therefore, this analysis was performed only for Co-60 in gamma gauges. Inferences based on available literature showed that dose equivalents received by members of the general public who purchase contaminated products would most likely not exceed 500 mrem /yr (0.005 Sv/yr) in most cases.
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| l l
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| ii
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| )
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| TABLE OF CONTE.NTS i 4
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| l I
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| Introduction.... ................................ ................ 1 Methods.....................
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| 4
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| ..................................... 1 Resul:s Portable Static Eliminators..... .......................... 15 Static Eliminators / Detectors (High Toxicity)............... 22 Static Eliminators / Detectors (Low Toxicity)................ 30 Camma Gauges....................
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| .......................... 34 Beta Cauges: Backscatter Type.
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| ........................... 46 Beta Gauges: Transmission Type............................ 58 Cas Chromatographs......................................... 63 X-Ray Fluorescence Analyzers (High Toxicity)............... 68 X-Ray Fluorescence Analyzers (Moderate Toxicity)........... 74
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| . Calibration or Reference Sources........................... 81 Self-Luminous Devices...................................... 92 i Self Luminous Devices in Aircraft.......... ............... 98 Analytical Instruments with Calibration Sources. . . . . . . . . . . 104 Calibration or Reference Sources (Am-241),................
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| 110 .
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| l i
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| Small Quantities of Source Material.............. ........ 114 Calibration or Reference Sources (Pu-239). .
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| i
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| .............. 115 S umma ry and Conc 1uc ions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 120 !
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| l R e f e r e nc e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
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| Appendix. Device Descriptions...................................A.1 111 l
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| l ll
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| . ._ ~. _ _ . - . . _ - - - . . _ . ._
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| : 1) Intreductien The Nuclear regulate the Regulatory Commission (NRC) and the Agreement States distribution and States that contain byproduct use of all products within the United material. The NRC classifies the 4
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| regulatory specific * ' control license, of byproduct general license,material into one of three categories:
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| or c'.assification depends on the fpe, quantity, and use of the materialexempt .
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| Many products general public andcontaining radioactive material can be used by the programs. These industry products contain without extensive radiation safety relatively small amounts of radioactive materials that are sealed without within the device (se so that they can be used by persons radiation small safety.ofThe amounts use material, source and distribution of such products, as well asforma licenses. General licenses are in effect are controlled through general radioactive materials without the filing offor NRC.
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| persons using certain an application with the Estimates indicate that over 240,000 such devices are in use in the United States under six categories of such licenses.
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| A study conducted by the NRC in 1984, 1985, and 1986 revealed several areas devices under of safety concern about the use of some sealed source general license.
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| accountability for some devices was Investigators observed inadequate and that users were that frequently be located unaware of regulations. Furthermore, some devices could not determined byand finalordisporition of some devices could not be the user the NRC.
| |
| This report summarizes work that was done to develop potential I scenarios for improper transfer or disposal of these devices and provides an assessment of both realistic and maximum dose equivalent to persons potentially equivalents for involved in each scenario. Additionally, dose indicate a pathway to the general public.the general public are given for
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| ; 2) Methods 2.1) Devices considered device The NRC maintains designs that area nationwide registry of sealed sources and data base was developed deemed acceptable for licensing purposes. A from this registry for over 600 examples of licensed arranged into devices containing radioactive materials. The devices were j activity, relative categories based on such charactistics as radionuclide, registry, etc. This radiotoxicity, resulted principal use, manufacture, date of devices for scenario development; in the selection of sixteen classes of these classes are listed in table 1.
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| 4 those An gamma analysis andof hazard to the public was performed by the NRC for i beta gauges having activities greater than 20 millicuries the gamma (740 MBq). Therefore, for this assessment, the activity for mci (740 MBq). and beta gauges (classes B, C-1, and C-2) was limited to 20 May 12, 1987 l
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| 1
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| : 1) Introduceion The Nuclear regulate the Regulatory distribution andCommission (NRC) and the Agreement States States that contain byproduct use of all products within the United material.
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| regulatorf ' control of byproduct The intoNRC oneclassifies of three the specific license, general license, material or categorie s:
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| lassification depends on the exempt from regulations. The type, quantity, and use of the material.
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| Many products general public andcontaining radioactive material can be used by the programs. These industry penducts without extensive radiation safety contain relatively small amounts of radioactive so that they can materials be used by that are sealed persons withintraining without formal the device in radiation safety. The small amounts of source material,use and distribution of such products, as well as licenses. General licenses are in effect are controlled through general radioactive materials without the filing offor NRC. an persons using certain application with the Estiestes indicate that over 240,000 the United States under six categories of such licensessuch devices are in use in A study conducted by the several areas of safe ty NRC in 1984, 1985, and 1986 revealed devices under concern general license.about the use of some sealed source accountability for some devices was Investigators observed that frequently inadequate and,that users were be located unaware of regulations. Furthermore, some devices could not determined byand finalordisposition the user the NRC. of some devices could not be This report summarizes work i that was done to develop potential '
| |
| scenarios for improper transfer or disposal of these devices and provides an assessment of both realistic and maximum dose equivalent to persons equivalents potentially for involved in each scenario. Additionally, dose indicate a pathway to the general public.the general public are given f
| |
| : 2) Methods 2.1) Devices considered device Thedesigns NRC maintains that are a nationwide registry of sealed sources and data base was developed deemed acceptable for licensing purpessa. A licensed devices containingfrom this registry for over 600 examples of radioactive arranged into materials. The devices were activity, relativecategories based on such charactistics as radionuclide, registry, etc. This radiotoxicity, principal use, manufacture, date of devices for scenario development;resulted in the selection of'fourtee classes of these classes are lis'e in table 1.
| |
| An analysis those gamma and of hazard to the public was performed by the NRC for beta gauges having activities greater than 20 millicuries the gamma (740 MBq). Therefore, for this assessment, the activity for mci (740 MBq).and beta gauges (classes B, C-1, and C-2) was limited to 20 April 10, 1987
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| a -- m _ _-..----m.a a_. -m..- . ._..-._m,_,.a. _ m 2A, , . _ _ , ,
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| 1 11 l1l
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| \
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| i 1 II .
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| 4 1
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| j.
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| I[l in e x lil !
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| l}i y, i B h 1 i I i . 1.t p i- i.l In 2
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| A brief description Appendix. Radiotoxicity of r. f each type of device is given in the the classification of radionuclides according to relative per o oxicity radi tt A unit activitymay radionu*clide given in Safe Handling of Radionuclides (IAEA 1973) be classified in one of four groups: a) very high .
| |
| low radiocoxicity. radiotoxicity, b) high radiotoxicity, c) moderate , and d) radio 2.2) Scenario Development for The t4sk of developing scenarios for improper transfer or dispossl each class of device began with listing a wide variety of possibilities.
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| manageable number The number of oossible events was reduced to a familiar with the through devicesliterature and searches, interviews with people reviews, disposal technologies, and committee beginnings)The result that would isresult a set of six different initial events (scenario
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| " event trees" leading to in improper transfer or disposal, with endings). A r ight potential final conditions (scenario rounded " generic scenario" is illustrated by Figure 1. In this I
| |
| figure, rectangles rectangles are assigned to initial events, squared represent j
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| intermediate ellipses represent final conditions. processes or individuals, and listed table 3.
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| in table 2 and the potential The different initial events are final conditions are listed in Probabilities were assigned to the intermediate process pathways on initial event pathways and inferences about the basis of available literature, knowledgeable sources devicesuchcharacteristics, and personal interviews with 3
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| literature search was performed; as manufacturers and users. An extensive i could be used to assign probabilities.however, few data were found which educated guesses and may The estimated probabilities are Data may vary be subject to many sources of variability.
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| ordinances. from region to region depending on local practices and Assigned probabilities for event or intermediate process add all pathways leading from an initial leading into an up to 1.0. Pathway probabilities the total intermediate process or final conditions add to give intermediate probability events, of a device reaching that point.
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| these probabilities For appropriate are then multiplied by the intermediate pathway probabilities process or final condition. to give the contribution to the next <
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| for all final conditions also add up to 1.0. The resulting probabilities k number was used More than one significant estimates and should be to calculate the results, but final results are only figure. considered accurate to only one significant Probabilities assigned to the various pathways are based on the assumption that a device has been probability of 1.0 mishandled. That is, the total mishandled, for all it'itial events applies to the devices each device not the total type, number of devices in use. Table 4 shows, for
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| ~
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| the 4nown number of cases of mishandled devices.
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| This report presents possible scenarios, probabilities assigned to April 10, 1987 3
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| I
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| those scenarios, and dosa estimatas to th3 public, assuming that a device has been mishandled in some way.
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| 2.3) Dose Assessment situations.
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| Radiat'i'on dose estimates were derived for several diffe for a particular The largest class source activity recorded in the NRC regi try doses for that device. f d:vice was sed to estimate the radiation public was estimated The radiation hazard to members of the general itself for (1) an ' intact' device, in which the source dispersed, is not damaged and or only slightly damaged and the activity is not (2) a device breached and the activity has in which the source integrity has been According to the been dispersed over a wide area.
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| might contact scenario format used, members of the general public two general radioactive material at several points. Because of the dispersed sources), of radiation hazard envisioned types
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| (' intact' and with one or the other these typevarious members of the public were associated of hazard.
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| from both external and Dose equivalents were estimated expmed otherwise and average noted, individual and for the exposed population. Unless in was estimated to the activity intake value for the average individual be half 1980). of that for the maximum individual (NUREG The be used to L dify the associated dose estimates; the scen estimated if it is mishandled. probabilities that a device will follow a particular pathway a device, If a member of the publi: :omes into contact with a theoretical probability.the dose that the individual receives will not be a ,
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| 2.3.1) 'Intace' source Because individuals the source might be is assumed to be mostly intact, only a few the scenarios defined in thir report, expected to contact the source atInany time.
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| intact source individuals who might contact an device, salvage include trash dealers, handlers, persons who find or receive the and workers incinerators, at sanitary landfills, such situationsor metal recyclinh Plants. The dosimetry concerns for exposure to the include external doses, either to the whole body from internal doses, radiation field or to the skin from contact, and from inhalation or' ingestion of some of the activity.
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| Maximum and realistic dose equivalents were estimated for external and internal exposures to activity in the source. Because only one or, at most, a few individuals would be expected to contact ' intact' sources, population dose equivalents were not calculated. !
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| geometry.WholeCamma body doseray equivalents were estimated based on a point source and Trubey (Unger dose constants were taken from a report by Unger 1981),
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| 1970), the Radiological Health Hnadbook (USDHEW or AECL 7617 (Cross 1982). Some estimate of the time and distance relationships involved in a worst case scenario may be derived from the incident in Mexico in 1962 (Andrews 1963) in which a young boy April 10, 1987 4
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| 1
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| found a Co-60 source in a field, carried it around in his pocket for several days, and then left it in his home for 15-25 weeks. Five people, including an unborn child, died in that incident; however the source intensit,y. was much higher in that incident than would be involved with the sources considered in this report. The average distances were not kvawn with any reasonable accuracy, but 3. 3 feet (100 cm) may be used as the average distc ae from a source in the center of a small room.
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| Surface dose rates to the skin were estimated from the values for sealed sources in NCRP Report 40 (NCRP 1972) or from a study by Kocher and Eckerman (Kocher 1986). The values in NCRP 40 are for encapsulated 4
| |
| sources of Co-60, Cs-137, and Ra-226, and therefore only include photon emissions. These values are given with the qualification that the listed values must be increased by 25-45 percent to account for electron production in the stainless steel walls assumed to encapsulate the source. For the calculations in this report, the values were all 1 increased by 45 percent. Values in the report by Kocher and Eckerman include only beta emissions. Dose rates at a depth of 7 m /cm s 2 in the report by Kocher and Eckerman were used to estimate the maximum dose from an unencapsulated source to healthy skin. No reference was found that has information needed to estimate the photon dose rates from contact with other sources. For Co-60, Cs-137, and Ra-226, dose rates for the encapsulated and unencapsulated source are given, based on the information in these two reports. l If an individual carries a source for several hours in contact with the skin, some transitory or permanent effects may occur. In the Mexican incident, the source was carried for several days, but was not in direct contact with the skin. Radiation symptoms were evident, but f were not identified as such, even after the child's death (Andrews 1963). Because of the size of most of these devices, they cannot be carried in a pocket, so a maximum contact time of three hours was hypothesized.
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| Internal dose estimates were based on the fifty-year dose equivalents per unit ingestion or inhalation of the individual radionuclides listed in ICRP Publication 30 (ICRP 1978). Because the chemical form of the material was not known, estimates were derived for all listed inhalation classes (D, V, or Y) and values of gastrointestinal absorption (ft ).
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| Intakes for the maximum individual were hypothesized based on device characteristics. To derive intake values for the more realistic case, available literature, which presented intake fractions for the general public in certain accident scenarios, was studied. In an scenarios involving spent fuel analvsis of possible accident transportation (Wilmot 1981), the maximum fraction of Cs-137 released to the environment in the form of respirable material was postulated to be 10~3, that for Co-60 was 10-1, and that for actinides was 3 x 10-6, These scenarios involved impacts, immersion in water, and punctures. A report (Ricks 1981) prepared for the Department of Transportation (DOT) cites a DOT rationale that in an accident of moderate severity, 1/1000 of the contents of a package containing radioactive materials might be April 10, 1987 5
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| . _ _ ._ .. _ _ _ __m - . _ . _ _.
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| l l
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| released. The authors further hypothesize that an individual in the l vicinity of the accident might take in 1/1000 of the released material, l implyingthatthefractionoforiginalactivitywhichmightbetakenup The.. fraction released (10-3) is of the same order of l is 10" .
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| y magnitude as the value specified by Wilmot for Cs.137, but is lower
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| < than his assumed values for Co-60 and higner than his assumed value for l actinides. Although the type of taclosures
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| * devices in these two l studies were much different raan those described in this document, i these numbers give some guidance for postulation of the amounts of 3
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| i l materials which could be released from these devices. If a radionuclide
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| ) !
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| was not specifica11/ treated by these studies, an attempt was made to use the maximum value listed for a similar nuclide, i
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| l i 2.3.2) Dispersed source I
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| If the source in a device becomes extensively damaged through a destructive process, the activity may be dispersed over a large area with the possibility that a large number of people may come in contact with smaller amounts of activity than would be encountered with an
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| . ' intact' source. This might occur at an incinerator (with material 4
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| subsequently released to the environment or buried in a landfill), in metals recycling (with material subsequently incorporated into consumer products or construction materials), or at a landfill (where material I
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| may leach from tbs landfill into nearby surface water or groundwater I supplies and be ingested).
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| , l 2.3.2.1) Incineration i Incineration is used by some cities and, in some urban tress, in l
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| } apartment buildings and other smaller applications to reduce the volume fii
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| ' of municipal wastes. Refuse may be pretreated (size sorted, shredded, or ground) and is oxidized at temperatures between 1000 and 1500
| |
| ' degrees Farenheit (540 and 820 degrees Centigrade). The residue, which consists primarily of metals, glass, and other unburned materials, may then be compacted and sent to a landfill or to salvage if the metals can be identified and isolated. Some residual materials may be used in building or road construction materials (NUREG 1980),
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| t
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| /
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| Effluents from incinerators are carefully controlled by combustion removal technologies, with particulate removal i and particle efficiencies of over 0.9 in most cases. Buckley et al. (NUREG 1960)
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| ' estimated the amounts of activity expected to be inhaled by the maximally exposed individual, the average individual, and the '
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| population near an incinerator which burns consumer products containing radioactivity as part of its normal refuse. Assumptions needed to derive the estimates include the number of devices incinerated per year, the fraction of activity which would be released by the incineration process, and the traction of released acervity which would escape with stack emissions. The other dose calculations in this report
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| ! were done on the basis of activity from one device. Because this model j requires input of a number of devices per year, a value of 5 devices
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| ; per year was arbitrarily chosen for model input. With the exception of portable static eliminators (which were treated differently), the devices were assumed to be incinerated at one site, with a nearby f April 10, 1987 6
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| ]
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| c
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| populction of 73000.
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| estimation Other built in modal assumptions allow the of the activity individual and the surrounding inhaled by the average and maximum population.
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| is known, the Once the activity inhaled the committe,d dose equivalents. standard models of ICRP 30 can be employe nedicalSeveral authors have studied or other institutions. incinerators.the release Materials of include burned radionuclide medicalcarcasses, animal and research wastes (planchets, filter paper, syringes, etc.),
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| and mixed general wastes.
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| findings, which were Table 5 summarizes these may be released during usedincineration.
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| to estimate the fraction of material which consistent, with The results were fairly Cs-137 retained inthe majority of the Co-57, Co-60, Rb-86, Ru-103, and I-125, and T1-204 were the incinerator ash, while most of the H-3, C-14, released. A significant difference exists between the results of Bush and Hundal for cobalt, with Classic et al. and those of Brekke et al. and higher fraction (0.273) than the Classic et al. predicting a much largest value for a particular element was used other authors (0.006-0.038). The fraction of burned material which t'o estimate the incineration. A would be released during value specifically listed in the table, of 0.3 was cypically assigned to nuclides not to special although higher values were assigned et al. that radionuclides 0.1 (e.g. H-3, Kr 85). The assumption of Buckley released of the material released durin5 incineration is from the stack, as well as other assumptions about the number of incinerators, dispersion characteristics, etc., were~used directly.
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| 2.3.2.2) Metals Recycling Lubenau involving licensedand Nussbaumer (Lubenau 1986) have reviewed incidents devices containin5 Co-60 and Cs-137 that have been inadvertently incorporated materials. into consumer products and construction surveyed. They quoted surface exposure rates for some products of 80 uR/hrExposure rates near contaminated products were in the range (21 nC/kg-hr). This was the reported exposure rate near steel products from the Brazilian activity concentration in contamination incident; the measured 1985). the products was 26 pCi/g (0.96 Bq/g) (NRC activity Products involved concentrations ofinup thetoTaiwan contamination incident had an 150 pCi/g (5.6 Sq/g) and exposure rates on contact of 100 uR/hr (26 nC/kg hr). The activity concentration expected in steel contaminated with activity from products such as those considered in this report may be implied from a report on the Auburn by Steel(Lubenau Lubanau company contamination 1985). The former incident (NUREO 1986) and a report report su5 gests a value of 50 tons for contaminationone steel melt; the report of Lubenau analyzes consequences of of melts of up to 700 tons. Using the 50 ton value, the activity concentration from loss of a 10 mci (370 MBq) Co-60 source would be 220 pCi/g (8.1 Bq/g). Based on the average value for exposure
| |
| :2te (1.9 uR/hr per pCi/g), this would result in a product with an exposure rate of 410 uR/hr (106 nC/kg-hr) . Continuous exposure to this product however, would result in an annual dose equivalent of 3.6 rem (36 mSv);
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| the dose rate will fall off quickly from the surface value.
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| Using equivalent a reduction factor of 0.1 for distance, the annual dose to an individual would not exceed 500 mrem (5 mSv). If the April 10, 1987 7
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| cetivity the productio inand a ccnstruction currounding mate.ricl, shielding of the radiation from equivalent. structures will further reduce the dose 104 The steel kg) each billets 1986).
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| (NUREG from the Auburn plant weighed about 55 tons (5 x lb (350 g) each (NRC 1985). The Taiwanese steel fittings weighed 0.77 ould result from choosing che The maximum number of persons exposed smaller of these two numbers. If the product was 350 activity g, concentration were 220 pCL/g (8.1 Bq/g) and the weight possibility that upwould this imply 77 nci (2.8 kBq) per product, and the to 130,000 the product weight were 5 x 10' g, the nunber of products would Therefore, .
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| be such an between incident. 1,000 and 100,000 products may be contaminated in equivalents of betweenThese 2.7 x values 102 andimply annual population dose 2.7 x 104 person-rem for the product exposure rate assumed above distance and using the factor of 0.1 for factor of 0. 75. assuming one person per p,ro6tet and an average occupancy the Thisstatedresult only applies to co-60 gauges of 10 mci (370 MBq) under assumptions.
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| this subj ec t , Because of the paucity of data and models on radionuclides and devices cannot be calculated.more refin 2.3.2.3) Burial in Landfill Buckley et from disposal of al. (NUREG 1980) also discussed hazards to the public concluded radioactive consumer products in landfills.
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| of suspended that particles, of three major pathways (direct irradiation, inhalationThey only and ingestion of contaminated food or water) j ingestion of contamination would be of concern because of the low activities activities and and energies of the nuclides involved. Although the be higher, energies of the nuclides considered in this report may suspension of radioactive particles with subsequent inhalation from is probably not a significant pathway to consider. Risks direct irradiation are highest for exposure to the intact source so ingestion of contaminated groundwater is the only pathway considered in this report.
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| rainfall Through a study of. landfill characteristics, soil characteristics, et activityal. constructed a model which allows estimation of total a ;
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| includes ingested per year from a given type of device.
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| The model in the contributions from sources which have been directly deposited incinerator landfill beforeas well as those reaching which have passed through an the landfill.
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| i typically has a higher fraction of activity release.The latter type of device Parameters assumptions of Buckley for leaching et from landfills were derived from the Site' and for the various radionuclides.al.
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| reference soil were given for all of the Retardation for their defined ' Reference factors for their this section. Leach radionuclides considered in products rates of 0.01 yr'1 and 1.0 yr-1 from intact and previously incinerated products, respectively, were April 10, 1987 8
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| j
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| assignad.
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| withdrcwel point was calculated for the radionuclides This c
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| parameter allows calculation of the total activity ingested from l
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| landfill leaching. v applied to.- estimate the organ committed dose equivalents. Stan Buckley et al. ;
| |
| individual dose for did not explicitly define an average and maximum '
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| enat the might same be affected by activity leaching from a single If lan) '
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| individual criterion as for the incineration model is used, half might be expected to receive a dose equivalent the average as large as the maximum individual. that is one They do state that there are 18,500 landfill population servedsites in the by one United landfill wouldStates, be so a first approximation of j
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| 12,000 (220,000,000/18.500)
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| This is a crude estimate, because many of these landfill sites are small
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| . i and population. because landfills are not evenly distributed with respect to -
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| l l
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| i-i April 10, 1987 9
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| l TABLE 1 CLASSES OF DEVICES FOR SCENARIO DEVELOPMEN APPLICABLE RECULATORY SECTION** RADIONUCLIDES AND
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| . C. LASS DEVICE MAXIMUM ACTIVITIES 31.3 A1 Static Eliminators: Po-210 - 0.50 mci (18.5 MBq) '
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| Hand Held /Pcreable/
| |
| Small Brushes 31.5 A-2 Static Eliminators Po-210 - 100 mci (3700 MBq) or Detectors: In Equipment or Process l
| |
| Line Am-241 - 0.0005 mci (0.0185 MBq) l (Very High Toxicity) Ra-226 - 0.0005 mci (0.0185 MBq) 31.5 A-3 Static Eliminators H 3 - 250 mci (9250 MBq) or Detectors: In Equipment or Process Line Kr 85 - 2 mci (74 MBq)
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| (Low Tox* city) r
| |
| .5 B Camma Gauges Co 10 mci (370 MBq) I Cs-137 - 20 mC1,(740 MBq)
| |
| ! Am-241 - 20 mci (740 MBq) i Ra-226 - 10 mci (370 MBq) 1 i
| |
| 31.5 ;
| |
| l C-1 Beta Cauges:
| |
| Backscatter Type Sr 0.025 mci (0.925 MBq) l i
| |
| ( T1-204 - 0.10 mci (3.7 MBq)
| |
| Ru-106 - 0.025 mci (0.925 MBq)
| |
| Pm-147 - 0.050 mci (1,85 MBq) i
| |
| ! C 0.050 mci (1.85 MBq)
| |
| Pb-210 - 0.010 mci (0.37 MBq) 31.5 C-2 Beta Cauges:
| |
| , Transmission Type Sr 20 aci (740 MBq) 31.5 D Cas Chromatographs Ni 20 aCi (740 MBq)
| |
| H 1000 mci (37 CBq) 31.$ E-1 X-Ray Fluorescence Analyzers Am-241 - 30 mci (1100 MBq)
| |
| .! (Very High Toxicity) Cm-244 - 100 mci (3700 MBq)
| |
| April 10, 1987 10 I
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| i
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| | |
| TABLE 1 CLASSES OF DEVICES FOR SCENARIO _ DEVELOPM APPLICABLE REGULATORY SECTION CLASS RADIONUCLIDES AND I
| |
| DEVICE MAXIMUM ACTIVITIES ;
| |
| 31.5 E-2 X Ray Fluorescence Analyzers cd-109 - 20 mci (740 MBq)
| |
| ! (Moderate Toxicity) Fe 100 sci (3700 MBq) !
| |
| 31.5 F Calibration or Cs-137 - 0.10 mci (3.7 MBq)
| |
| Reference Sources ,
| |
| i l Co 0.01 mci (0.37 MBq) '
| |
| Ra-226 -~0.004 mCiL(0.15 mBq)'
| |
| Sr-90 .0.001 mci (0.037 MBq) i 31.5 G-1 Self Luminous Devices H 5000 mci (185 CBq)
| |
| ~
| |
| Kr 1700 mci (62.9 CBq) i i
| |
| C 0.10 sci -(3.7 MBq) 31.7 C-2 .Self-Luminous H 5000 sci (185 CBq)
| |
| Devices in Aircraft Psa l47 - 300 mci-(11 CBq) >
| |
| 31.8 H Analytical L
| |
| Instruments Cs-137 - 0.040 mci (1.5 MBq) l Containing Small Ni-63 j Calibration or 15 sci (555 MBq) j Reference Sources !
| |
| I 31.8 I Calibration or An-241 - 0.005 mci (0.185,MBq) l Reference Sources 40.22 J Small Quantities of U-238 and Th-232 - 15 pounds at Source Material any one time, no more than 150 1
| |
| pounds per calendar year 70.19 K Calibration or l Pu-239 - 0.005 mci (0.185 MBq)
| |
| ! Reference Sources
| |
| * See Appendix for device descriptions
| |
| ** Code of Federal Regulations, Title 10 i
| |
| April 10, 1987 i 33
| |
| + , -. . -
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| | |
| TABLE 2 INITIAL EVENTS (SCENARIO BEGINNINGS)
| |
| I. Owner takes device out of use. Device remains in place.
| |
| Loss of control results.
| |
| II. Owner takes device out of use. Device is transferred to a storage location.
| |
| Loss of control results.
| |
| III. Owner takes device out of service and improperly transfers device Loss of to an unauthorized control results. user (or some other individual).
| |
| IV. Device is discarded in the trash.
| |
| V. Device is discarded in the environment. "
| |
| VI. Device is sold to a salvage dealer.
| |
| TABLE 3. FINAL STATUS OF DEVICE (SCENARIO ENDINGS) i I.
| |
| Device remains in place out not in use. i No control of device. !
| |
| II.
| |
| Device remains in uncontrolled storage on-site.
| |
| III.
| |
| Possession by unauthorized individual. Potential for improper use. No control of device.
| |
| IV.
| |
| Device remains in uncontrolled storage at salvage yard.
| |
| V.
| |
| Device reprocessed; incorporated into consumer products.
| |
| VI.
| |
| Device reprocessed; incorporated into construction materials. 1 t
| |
| VII. Device buried in sanitary landfill.
| |
| VIII. Device remains uncovered in the environment. l 4
| |
| April 10, 1987 12
| |
| | |
| TABLE 4 KNOWN KUMBER OF CASES OF MISHANDLED DEVICES Estimated Number Class
| |
| * Number of Known Distributed Incidents **
| |
| Percentage A-1 20,000 l 1
| |
| 5.0 x 10 3 A-2.A-3 170,000 21 1.2 x 10-2 i B
| |
| 4,200 \
| |
| 11 2.6 x 10-1 l C 1,C-2 8,000 l 1
| |
| 1.2 x 10-2 l D 8,000 !
| |
| 1 1.2 x 10-2' )
| |
| E-1 E-2 720 0 \
| |
| 0 t F --.
| |
| 2 ---
| |
| C-1 180,000 4
| |
| 2.2 x 10-3 G-2 90,000 1
| |
| 1.1 x 10-3 H 7,000 i 1
| |
| 1.4 x 10-2 I 1
| |
| 2,000 !
| |
| 0 0 J ,,,
| |
| K ---
| |
| * See table 1
| |
| ** Sources:
| |
| : 1) Incident Summaries. Texas Dept. of Health, 1985-86,
| |
| : 2) The U.S. Nuclear Regulatory Commission and the Agreement States.
| |
| 1983. Licensing Statistics and other data,
| |
| : 3) Agreement State Incidents Involving Generally Licensed Gautes.
| |
| 1985. Licensing Statistics and other data, 1984, April 10, 1987 13
| |
| | |
| . . - . - .. . . _ . - _ _ . . _ - . . . _ . . . _ . . - . . _ . . . ~ . . . . ~ . - . . ~ . .- -. .
| |
| l l
| |
| TABLE 5.
| |
| RADIONUCLIDE RELEASE DURING INCINERATION IN VARIOUS STUDIES Nuclide Form Fraction Released Re ference H-3 Animal carcasses 1.0 Classic 1985 C-14 Animal carcasses 0.992 Classic 1985 C-14 Planchets 0.96,0.98 Bush'and Hundal 1973 Co-57 Animal carcasses -
| |
| microspheres 0.038 Brekke 1985 t
| |
| Co 57 Animal carcasses 0.273 Classic 1985 Co 60 Filter paper 0.009,0.006 Bush and Hundal 1973 Rb 86 Filter paper 0.449,0 Bush and Hund 1 1973 Ru-103 Animal carcasses -
| |
| microspheres 0.165 Brekke 1985 '
| |
| I-125 Animal carcasses 1.0 Classic 1985 1-125 Filter paper and 0.056-0.9998 Bush and animal carcasses Hundal 1973 Cs-137 Syringes 0.525,0.18 Bush and Hundal 1973 T1-204 Planchet 0.948,0.993,0.953 Bush and Hundal 1973 April 10, 1987 14
| |
| : 3) Romults i
| |
| 3.1) CLASS A PORTABLE STATIC ELIMINATORS !
| |
| 3.1.1) Device Description This aatis tatic class of static eliminators includes small, hand-held film, or brushes artists' used for reducing static from records, photographic joined between twocanvases. They consist of a small plastic handle the front. aluminum braces with a sof t bristle attachment o metal Po 210 strip located inside of the handle and behind es.
| |
| the activities(an up alpha to emitter) is the most commonly used radionuclide , with half life (138 0.50 mci days), the(18.5 MBq) when sold. Because of the short
| |
| )
| |
| i (about one with the year) is about 0.080 mci (3.0 MBq). activity at the end of the{
| |
| Instructions included )
| |
| but distributors indicate that this is rarely done (NURE .
| |
| may Static be eliminators exposed internally are themselves consumer products and the publi discarded in a landfill or a salvage or externally yard. before the eliminator is ;
| |
| assumed to External exposure is often be negligible with internal exposure via ingestion, inhalation, they are aor absorption being a greater hazard (NUREG 1980) .
| |
| Because the portable consumer product, it is difficult to account for ;
| |
| approximately 7,000 distributed per year. However, distributors interviewed claim that users frequently discard the used devices in the trash. )
| |
| 3.1.2) Scenario Development ,
| |
| If most of these they will devices are thrown away, the probability that in the final be buried in a sanitary landfill is the highest probability High conditions of the scenario (0.63), as seen in Figure 2.
| |
| controlprobabilities are found for the device remaining in place without (0.1). (0.1) and the device remaining in uncontrolled storage on site devices Distributors and users of the devices may continue to keep the These after their usefulness has expired instead of discarding them portable static eliminators may also fall into the possession of unauthorized from individualindividuals (0.078) because they can be transferred easily in the environment (0.08).to individual. The device may also remain unrecovered static eliminators contain Other outcomes seem less likely. Portable little salvagable material, and would not generally be sent through metals recycling. This accounts for the salvage dealers would not save them.salvage yard (0.001) since lower probabilities in storage at the devices will It is also unlikely that the products (0.003).be used in construction materials (0.007) or in consumer April 13, 1987 15
| |
| | |
| + -
| |
| _. ; *_<*4 m._u__ 6- . _ m _
| |
| =
| |
| I CLASS A-1 i
| |
| PORTABLE STATIC ELIMENATORS themeseanne at hpense huemee i
| |
| i home
| |
| [ h Oswese sema.m yemas.enome Es senees l
| |
| Et h eserage Oswtesm g, ,,,,,,,aa,hn
| |
| , , , g, meeragesu>emo s l
| |
| t Devue .anetweed 8M es unausnetsy m
| |
| U""*"*
| |
| * =* m8 e eis" 08 py,emo Gerareed O#I gestessecoumed
| |
| * as waam by masener &I g'gg E.bvhea seerageremake as Eget hameemus ~
| |
| 'N P''d aos desear
| |
| -' '" ~et, esse
| |
| ~~'
| |
| t
| |
| := = ..
| |
| / s .. .. . .
| |
| ES - ..,
| |
| esaurents r
| |
| m e,r
| |
| ; x namens Oswenosassehe h g.gg Figure 2 Scenario development fordevice class A-1, portable shnic eliminators
| |
| | |
| 3.1.3) Dosn Ass 2ssm:nt l 3.1.3.1) ! Intact' Source External dose -
| |
| maximum individual ssurcesTable 6 lists the estimated dose equivalents for exposure to point !
| |
| containing the maxic.um amounts of activity from the NRC Registry exposure for each device class and radionuclide. ;
| |
| to The estimate for I is given fora the 0.5 mci (18.5 MBq) gamma emissions.
| |
| Po-210 source at 100 cm for 20 weeks magnitude as natural The value is of the same order of j energy sufficient background. Po-210 emits no beta particles with to penetrate the skin dead layer. j External dose - realistic case Because the 10-6 rem, maximum external dose equivalent was on the order of than this value, the dose equivalent for a more realistic case will be lower and will not represent a significant harard.
| |
| Internal dose - maximum ipdividual l ICRP 30 assips class V oxides, hydroxides, and nitrates of polonium to an ft of 0.1. and all others to class D, with all compounds assumed to have models for Po-210 Table if 7 lists dose equivalents predicted by the ICRP assumption all of the device activity were.taken in. The because the that 100 percent of the activity might be taken in was used public, devices are routinely used by members of the general the and children or others who tamper with the devices could cause used release in of all of the activity. Even the small amounts of activity , i i '
| |
| spleen andthese lungssources can result in very high dose equivalents to the sources, if inhaled or ingested (table 7). However, these much like smoke detectors, only constitute a significant risk if broken for intake. open in such a way that the radioactive material is availabl e In al. 1975)a study of the hazard from these devices, Webb et al. (Webb et activity would found that under r.:.rmal handling, very small amounts of the be released from the microspheres, and that the dose equivalents to major organs would table 7. be much lower than the values in The microspheres were not susceptible to leaching and were assumed to pass directly through the gastrointestinal system. However, up to 50 percent of the activity could be removed from the microspheres if they hadmore probably been subjected to impact or heating. Therefore, there is sources after risk to trash handlers and others who might contact the they are thrown away. However, because of the short half-life of Po-210, less activity will remain associated with the devices at the time these people could contact them.
| |
| Internal dose - realistic case Wilmot does of 10-6 was assigned. not assign a release fraction for Po-210, so a value inhalation of 5 x 10-7 mci Table 8 lists dose equivalents for ingestion or (18.5 Bq).
| |
| April 13, 1987 17
| |
| | |
| l 3.1.3.2) Dispersed source 3.1.3.2.1) Incineration i
| |
| BecauAe these devices are routinely thrown into the trash, the l number of devices device types. incinerated per year will be higher than for other year (7000) and from the estimate in Figure 2 that 6.4 up at year isthe450._ incinerator, Using the estimated number of devices incinerated per intake by the maximumthis value in the model of Buckley, the activity Bq/ year). Table individual is 5.2 x 10 10 mci / year (0.019 individual, average9 lists the dose equivalents for the maximum individual, and population .
| |
| incinerators (22 x 106 residing near persons) (NUREG 1980) based on this conclusion.
| |
| 3.1.3.2.2) Metals Recycling !
| |
| i the As stated in section 2.3.2.2, a general result for Co-60 gauges is discussion only result afforded of this result. by the available data. See that section for a 3.1.3.2.3) Burial in Landfill Based on the i withdrawal model of Buckley, the time for Po-210 to reach the it reaches that point. point is so long that all of the activity will decay before !
| |
| 1 1 i.
| |
| l l
| |
| 1 I
| |
| l
| |
| (
| |
| l April 13, 1987 1 13 1
| |
| | |
| TABLE 6. !
| |
| ESTIMATED MAXIMUM DOSE EQUIVALENTS i FROM EXTERNAL EXPOSURE TO INTACT SOURCES -
| |
| i Device Estimated Dose Equivalent (rern) k Class Nuclide Contact Contact L' hole B.dy*
| |
| Er. .psulated** Non-Encapsulated **
| |
| A-1 Po-210 8.8 x 10 6 .
| |
| A-2 Po-210 !
| |
| 1.8 x 10-3 -
| |
| Am-241 5.3 x 10 4
| |
| )
| |
| Ra-226 1.4 x 10-3 1,7 x 10-1 1.4 x 10-2 ,
| |
| 3,9 x gol -
| |
| A3 Kr-85 1.05 x 10-2 .
| |
| H-3 -
| |
| i B Co-60 4.6 x 101 5.4 x 103 Cs-137 2.6 x 101 1.25 x 105 Ra-226 2.7 x 103 4.05 x 105 2.7 x 101 3.4 x 103 7.8 x 105 Am-241 2.1 x 101 -
| |
| 5.6 x 102 ;
| |
| C-1 Sr-90 -
| |
| l T1-204 3.7 x 10-4 1.2 x 103 Ru-106 -
| |
| 2.0 x 103 '
| |
| ?.0 x 102 Pm-147 4.5 x 10-7 -
| |
| 3.4 x 102 1 C-14 - l Pb-210 -
| |
| 1.8 x 102 <
| |
| 2.4 x 102 C-2 Sr-90 -
| |
| 9.4 x l'05 D N1-63 -
| |
| H-3 -
| |
| E-1 Am-241 3.2 x 10 1' -
| |
| 8.4 x 102 cm-244 2.2 x 101 -
| |
| E-2 Cd-109 1.2 x 101 -
| |
| Fe-55 2.6 x 101 -
| |
| 1 F Ra-226 1.1 x 10-2 1,4 x 100 3.2 x 102 l
| |
| Cs-137 1.3 x 10-1 1.4 x 101 2.0 x 103 Co-60 4.6 x 10-2 5.4 x 100 Sr-90 -
| |
| 1.25 x 102 4.7 x 101 G-1 H-3 - -
| |
| Kr-85 9.0 x 100 .
| |
| C-14 -
| |
| l 3.7 x 102 April 13, 1987 19
| |
| | |
| TABLE 6.
| |
| ESTIMATED MAX 1 MUM DOSE EQUTVALENTS_ .
| |
| FROM EXTERNAL EXPOSURE TO INTACT SOURCES - COi1
| |
| ** i
| |
| '\
| |
| Device Estimated Dose Eauivalent (rem) '
| |
| Class Nuclide
| |
| | |
| ==Contact:==
| |
| Contact Whole Body * !
| |
| Eticapsulated** Non-Encapsulated **
| |
| j G-2 H-3 -
| |
| - 1 Pm-147 2.7 x 10-3 2.05 x 106 H Cs-137 5.1 x 10 2 5.4 x 101 8.2 x 102 Ni-63 -
| |
| I Am-241 .1.06 x 10-2 -
| |
| 1.4 x 10*1 J U-238 -
| |
| Th-232 -
| |
| K. Pu-239 !'
| |
| l
| |
| *-20 week exposure at 100 cm.
| |
| ** 3 hour contact time. 1 I
| |
| i
| |
| 'I l
| |
| i l
| |
| I i
| |
| I April 13, 1987 20
| |
| | |
| IABLE 7. COMMRTTED DOSE EQUIVALENTS FOR IN j OF 5 x 10'1 mci (18.5 MBq) Po-210 Dose Equivalent (rem)
| |
| Organ _ Inhalation Ingestion Class D Class W I Kidneys Liver 4.6 x 103 2.2 x 104 8.1 x 102 7.2 x 103 Spleen 4,1 x to3 .
| |
| Lungs 8.1 x 103 4.1 x 104 1.2 x 104 2.4 x 104 TABLE 8. COMMITTED DOSE EQUIVALENTS FOR INGES OF 5 x 10-7 mci (18.5 Bq) Po-210 Dose Equivalent (rem)
| |
| Organ Ingestion Inhalation Class D Clasis W Kidneys 4,6 x 10*3 Liver 2.2 x 10-2 7.2 x 10-3 Spleen 8.1 x 10-4 4.1 x 10 -
| |
| 8.1 x 10*3 4.1 x 10*2 !
| |
| Lungs -
| |
| 1.2 x 10-2 2.4 x 10-2 TABLE 9.
| |
| COMMITTED DOSE EQUIVALENTS FOR INHA1ATION OF 5.2 x 10-10 mci (0.019 Bq) Po-210 FROM INCINERATOR EMISSIONS Maximum Average Population Organ Individual (rem) Individual (rem)
| |
| Class D Class W Class D (Person-rea)
| |
| Class W Class D Class W Kidneys Liver 2.3 xx 10-6 4.2 10-5 7.5 x 10-6 1.2 x 10-5 3.8 x 10-6 2.5 x 102 8.2 x 101 Spleen 2.1 x 10-6 -
| |
| 4.6 x 101 Lungs 4.2 x 10-5 t,3 x to-5 2.1 x 10-5 6.4 x 10-6 -
| |
| 4.6 x 102 1,4 x to2 2.5 x 10-5 -
| |
| 1.2 x 10-5 -
| |
| 2.7 x 102 April 10, 1987 21
| |
| | |
| 3.2)
| |
| CLASS A 2 - STATIC ELIMINATORS OR DETECTORS: IN EQUIPMENT OR PROCESS LINE (VERY HIGH T0XICITY SOURCES) 3.2.1) Device Description The io charge ' 'nization build-up sourcesonin these devices are in equipmenr. used move staticto re roller systems) or in air ducts. production lines (conveyor belts, sense and measure Detectors in this class are used to line or be portable. static charge. They may be attached to the process Commonly used activities up to 100 mci (3700 MBa), nuclides include Po-210 with (0.0185 MBq), Am-241 with activity of 0.0005 mci and Ra 226 with activity of 0.0005 mci (0.0185 MBq).
| |
| This class the CFR as opposed of static to eliminator is covered under section 31.5 of section 31.3. Industrialthe portable static eliminator covered under required to perform leak static eliminator users are therefore maintenance restrictions. tests, and observe transfer, disposal, and products Most of the these devices differ from other devices rather than selling them. covered by section 31.5 because months to a year. The lease period is usually six The major distributor of the davices revealed that approximately to the client. 95 percent of the devices are returned without a reminder the be warning charged. that if the device is not returned,s fee another will yea They report good results using this method.
| |
| In a 1984 static eliminators to determine their awareness of license re Eight of regulations the twelve (67 percent) licensees surveyed were aware of the and 80 percent gj records (NRC 1987). were keeping proper receipt and transfer +
| |
| general license requirements Even with a seemingly greater awareness of the for filed concerning missing this class of devices, reports are static eliminators or incidences of damage.
| |
| Four of the licensees contacted in the NRC survey reported misplacin total of ten static eliminators (NRC 87). Fixed static eliminators also line and have a problem may encounter similar a to the gauges in that they are on a process disappear),
| |
| have left theor the devices may be lost because knowledgeable plant. pe being lost. Plant closings may also contribute to the devices 3.2.2) Scenario Development Due to plant closings, the devices may remain in place without .
| |
| control (0.2) or be placed in storage (0,1) (Figure 3). Probabilities environment, assigned to the devices being discarded to the trash or the of 0.3 were individual with 0.15 each going to recovery of the device by another the and environment. to either the trash handler or remaining unrecovered in ProbaL.tities transferred are lower for the device being dealer (0.05).to an unauthorized individual (0.05) or sold to a salvage Final probabilities possession of are highest for the device remaining in an unauthorized individual (0.23). This is a result of April 10, 1987 22
| |
| | |
| ,_.._.m..m._m. ..m.- _ _ . . _ - . . . . ... m -_. _m._. .m.. . .-.. _. . .. ..m- .m. .. ..m..- . _ . . .. -. .
| |
| t CLASS A-2 EQUIPMENTSTATIC ELIMINATORS
| |
| %% 0.2 h pense but not me
| |
| ; Dev
| |
| [ h enee. sen.emie,
| |
| ,s. ., i eentred 6 r
| |
| El h e sse tiewemerunehe he amusenesend gg masrego adame f
| |
| 0.95 !
| |
| Device tenetersed 4.eenedhartend t he,teuss Dowdne rumette h poemessean at [
| |
| e.23e annuenesteeg ES i 0.1 mw encareas als Desenesomoussed Dewho sumah.
| |
| seeresh ty e., o.s N ensninos e.3
| |
| ~ "8'ase*yere s.ss -
| |
| i a.se seaween desear ~
| |
| ansane.
| |
| 8.13 Devemo encarded
| |
| ~
| |
| S.13 humuAur ggs
| |
| &l O.45 SMS f _
| |
| 84 g,g y eA27 i
| |
| S.SSS Destes sett to amepage emeter Devise heated h eenhary 9.130 '
| |
| gangggg Devene sumiseen h O'II
| |
| .n e ,
| |
| Floure 3 Scenario &;@..iid for device class A-2, equipment static eliminators
| |
| | |
| l l
| |
| 1 tha various pathways loeding to this condition.
| |
| remaining in place without control (0.20), The probabilities of landfill (0.16), and remaining unrecovered in the environmentbeing b are also high. (0.15) 3.2.3) Dos'e' Assessment 4
| |
| 3.2.3.1) IIntact' source 1 External dose - maximum individual Table 6 shows the activity levels dose equivalents for exposure to point sources at stated Ra-226 for 20 weeks at in section 3.2.1 for Po-210, Am-241, and millirem 100 cm. The values are all on the order of 1 with energy (0.01 mSv). As stated in section 3.1.3.1, Po-210 has no betas sufficient equivalents to the to penetrate the dead layer of the skin. Dose for -unencapsulated skin at a depth of 0.007 cm are listed in cable 6 sources of Am 241. and for encapsulated and unencapsulated sources of Ra-226, assuming a three hour exposure in all cases.
| |
| transitoryNone of the values listed are sufficient to produce even erychema.
| |
| External dose - realistic case around Because the maximum external dose equivalents for photons were not be 1significant.
| |
| millirem, the realistic estimates will be much-lower and will Ra-226 were 14 mrem The maximum beta dose equivalents for Am-241 and Realistic (0.014 mSv) and 39 rem (0.39 Sv), respectively.
| |
| will not beestimates significant.for thcae nuclides will be lower; those for An-241 for radiation workers Those (18.75 for resRa-226 will be lower than the limits '
| |
| negligible if the source activity is not exposed.per calendar quarter), or will be Internal dose - maximum individual Although it would take a significant effort, damaged in such a way as to make the material available for uptake irtothe the body by inhalation or ingestion. The maximum dose equivalent possible would result from intake pathway. of all of the material by either so the that the activity was scattered, it is not likely that even radioactivity would be inadvertently swallowed or inhaled.
| |
| Material suspended in air may be directly inhaled, resulting in activity deposition in the lungs and gastrointestinal tract. Scattered material could be spread over the enough to produce a dose to the skin which might result in some d The most likely pathway for ingestion would be from activity on the skin transferred to the mouth.
| |
| equivalents which would resultTables 10-12 list from intake of the 0.3committed of the dose source activity. The fraction of amount which might be inadvertently taken in.0.3 was arbitrarily chosen as a ma April 13, 1987 24
| |
| | |
| . . - . - .-- -..-. . __ . . _ . . . . -. . .- -.- . . . - ~ . _ - - -
| |
| 1
| |
| \
| |
| 1 Interns 1 doan - realistic cesg '
| |
| Based on assigned to allthe conclusions of Wilmot, uptake fractions of 10-6 are list the associated dose equivalents for intake of 10-6three radi ;
| |
| of the maximum source ' activities of Po-210, Am-241, or Ra-226, respectively. l 3.2.3.2) Dispersed Source
| |
| \
| |
| 3.2.3.2.1) Incineration ,
| |
| Based on the maximum the model of Buckley, the amount of activity inhaled by (1.3 x 10-5 individual near an incinerator would be 3.47 x 10 7 mci i
| |
| i either. Am-241 MBq) of Po-210,'and 1.7 x 10 12 mci (6.4 x 10'11 MBq) of 1 equivalents fororintake of Tables Ra 226. these 16-18 show the associated dose )
| |
| j average individual, and population. levels for the maximum individual, 3.2.3.2.2) Metals Recycling g 4
| |
| the only result afforded by the available data.As stated in sectio discussion of this result. See that section for a j
| |
| 3.2.3.2.3) Burial in Landfill landfillThe to model of Buckley predicts that the transit time from the l
| |
| that For all of the activity decays before it reaches the withdraw Ra-226, the model predicts an uptake of 1.7 x 10-6 .
| |
| # MBq); mci (6.2 x 10-5
| |
| < 19. dose equivalents corresponding to this uptake are shown in table d
| |
| 4 i
| |
| i i
| |
| 4 i
| |
| 4 4
| |
| April 13, 1987 4
| |
| 25 e
| |
| E
| |
| | |
| I I 1QLE 10.
| |
| COMMITTED DOSE EQUIVALENTS FOR INCESTION 4
| |
| {
| |
| i OF 30 mci (1.11 x 109 B0) 0F Po-210 j
| |
| Organ Dose Equivalent (rem) i Ingestion t
| |
| Inhalation i Class D Class W Kidneys Liver 2.8 x 105 -1.3 x 106 4.3 x 105 Spleen 4.9 x 104 2.4 x 105 .
| |
| Lungs 4.9 x 105 2.4 x 106 7,4 x 105
| |
| -: 1.4 x 106 TABLE 11.
| |
| COMMITTED DOSE EQUIVALENTS FOR INCESTION OR INH OF 1.5 x 10-4 mci (5.6 x 103 Ba) 0F An-241
| |
| . Organ Dose Eaulvalent (rem)
| |
| Ingestion Inhalation Conads Red Marrow 7.7 x 10-2 1.8 x 101 Bone Surfaces 4.7 x 10-1 1.1 x 102 Liver 6.1 x 100 .1,4 x 1.03 1.3 x 100 3,1 x 102 TABLE 12.
| |
| - COMMITTED DOSE EQUIVALENTS FOR INCESTION OR INHA 0F 1.5 x 10-4 mci (5.6 x 103 B0) 0F Ra-226 l-Organ Dose Equivalent (rea)
| |
| Ingestion Inhalation Donads Red Marrow 5.1 x 10-2 .
| |
| 3.3 x 10-1 -
| |
| Bone Surfaces 3.8 x 100 Lungs 4.2 x 100 8.8 x 100 April 10, 1987 26
| |
| | |
| I l
| |
| TABLE 13. COMMYTTED DOSE EQUIVALENTS FOR ING OF 10-4 mci (3700 Bq) Po-210 Dose Ec.. valent (rem)
| |
| Organ Ingestion Inhalation Class D Class W Kidneys Liver 0.92 4.4 1.4 Spleen 0.16 0.81 -
| |
| 1.6 8.1 Lungs -
| |
| 2.5 4.8 TABLE 14. COMMITTED DOSE EQUIVALENTS FOR INGEST OF 5 x'10-10 mci (0.0185 Bq)~Am-241 Organ Dose Equivalent (rem)
| |
| Ingestion Inhalation Gonads Red Marrow 2.6 x 10-7 5.9 x 10-5 1.6 x 10-6 3,7 x 10-4 Bone Surfaces 2.0 x 10-5 Liver 4.6 x 10-3 4.3 x 10 1,o x to-3
| |
| )
| |
| TABLE 15. COMMITTED DOSE EQUIVALENTS FOR INGESTIO OF 5 x 10-10 mci (0.0185 50) Ra-226 Organ Dose Equivalent (rem)
| |
| Ingestion Inhalation Gonads 1.7 x,1.0-7 Red Marrow 1.1 x 10-6 Bone Surfaces 1.3 x 10-5 Lungs -
| |
| 1.4 x 10-5 3.0 x 10-5 i
| |
| l April 10, 1987 27 i
| |
| )
| |
| | |
| TABLE 16.
| |
| l COMMITTED DOSE EQUIVALENTS FOR INHA1ATION OF Po-210 FROM INCINERATOR EMISSIONS Maximum Average Populatiot. -
| |
| Organ Individual (rem) Individual (rem)
| |
| Class D Class W C1a=s D (Person-rem)
| |
| Class W Class D Class W Kidneys 1 5 x 10-2 2.8 x 10-3 5.0 x 10~3 1.4 7,7 x 10-3 2.5 x 10-3 5.6 x 102 1.8 x 102 Liver Spleen x 10-3 2.8 x 10-2 8.6 x 10-3 1.4 x 10-2 4,3 x 10-3 1.0 x 102 Lungs -
| |
| 1.7 x 10-2 1.0 x 103 3.1 x 102 8.4 x 10*3 -
| |
| 6.1 x 102 TABLE 17.
| |
| COMMITTED DOSE EQUIVALENTS FOR INHALATION I 0F Am-241 FROM INCINERATOR EMISSIONS i I
| |
| Maximum- Average Organ Population Individual (rem) Individual (rem) (Person-rem)
| |
| Gonads 2.1 x 10-7 1,o x to-7 7.5 x 10-3 Red Marrow 1.3 x 10-6 Bone Surfaces 1.6 x 10-5 6.4 x 10 ' 4.7 x 10*2 Liver 8.0 x 10-6 5.9 x 10-1 3.5 x 10-6 1.8 x 10-6 1,3 x to-1 l l 1 \
| |
| TABLE 18. 1 COMMITTEC DOSE EQUIVALENTS FOR INHAIATION l OF Ra-226 FROM INCINERATOR EMISSIONS Maximum Average Organ Population Individual (rem) Individual (rem) (Person-rem)
| |
| Lungs 1.0 x 10-7 I 5.1 x 10-8 3.8 x 10-3 I Bone Surfaces 4.9 x 10-8 2.4 x 10-8 1.8 x 10-3 i :
| |
| i l
| |
| 1 I
| |
| April 13, 1987 28 l
| |
| | |
| TABLE 19.
| |
| C0KMITTED DOSE EQUIVALENTS FOR INGESTION j 1.7 x 10-6 mci (6.2 x 10-5 MBq) of Ra-226 0"rg'an i Dose Equivalent (rem)
| |
| Gonads Red Marrow 5.7 x 10-4 Bcne Surfaces 3.8 x 10-3 i 4.3 x 10-2 i
| |
| i s
| |
| \
| |
| l
| |
| 'l O
| |
| April 13, 1987 29
| |
| | |
| 3.3) CLASS A STATIC ELIMINATORS OR DETECTORS: IN EQUIPMENT OR PROCESS LfNE (LOW T0XICITY SOURCES) 3.3.1) Device Description This class of static eliminators is also used by industry in a variety of ways.
| |
| paint and ink jet They remove static build-up in areas such as spray other areas. nozzles, plastic laminates, printing presses, and They are similarDetectors are used to sense and measure static charge.
| |
| except in every way to the A-2 class of static eliminators and Kr-85typical nuclides are H-3 of with an activity with2anmciactivity of 250 mci (9250 MBq) distribution of these devices was not readily (74 MBq).
| |
| A-2 class, available. Estimates for As with the covered by these static eliminators may be attached.or portable and are the the same section of the CFR. They also are covered under same leasing arrangements as the previous class of static eliminator.
| |
| 3.3.2) Scenario Development ,
| |
| The same probabilities were assigned to this class of static eliminators as to the A-2 class of static eliminators (Figure 4).
| |
| 3.3.3) Dose Assessment 3.3.3.1) ' Intact' Source External dose - maximum individual Table 6 shows dose equivalents for exposure to point sources at the activity 100 cm. levels stated in 3.3.1 for Kr-85 and H-3 for 20 weeks at H-3 presents no external hazard; the estimate for Kr-85 under these assumptions is around 10 mrem (0.1 mSv) from the photons. If the Kr 85 source were exposed, the activity would not pose a risk to the skin as a point into the nearby environment. source beta emitter, because the gas would disperse External dose - realistic case Because the maximum external dose equivalent was around 10 mrem, the realistic estimates will be much lower and will not be significant .
| |
| Internal dose - maximum individual Tritiated body water, water taken into the body quickly equilibrates with the body and and is assumed to be distributed uniformly throughout the As for the retained with a biological half time of 10 days (ICRP 1978).
| |
| arbitrarily chosen high toxicity static eliminators, a fraction of 0.3 was as the maximum amount which might 'nadvertently be taken in. The dose equivalent for intake of 75 mci (2800 MBq) of H-3 as tritiated body. water is 4.7 rem (0.047 Sv) to the soft tissue of the whole to Because H-3 in the form of gaseous hydrogen is usually converted tritiated water in the atmosphere (Jacobs 1968), dosimetry for H-3 as gaseous hydrogen is not considered in this report.
| |
| 3 I April 10, 1987 30 i
| |
| | |
| CLASS A-3 STATIC ELIMINATORS / DETECTORS ot. _
| |
| ==
| |
| h MW hN w 0.2 eentral
| |
| .1 newesepimend h senses De.t.e %
| |
| b ** n**'bd "8 eterage se one Device tenetereed 0.05 ts unauthertmed
| |
| ~ op. tan,esasha b .- . ef 0.228 W
| |
| %m e.t s
| |
| ':"l".,"l"':l- .
| |
| e.s e.t s U >
| |
| o.se W*****r es. ten.
| |
| ..m """'r * *2 o
| |
| ::=
| |
| 0.45
| |
| ~~
| |
| f .>
| |
| Canseuceu e.s '"*"''''
| |
| o e.w e.s he o ,t,.
| |
| h monitary gg33 pertes sumisme
| |
| ^
| |
| _ _ kn 0.15 anutsunant Figure 4 Scenario development for device class A-3, static eliminators / detectors
| |
| | |
| I l
| |
| Tha noble gases ovsrriding concern in the dosimetry of immersion in clouds of 1978). The is the dose to the organs from external irradiation (ICRP ,
| |
| mci (74 MBq) of Kr-85 in a 100 3 mdoses to the major organs from immersio equivalent rates are room are listed table 20. These dose hazard exists from mishandling of these devices.also extremely internal low, and Internal dose - realistic individual i The values given cannot be directly applied by Wilmot for release of various radionuclides to tritium 10*l. Using 10-1 with a value of 10~3 . The largest value he gives is for uptake, the fraction of the ori ral source activity which would be taken in under this model is
| |
| .10 -
| |
| equivalentFor intake to theof soft 0.025 mCL (0.925 MBq) of tritiated water, the dose rem (1.6 x 10-5 tissues of the whole body would be 1.6 x 10-3 low for the maximum Sv). Dose case,equivalents for exposure to Kr-85 were very so realistic significance.
| |
| estimates would not be of 3.3.3.2) Dispersed Source 3.3.3.2.1) Incineration (6.4 The model of Buckley et al predicts an uptake of 1.7 x 10-6 mCL/yr x 10-5 250 mci device. (9250 MBq)
| |
| This would result of H-3, if 100% of the activity is relea (1.1 x 10-9 in a dose equivalent of 1.1 x 10-7 rem /yr 10-10 Sv/yr) to the maximum individual, 5.5 x 10-8 (4.0 x Sv/yrg 10- to the average individual, and 4.0 x 10-3 rem /yr (5.5 x person-rem /yr Kr-85, 100% person-Sv/yr) to tha population near the incinerator. For ;
| |
| activity from release from the device was assumed and no removal of the in a concentration the stack gases was assumed. These assumptions resulted from the incinerator.of 1.9 maximum x 10~11 mci /m3 (7.0 x 10~4 Bq/m3 ) downwind Corresponding annual dose equivalents to the to this concentration are listed in table 21.and average individual an 3.3.3.2.2) Metals Recycling the only result afforded by the available data.As stated in section 2 discussion of this result. See that section for a 3.3.3.2.3) Burial in Landfill landfill Thepredicts model ofanBuckley intake for transport of 0.866 mciof radionuclides from a withdrawal point (32 MBq) of H-3 at the was buried. Intakenear a landfill in which a 250 mci (9250 MBq) source equivalent of 0.054 of this amount of activity would result in a dose rem (5.4 x 10-4 whole body. No groundwater pathway for Kr-85 was envisioned.Sv) to the soft tis April 10, 1987 32
| |
| | |
| i TABLE 20.
| |
| DOSE EQUIVALENT RATES FOR IMMERSION IN '
| |
| 2 mci (7.4 x 107 Bq) 0F Kr-85 IN A 100 M3 RO0t!
| |
| Organ Dose Equivalent (rem /hr)
| |
| Conads Breast 8.9 x 10-7 Red Marrow 8.1 x 10-7
| |
| ' Lungs 9.6 x 10~7 Bone Surfaces 7.1 x 10*7 I Stomach Vall 1.0 x 10-6 Kidneys 7.2 x 10~7 Liver 6.7 x 10*7 Spleen 6.3 x 10-7 i Adrenals 7.4 x 10-7 Skin 6.6 x 10-7 Lens of the eye 3.4 x 10-3 1.5 x 10-6 TABLE 21.
| |
| ANNUAL DOSE EQUIVALENTS FOR INHALATION __
| |
| OF Kr-85 FROM INCINERATOR EMISSIONS Maximum Average Organ Population Individual (rem) Individual (rem) (Person-rem)
| |
| Gonads 3.2 x 10-10 Breas't 1.6 x 10-10 1.2 x 10-5 2.8 x 10 10 1.4 x 10-10 4 Red Marrow 3.5 x 10-10 1.0 X 10-5 4 1
| |
| Lungs 1.7 x 10-10 1.3 x 10-5 2.6 x 10-10 1.3 x 10-10
| |
| . Bone Surfaces 3.8 x 10-10 9.6 x 10-6 Stomach Vall 1.9 x 10-10 1.4 x 10-5 2.6 x 10-10 1.3 x 10-10 Kidneys 2.4 x 10-10 9.6 x 10-6 Liver 1.2 x 10-10 9.0 x 10-6 2.3 x 10-10 1.2 x 10-10 Spleen 2.8 x 10-10 1,4 x to-10 8.5 x 10-6 Adrenals 1,o x 10-5 2.4 x 10-10 1.2 x 10-10 Skin 2.9 x 10-8 8.8 x 10-6 Lens of the eye 3.6 x 10-10 1.4 x 10-8 1,1 x 10-3 .
| |
| 1.8 x 10-10 1.3 x 10 5 l
| |
| l
| |
| _'* April 10, 1987 33 l
| |
| l
| |
| | |
| 3.4) CIASS B - GAMMA GAUGES l
| |
| 3.4.1) Device Description Gamma !
| |
| a product' ion gauges contain sources which are mounted with a detector on 1 reflected line for monitoring and quality control. j the intensity rays of from thea radioactive source are sensed by a detector, andThe e t 1
| |
| thickness, density, etc.' signal may be translated into a measure of the this study to of the material studied. The limitation of popular devices sources less chan 20 mCL (740 MBq) removes some of the include Co-60, from consideration. Radionuclides commonly used Cs-137, Am-241, and Ra-226.
| |
| 50,000 industrial gauging devices It is estimated that some 85).
| |
| are probably in use today (Peters Gamma gauges are normally not '
| |
| removed from the process lines.
| |
| ; Because of the work environments commonly encountered, there is significant potential for damage of workplace. the devices and sources in the treatment, or The poor corrosive maintenance atmosphere of industrial lines, rough indicates be painted over or removed.the presence of labels These a radioactive may also source from other parts on the processing Gamma gauges are often hard to distinguish line j
| |
| (Peters 1985). Devices and vary greatly in size with them have left the are often lost because the persons familiar gauges may be discarded inplant or the plant has clos.ed. These lost 1
| |
| (Lubenau 1986). landfills or included eith scrap metal i
| |
| < i
| |
| ; Several incidents. have
| |
| ' become occurred in which steel products have In 1984,contaminated significant by radioactivity assumed to be from a gamma gauge imported from levels of Co-60 were detected in steel fittings ll of a source with Taiwan;scrapthe source feed. was traced to the inadvertent mixing In 1984,
| |
| { Carolina steel plant was struck and melted by a stream of moltea Cs-137 gaugei n steel.
| |
| No radioactive confined products to the plant site. were released, and all . contamination was ;
| |
| Cs-137 gauge mixed with scrap An event in California in 1985 involved a metal, again in i activity was kept on the plant site. which most of the
| |
| ! gauges recovered were lost by a licensee in Four Cs-137 level and density e
| |
| ]
| |
| from a nearby scrapyard and contaminated steel products we 1
| |
| detected at another location (Lubenau 1986).
| |
| 3.4.2) Scenario Development
| |
| ' Figure 5 shows the pathways for gamma gauges. probabilities assigned to the events and i
| |
| from initial events were Highest probabilities for pathways leading (because assigned to the device remaining in place to these devices are designed to be fixed on a process line) and into the device
| |
| { the trashbeing (0.3).recovered by a trash handler after being discard Lower values (0.1) were assigned to the device e
| |
| d being directly placed to theinsalvage storage dealer.
| |
| or transferred to an unauthorized individual or familiar with their use was thought The apparent value to persons not (steel and lead). to be in the recoverable metals
| |
| .; Therefore, the pathway leading from recavery by an April 10, 1987 34 l
| |
| | |
| 'a CLASS B - GAMMA GAUGES r
| |
| Devkoseenhe 83
| |
| " T,,*,e"' "" i De.ne nehe '
| |
| h pisse weheut 0.3 eentrol
| |
| [
| |
| &I Dowkepinned h eeween
| |
| < ; omen ,,,,,, 5 El i
| |
| Device tenefereed Al 2 Dewkesesenho
| |
| * h poseession of 0.136 emmelhartmed heawidual 85 0.1 Destes seewe,q 0.033 Dueles sessemed -~
| |
| Devine senehe u i, irm by anWher SA ,,,
| |
| be eterste et 0.023 Art nimete g esfvego yard 0.3 ' --
| |
| 0.080 h%
| |
| doeter km Devko escarded
| |
| /0.033 0.033 Tresh handler 0.g bacteerecer 0.S e,3e g,g3 g,, W 0.7 t 0.3 Canedruction analeriete 0.107 Dowton eett El sp eefrege Device tourled h annaery 0.240 landfie i
| |
| Dawke rseinho Nh 0.033 enetenment Figure 5 Scenario development fordevice class B, gamma gauges
| |
| | |
| . . _ _ - - . .~.. - - - .-- . . - . - _ _ - _- - - - . .
| |
| 1 I individual to i
| |
| thzt for discarded tha individual keeping the device (0.2).the salvaga .
| |
| an dI into the. trash, If the device were disposal in a the most probable event was assumed to be recovered, the landfill (0.7) if not recovered from the trash; if more heav1fy pathway leading to the salvage dealer was again weighted 4.
| |
| Because of the high metal content, the pathway from the salvage dealer to metals recycling was weighted the most heavily (0 with the materials other pathways weighted less heavily (0.1 -each). .
| |
| Of to ~
| |
| (Metal construction Statistics 1986).
| |
| materialsNo and one-third (0.33) go to con assess the probability adequate information was available to 1
| |
| incineration, so ~ probabilities regarding the flow of materials after leading from this point. of 0.5 were assigned to both pathways Predicted final .
| |
| 1andfill through (0.25),
| |
| metal recycling remaining (0.16 total).in place without control (0.30) trends !
| |
| i These results show some of the these devices if mishandled.noted by Lubenau et al. (Lubenau 1986) for final disj i j.
| |
| i 3.4.3) Dose Assessment i 3.4.3.1) f_ Intact' source well-shielded Gamma gauges typically employ radioactive source's~ enclosed in a 4
| |
| forth to expose or cover geometry with a mechanical shutter which moves back and the source. Because a built-in mechanism exposes the devices could source, a person not familiar with safe handling of these l radiation dose. conceivably expose the source and' receive a significant ii External dose - maximum individual -
| |
| i Table \
| |
| , 6 shows the estimated radiation doses for exposure to these I i
| |
| sources at 100 cm for 20 weeks or on contact for three hours.
| |
| j External dose - realistic case ,
| |
| External i containing moreexposure rates around several Cs-137 gauges with sources l e
| |
| i Source and Devicethan 20 mci (740 MBq) were quoted in the NRC's Sealed 4
| |
| Registry. ;
| |
| The devices were shielded so that the !
| |
| external contact exposure rates were less than 10 mR hr~1 (2.6 uC kg'1 hr~1) on and less than 5 mR hr-1 (1.3 uC kg*1 hr-1) at i foot (30.5 cm) from the device with the shutter closed.
| |
| the devices considered in this report,Because of the lower activity of i
| |
| shuttered device would probably not result continuous exposure to the exceeding 500 mrem /yr (5 mSv/yr). in annual dose equivalents i
| |
| ' If the shutter were briefly opened, dose the equivalents for photon exposure would be less than those quoted in -
| |
| i to theprevious section for a 20 week (3400 hour) exposure, in proportion time of exposure.
| |
| J April 13, 1987 36 J
| |
| T A
| |
| a
| |
| | |
| _ Internal dose - maximum individual For Co-60, material in ICRP the 30 (ICRP 1979) assigns an fl value (fraction of gastrointestina) bloodstream) of 0.05 to oxides and hydroxides, system which passes into the and an f to all other, compounds. 1 value of 0.3 assigned to inhalation class Oxides, Y: hydroxides, halides, and nitrates were
| |
| . )
| |
| class W.
| |
| all compounds For Cs-137, all classes were assigned an fall others were assign were assigned to inhalation class D.l value compounds were assigned.an f of 1.0 and For Ra-226, all W.
| |
| an For Am-241, all compounds were assigned an f1 value of 0.2 and an inhala inhalation l value of 5 x 10'' and of 0.3 class of W. As discussed in section 3.2.3.1, a fraction was chosen an individual might as the maximum take fraction of the source activity that equivalents resulting from this assumption.22-25 list the organ dose Tables in.
| |
| Internal dose - realistic case In most circumstances, the released. If a gauge were exposed activity toin the gauges would not be otherwise mistreated, a a corrosive atmosphere or from the small fraction of the activity might be lost list the source in the form of removable contamination. Tables 26-29 10-4 of the assumeddose committed equivalents which would result from intake of source activity for Co-60, and 10-6 of the source activity for Cs-137, Ra-226, and Am-241.
| |
| 3.4.3.2) Dispersed source 3.4.3.2.1) Incineration From table 5, stack release assigned to Co-60 and fractions of 0.273 and 0.525 were '
| |
| Buckley et al. Cs-137, respectively, for use in the model of for incineration. Fractions of 0.5 were assigned to Ra-226 and Am-241, because nuclides other than those forthis was the approximate maximum for which very high release fractions were observed (H-3, C-14, based on the model of1-125, Buckley andetT1-204).
| |
| al. (NUREGThe total activity inhaled, 1980), is 1.9 x 10-8 mci (7.0 x 10 7 MBq) of Co-60, 7.3 x 10-8 mci (2.7 x 10-6 MBq 6.9 10-6 x MBq) 10-8 mci (2.6 x 10-6 MBq)ofAm-241,and3.5x10gofCs-137, mci (1.3 x of Ra-226. Tables 30-33 list individual, and the maximum individual, average of these activity levels. population committed dose equivalents from inhalation 3.4.3.2.2) Metals Recycling As stated in section 2.3.2.2, a general result for Co 60 gauges is the only result afforded by the available data.
| |
| discussion of this result. See that section for a 3.4. 3. 2. 3) Burial in Landfill l
| |
| be The low very total activity reaching the withdrawal point was calculated to radioactive decay for Cs-137, Co-60, and Am-241 (as the equation predicted t
| |
| before reaching the withdrawal point) and 0.033 mci j April 13, 1987 37
| |
| | |
| 4 I
| |
| \
| |
| (1.2 MBq) of Ra-226. Table 34 lists committed dose equivalents for ingestion of 0.033 mci (1.2 MBq) of Ra-226.
| |
| l 1
| |
| i i
| |
| i i
| |
| i i
| |
| I l
| |
| . l April 13, 1987 1 38 1 l l
| |
| l
| |
| | |
| . - . . .. . .. . . . . - . . . - _ . - - - - . . . _ - . . - - . . . . - . .- . - . . ~ . - . - . _ . - . - -
| |
| 1 TABLE 22.
| |
| COMMRTTED DOSE EQUIVALENTS FOR INGESTION OR INHALATI
| |
| ( OF 3 mci (111 MBq) OF Co-60 ,
| |
| . i Dose Equivalent (rem) '
| |
| It.ges tion Organ f i-0.05 Inhalation fi -0.3 Class W Class Y i Conads 36 f 80 44 Breast 12 -
| |
| 57 47 Red Marrow 14 61 47
| |
| ' Lungs 9.7 56 400 Small Intestine 40 3800 91 Upper Large Intestine 63 110 Lower Large Intestine 120 160 Liver 91 -
| |
| 26 140 100 Remainder of body 23 97 89 -
| |
| TABLE 23. COMMITTED DOSE EQUIVALENTS FOR INCESTION OR INRALATI OF 6 mCL (222 MBq) OF Cs-137 i
| |
| Organ Dose Equivalent (rem)
| |
| Ingestion Inhalation Gonads 310 Breast 200 270 170, -l' Red Marrow 290 Lungs 180 290 200 Thyroid 290 180 Bone Surfaces 290 180-Small Intestine 310 200 Upper Large Intestine ~310- 200 Lower Large Intestine 310 200 Remainder of body 330 210 April 10, 1987 39
| |
| | |
| TABLE 24. COMM7TTED DOSE EQUIVALENTS FOR INCESTION OR INHALATION OF 3 mci (111 MBq) OF Ra-226 Dose Equivalent (rem)
| |
| Organ Ingestion Inhalation Conads 1000 -
| |
| Red Marrow 6700 -
| |
| Bone Surfaces 75000 84000 Lungs -
| |
| 180000 TABLE 25.
| |
| COMMITTED DOSE EQUIVALENTS FOR INCESTION OR INHALATION OF 6 mci (222 MBq) of Am-241 Dose Equivalent (rem)
| |
| Organ Ingestion Inhalation Conads 3100 710000 Red Marrow 19000 4400000 Bone Surfaces 240000 56000000 ;
| |
| s Liver 51000 12000000 l
| |
| 1 April 10, 1987 40
| |
| | |
| l TABLE 26.~ COMMITTED DGSE EQUIVALENTS FOR INGESTION __
| |
| 1 OF 1 uCi (37 kBq) 0F Co-60
| |
| . Dose-Equivalent (mrem) l Ingestion '
| |
| . Organ f i-0.05 Inhalation fi -0.3 Class W. Class Y Gonads 12 27 Breast 4.1 15 -
| |
| ! 19 16
| |
| ' . Red Marrow -4.8' -
| |
| Lungs 20 16 -
| |
| 3.2 18 130 Small Intestine 13 30 1300
| |
| ' Upper Large Intestine- 21 36 Lower Large Intestine- 41 52 Liver :30 -
| |
| 8.5 48 34 Remainder of body 7.8 32 30 -
| |
| l TABLE 27. COMMITTED DOSE EQUIVALENTS FOR INGESTION OR INH OF 20 nCi (740 Bq) 0F Cs-137-Organ Dose Equivalent (mrem) l Ingestion Inhalation .
| |
| , i Gonads 1.0 0.65 Breast 0.89 i l
| |
| Red Marrow 0.58 ''
| |
| 0.96 0.61 Lungs I 0.96 0.65 Thyroid. 0.96 l Bone Surfaces -0.58 ,
| |
| 0.96 0.58' !
| |
| Small Intestine 1.0 Upper Large Intestine 0.67 .
| |
| ' 1. 0 . 0.67 Lower.Large Intestine 1.0 Remainder. of body 0.67 i 1.1 0.70 i
| |
| i l
| |
| 1 I
| |
| i j . April 10, 1987 41 )
| |
| i 1
| |
| | |
| TABLE 28. COMMITTED DOSE EQUIVALENTS FOR INGESTION OR IN
| |
| , OF 10 nci (370 Bq) 0F Ra-226 l
| |
| h Dose Equivalent (mrem) i Organ Ingestion Inhalation Gonads 3.4 -
| |
| Red Marrow 22 -
| |
| Bone Surfaces 250 280 Lungs -
| |
| 590 TABLE 29.
| |
| COMMITTED DOSE EQUIVALENTS FOR INGESTION OR INHALATIOh !
| |
| 0F 20 nCi (740 Bq) of Am-241 !
| |
| Dose Equivalent (mrem)
| |
| Organ Ingestion Inhalation i
| |
| ~
| |
| Canads 10 l 2400 Red Marrow 62 15000 j; Bone Surfaces 810 180000 1
| |
| Liver 170 41000 I
| |
| 9 J April 13, 1987 42
| |
| | |
| TABLE 30. COMMITTED DOSE EQUIVALENTS FOR INHALATION OF Co-60 FROM INCINERATOR EMISSIONS Maximum Average Population '
| |
| Individual (urem) Individual (urem) (person-rem)
| |
| Organ Class W Class Y Class W Class Y Class W Class Y Gonads 0.30 0.15 0.011 -
| |
| Breast 0.31 -
| |
| 0.16 -
| |
| 0.012 -
| |
| Red Marrow 0.31- -
| |
| 0.16 -
| |
| 0.012 -
| |
| Lungs 2.7 25 1.3 13 0.095 0. 9'S Lower Large Intestine 0.61 -
| |
| 0.30 -
| |
| 0.022 -
| |
| Liver 0.68 -
| |
| 0.34 -
| |
| 0.025 -
| |
| Remainder of body 0.60 0.30 0.022 -
| |
| l l
| |
| l l
| |
| 1 i
| |
| -1 April 13, 1987 43
| |
| | |
| TABLE 31.
| |
| COMMITTED DOSE EQUIVALENTS FOR INHALATION
| |
| 'OF Cs-137 FROM INCINERATOR EMISSIONS Maximum Average Population Organ i
| |
| Individual (urem) Individual (urem) (person-rem)
| |
| } Conads 2.4 l Breast 1.2 0.087 2.1 1.1 Red Marrow 0.077 2.2 1.1 Lungs 0.082 2.4 1.2 0.087
| |
| ; Thyroid 2.1 j 1.1 0.078 Bone Surfaces 2.1 1.1 i Small Intestine 0.078 2.5 1.2 0.090 l Upper Large
| |
| ; Intestine 2.4 1.2 Lower Large 0.089 Intestine 2.5 1.2 Remainder of 0.090 body 2.6 1.3 0.094 l-I I
| |
| TABLE 32. COMMITTED DOSE EQUIVALENTS FOR INHALATION l
| |
| OF Ra-226 FROM INCINERATOR EMISSIONS i
| |
| L Organ Maximum Individual (mrem)
| |
| Average Individual (mrem)
| |
| Population (person-rem) j Lungs 2.1 1.0 76 l
| |
| Bone Surfaces 0.99 0.49 1
| |
| 36 i
| |
| ; -* April 10, 1987 44
| |
| | |
| TABLE 33. COMMITTED DOSE EQUIVALENTS FOR INHATATION OF Am 241 FROM INCINERATOR EMISSIONS Maximum Average Population
| |
| ' Organ Individual (arem) Individual (mrem) (person-rem)
| |
| Gonads 8.3 Red Marrow 4.1 300 52 26 l
| |
| Bone Surfaces 1900.
| |
| 650 320 Liver 140
| |
| '24000.
| |
| 71 .5200 TABLE 34. COMMITTED DOSE EQUIVALENTS FOR INGESTION OF 0.033 mci (1.2 MBq) OF Ra-226 I
| |
| i i
| |
| t
| |
| ; Organ Dose Equivalent (rem)
| |
| .b i
| |
| i Gonads 11 l Red Marrow 72 !
| |
| l Bone Surfaces 820 ,I L i i
| |
| 'I!
| |
| .li>
| |
| l
| |
| : i. li l !!
| |
| l li j
| |
| i .
| |
| I t-i i April'10, 1987 45 I
| |
| r.
| |
| r t
| |
| | |
| 3.5) CLASS C BETA GAUGES: BACKSCATTER TYPE 3.5.1) Device Description l These . devices are widely used in production facilities in monitoring process lines or in measuring thickness, density, or composition of such materials as plastic, paper, steel sheets, precious metal platings, plating of circuit boards, and plastic coatings.
| |
| Devices can be permanently mounted or portable. Nuclides common to the '
| |
| beta and backscatter gauges include Sr-90, C-14, Pm-147. T1-204, Ru-106, Pb-210. Radiotoxicity ranges from moderate to very high.
| |
| Approximately 1,200 beta backscatter gauges are distributed per year.
| |
| i Because these devices may be portable with the sources being
| |
| . interchangeable, !
| |
| the beta backscatter gauges may be easily misplaced.
| |
| They may be lost or discarded in the trash. If found, they may appear.
| |
| to have some salvage value, and thus may be transferred to a salvage dealer.
| |
| 3.5.2) Scenario Development i
| |
| The probabilities assigned to the various pathways (Figure 6) were based on- the appearance of the beta backscatter gauges, their size, their portability, and their apparent salvage value. The highest probabilities were assigned to the pathways leading from the initial event of the device being discarded to the trash (0.21 each that the l
| |
| device would be recovered by another individual or remain in the~ trash) ;
| |
| and to the pathways in which the device remains in place or in storage on site (0.15 each). Because their external appearance might lead some to believe that the devices might have some salvage value, the~ pathway j leading from the device being recovered by an individual to the salvage !
| |
| dealer -was weighted more heavily (0.8). It was thought about equally probable that the device would be kept by the salvage dealer, sent to metals recycling, or buried in a landfill, but less probable that the salvage dealer would give it to another individual. Therefore, the ,
| |
| probability of 0.3 was assigned to each of the former pathways and 0.1 to the latter.
| |
| These assumptions indicate that the highest probabilities are that the devices would be buried in a sanitary landfill (0.28), remain in place without control (0.15), or remain uncontrolled in storage (0.15).
| |
| The predicted probability of possession by an unauthorized individual is also relatively high (0.15) due to their attractive appearance. The interchangeable sources and portability of the gauges also raises the probability of possession by unauthorized individuals. All other events appear to be less likely. These events include becoming used in consumer products or construction materials (0.12 total) or remaining in storage at the salvage yard (0.12).
| |
| 3 April 13, 1987 46
| |
| | |
| ._m. ,. ._ .. - - - - -
| |
| CLASS C-1 BETA BACKSCATTER GAUGES Dewks remehe 0.15 h place but not huse Device somehe h piece emneut 0.1s contres I
| |
| 0.15 t h storage ,
| |
| Dewho remake h a _ ._ m 0.1 S
| |
| % en elle ,
| |
| i Device tensferred 0.05
| |
| : k. dew h possemehn of 0.15 unautherhed hdev&ol Dewks specorded OSI
| |
| ,e.b. _.4 X~ '
| |
| 0.,
| |
| -h . - -
| |
| .O. .t .,,,
| |
| h onether 0.3 .
| |
| to trash eelvage yard N helvkfuel 021 Salvage 0.30 dealer esasene 0.040 Treen 0.5 Dowtoe discereed handter 0.1 bb*
| |
| 0.67 O.21 E021 D 0.03 Constructka 0.3 ***'''' '
| |
| Devbo sold 0.11 to meswege deste, Devloe buried h eenhery 0.278 landfig
| |
| =- _ .: h n03 envbenmmt 1
| |
| Figure 6 Scenario development for device class C-1, beta backscatter gauges
| |
| ,-.___2_____ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ . - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -_ . - _ _ - _ _ . _-__-4_
| |
| | |
| I 3.5.3) Dosa Assessment 3.5.3.1) ' Intact' Source External dose - maximum individual l Dose equivalents for exposure to point sources for 20 weeks at 100 i' cm (photon) or in contact for three hours (beta) at the activity levels stated in section Pb-210 are shown in table 6.3.5.1 for Sr-90, T1-204, au-106, Pm-147, C-14, and ;
| |
| External dose - realistic case External photon dose ratus near the devices were very low, listed in as high enough the NRC registry. The beta dose rates listed in table 6 are for desquamation to transitory e rythema or ulceration and moist occur, based on the estimates in NCRP 39 (NCRP 1971).
| |
| More realistic estimates coult. be based on other arbitrary estimates of i l
| |
| contact time less than th ree hours, with the estimates in table 6 scaled accordingly. It is still possible that estimates might exceed the per limits for radiation workers for dose to the extremities (18.75 rem calendar quarter). These devices present an appreciable hazard for external irradiation.
| |
| Internal dose - maximum f.ndividual i
| |
| D with ICRP an f 30 assigns all soluble compounds of Sr-90 to inhalation class l of 0.3. SrTiO3 is assigned to class Y with an fl of 0.01.
| |
| All compounds distributed of carbon are throughout allassumed to be instantaneously and uniformly are retained indefinitely. organs and tissues of the body where they For Pm-147, oxides, hydroxides, carbides, and fluorides are assigned to inhalation class Y and all others are assigned to class W. {
| |
| All compounds are assigned an ft of 0.0003. All {
| |
| compounds of thallium are assigned to class D with an ft of 1.0. All compounds of Ru-106 are assigned an ft of 0.05. Oxides and hydroxides of ruthenium are assigned to class Y, halides to class U, and all others to class D. All commonly occurring compounds of lead are assigned to class D, with an ft of 0.2.
| |
| As for the previously discussed devices with industrial applications, a fraction of 0.3 was assigned as the maximum amount of the total source activity which might be inadvertently taken into the body.
| |
| 0.3 Tables 35-39 list the dose equivalents corresponding to intake of of the source activities listed in table 1. The committed dose equivalent for intake of 0.015 mci (0.56 MBq) of C-14 is 0.031 rem (0.31 mSv) to any organ.
| |
| Internal dose - realistic case Intake lower than values from the study of Wilmot were used unless they were the 10-6 value suggested by Ricks et al., in which case an intake value of 10-6 was assigned. This scheme results in the assignment of fractions of 10-6 to all nuclides except C-14, for which a fraction of 10-0 was used. Tables 40-44 list the dose equivalents for
| |
| > April 13, 1987 48
| |
| | |
| intakeequivalent dose of those fractions for intakeofofthe 5 xinitial 10-6 source activity. The committed mci (1.85 x 10-4 MBq) of C-14 is 1.0 x 10-5 rem (1.0 x 10-7 Sv) to any organ.
| |
| 3.5.3.2) Dispersed Source 3.5.3.2.1) Incineration 10~9 The MBq)model of Buckley predicts the intake of 8.7 x 10-11 mci (3.2 x of.Sr-90, 6 10-11 mci (2 6 x 10-8 mci (1.1 x 10'9MBq9 x 10-10 2 of Pm-147, 3.5 x 10-10 m )Ci (1.3 x 10-8of Ru-106, 1.7MB mci (6.4 x 10- x 10-10MBq mci (1.3 x 10-9 MBq) of C-14, and 3.5- x 10'g{
| |
| MBq) of Pb-210 by the maximum individual. Tables 45-49 and' average individualscorresponding to these intakes to the maximum list the dose equivalents and the repulation near the incinerator. The committed dose equivalents from . inhalation of C-14 to the maximum xindividual, 10-12 Sv), average individual, and population are 7.3 x 10-10 rem (7.3 3.6 x 10-10 rem (3.6 x 10-12 gy),' and 2.6 x 10-5 person-rem (2.6 x 10-7 person-Sv), respectively.
| |
| 3.5.3.2.2) Metals Recycling As stated in section 2.3.2.2, a general result for Co-60 gauges is the only result afforded by the available data. See that section for a discussion of this result.
| |
| 3.5.3.2.3) Burial in Landfill The model of Buckley did not list values of K for several radionuclides considered in this section. A value of-1.0 was assigned in all such cases. The model then predicted'inrakes of 1.8 x 10-4 mci ;j (6.7 x 10-3 MBq) of Sr-90 6.2 x 10-5 mci (2.3 x 10-3 MBq) of T1-204 'i 8.7 x 10-8 MBq) of C-14, mci (3.2 x 10-6,MBq)5 of Ru-106,1.24 x 10~3 mci (4.6 x 10*I and 6.0 x 10- mci (2.2 x 10~3 MBq) of Pb-210 at the withdrawal point. The _model predicted that all of the Pm 147 would l decay before reaching the withdrawal point. Dose equivalents corresponding to these intakes are listed in tables 50-53. The I committed dose equivalent for ingestion of this amount of C-14 is 2.6 x ;
| |
| 10'3 rem (2.6 x 10-5 Sv) to any organ.
| |
| l l l
| |
| L i
| |
| 2 April 13, 1987 49
| |
| | |
| . - . . .-_ ~. . . . - . . .. - . ~ . . . . .. ._. . ~ . . . . , _ - . - - ~ . = -
| |
| i 1 i i
| |
| ~
| |
| i TABLE 35. COMMITTED DOSE EQUIVALENTS FOR INCESTION OR INHALATION OF b i
| |
| i
| |
| * 0.0075 mci (2.8 x 105Bq) sr-90 1
| |
| : I Dose Equivalent (rem)_
| |
| 1 l
| |
| Orgen Ingestion Inhalation- l j
| |
| f i -0.3 f i-0.01 Class D Class Y Red Marrow 5.3 x 10 0 1.8 x 10-1 9.2 x'10 0 .
| |
| j
| |
| = Bone Surfaces 1.2 x 10 1 3.9 x 10-1 2.0 x 10 1 -
| |
| ! Upper.Large Intestine
| |
| - 1.7 x 10-1 - -
| |
| i Lower Large 2
| |
| Intestine
| |
| - 7.2 x 10-l' - -
| |
| - - - 8.0 x 10 1 l Lungs i
| |
| i l -TABLE 36. COMMITTED DOSE EQUIVALENTS FOR INGESTION OR INHALATION i 0F 0.03 mci (1.1 x 10 6 Bq) OF T1-204 a
| |
| l S
| |
| i Dose Equivalent (rem)
| |
| Organ Ingestion Inhalation i
| |
| 4'.6 x '10-2 j; Gonada 7.3 x 10-2 ''
| |
| Breast 7.3 x 10-2' 4.6 x 10-2 i
| |
| 7.3 x 10-2 4.7 x 10-2 j Red Marrow 1.2 x 10-1
| |
| ' Lungs 7.3 x 10-2 1.0 x 10-1 5.1 x 10-2
| |
| ! Stomach Wall 4.6 x 10-2 Small Intestine 7.3 x 10-2 Kidneys 5.1 x 10-1 3.2 x 10-1 3
| |
| Remainder of Body 7.3 x 10-2 4.6 x 10-2 t -
| |
| )'
| |
| l l j
| |
| 4 i
| |
| s.
| |
| i 4
| |
| a 4
| |
| b 0 :
| |
| April 10, 1987 50 i
| |
| 3 1
| |
| | |
| i i
| |
| ]
| |
| l TABLE 37.
| |
| COHMITTED DOSE EQUIVALENTS FOR INGESTION OR INHALATION OF 0.0075 mci (2.8 x 105 Bq) 0F Ru-106 Dose Equivalent (rem)
| |
| Organ Inhalation Ingestion Class D Class W Class Y
| |
| , Conads -
| |
| 3.9 x 10 1 -
| |
| Breast -
| |
| 3.9 x 10-1 - -
| |
| Red Marrow -
| |
| 3.9 x 10-1 - -
| |
| ~
| |
| Lungs 5.0 x 10-1 5.8 x 100 2.8 x 101 Stomach Vall -
| |
| 3.9 x 10-1 - -
| |
| i Small Intestine -
| |
| 4.2 x 10-1 - -
| |
| Upper Large Intestine 6.9 x 10-1 4.7 x 10-1 - - -
| |
| Lower Large Intestine 2.0 x 100 6.9 x 10-1 - -
| |
| Thyroid -
| |
| 3.9 x 10-1 - -
| |
| , Bone Surfaces -
| |
| 3.9 x 10-1 - -
| |
| , Remainder of Body -
| |
| 3.9 x 10-1 - -
| |
| t TABLE 38.
| |
| C0KMITTED DOSE EQUIVALENTS FOR INGESTION OR INHALATION OF 0.015 mci (5.6 x 105 Bq) OF Pm-147 i
| |
| Dose Equivalent (rem)
| |
| Inhalation Organ Ingestion Class W Class Y
| |
| , Upper Large Intestine 6.1 x 10-2 - -
| |
| Lower Large Intestine 1.8 x 10-1 . . .
| |
| Red Marrow -
| |
| 4.6 x 10-1 -
| |
| Bone Surfaces -
| |
| 5.6 x 100 ,
| |
| Lungs -
| |
| 5.4 x 10-1 4.3 x 100 Liver -
| |
| 1.5 x 100 .
| |
| 4 April 10, 1987 51
| |
| | |
| TABLE 39.
| |
| i I
| |
| COMMITTED DOSE EQUIVALENTS FOR INGESTION OR INHALA 0F 0.003 mci (1.1 x 105 Bq) 0F Pb-210 Organ Dose Equivalent (rem)
| |
| Ingestion Inhalation Kidneys 3.1 x 101 Liver 7.9 x 101 6.8 x 101 1.7 x 102 Red Marrow 1.6 x 101 Bone Surfaces 4.1 x 101 2.4 x 102 i 6.1 x 102 t
| |
| I TABLE 40.
| |
| COMMITTED DOSE EQUIVALENTS FOR INCESTION OR INHALATION 0F 2.5 x 10-8 mci (0.925 MBq) Sr-90 Organ Dose Equivalent (rem)
| |
| Ingestion Inhalation f i -0.3 f t -0.01 Class D Class Y Red Marrow 1.8 x 10-5 5.9 x 10-7 3.1 x 10-5' Bone Surfaces 3.9 x 10-5 1.3 x 10 6 6.8 x 10-5 Lungs -
| |
| Upper Large Intestine 2.7 x 10 4 Lower Large Intestine 5.6 x 10-7 -
| |
| 2.4 x 10-6 . .
| |
| j TABLE 41. l i
| |
| COMMITTED DOSE EQUIVALENTS FOR INCESTION OR INHALATION 0F 5 x 10-8 mci (1.85 Bq) 0F Pm-147 Dose Equivalent (rem) -
| |
| Organ Inhalation Ingestion Clas: W cl.'ss Y Upper Large Intestine 2.0 x 10*7 -
| |
| Lower Large Intestine 5.9 x 10-7 .
| |
| Red Marrow -
| |
| 1.5 x 10-6 .
| |
| Bone Surfaces -
| |
| 1.8 x 10-5 .
| |
| Lungs Liver 1.8 x 10-6 1.4 x 10-5 l
| |
| 5.0 x 10-6 .
| |
| April 10, 1987 52
| |
| | |
| TABLE 42.
| |
| COMMITTED DOSE EQUIVALENTS FOR INGESTION OR INHAIATION OF 1 x 10-7 mci (3.7 Bq) 0F T1-204 Dose Equivalent (rem)
| |
| Organ Ingestion Inhalation Conads 2.4 x 10-7 Breast 1.5 x 10-7 2.4 x 10-7 1.5 x 10-7 Red Marrow 2.4 x 10-7 1.6 x 10-7 Lungs 2.4 x 10-7 4,1 x 10-7 Stomach Wall 3.4 x 10-7 1.7 x 10-7 Small Intestine 2.4 x 10-7 1.5 x 10-7 Kidneys 1.7 y 10-6 t,1 x 10-6 Remainder of Body 2.4 x 10-7 1.5 x 10-7 TABLE 43.
| |
| COMMITTED DOSE EQUIVALENTS FOR INGESTION OR INHAIATION OF 2.5 x 10-8 mci (0.925 MBq) 0F Ru-106 Dose Equivalent (rem)
| |
| Organ Inhalation Ingestion Class D Class W Class Y Stomach Vall -
| |
| 1.3 x 10-6 .
| |
| Small Intestine Upper Large Intestine 1.4 x 10-6 . . I.
| |
| 2.3 x 10-6 1.6 x 10-6 .
| |
| Lower Large Intestine 6.6 x 10-6 2.3 x 10-6 . .
| |
| Lungs Gonads 1.7 x 10-6 1.9 x 10-5 9.2 x 10-5 Breast 1.3 x 10-6 . .
| |
| 1.3 x 10-6 . .
| |
| Red Marrow -
| |
| 1.3 x 10-6 . .
| |
| Thyroid -
| |
| 1.3 x 10-6 . .
| |
| . Bone Surfaces -
| |
| 1.3 x 10-6 . .
| |
| Remainder -
| |
| 1.3 x 10-6 . , .
| |
| April 13, 1987 53
| |
| | |
| TABLE 44.
| |
| COMMITTED DOSE EQUIVALENTS FOR INCESTf0N OR INHAIATION OF 2 x 10-5 mci (0.37 MBq) OF Pb-210 Dose Equivalent (rem)
| |
| Organ Ingestion Inhalation Kidneys 1.0 x 10-4 Liver 2.6 x 10-4 2.3 x 10-4 5.5 x 10-4 Red Marrow 5.6 x 10-5 Bone Surfaces 1.4 x 10-4 8.1 x 10-4 2.0 x 10-3 TABLE 45.
| |
| COMMITTED DOSE EQUIVALENTS FOR INHAIATION OF Sr-90 FROM INCINERATOR EMISSIONS Maximum Average Population Organ Individual (rem) Individual (rem) (Person-rem)
| |
| Class D Class Y Class D Class Y Class D Class Y Red Marrow 1.1 x 10-7 Bone 5.3 x 10-8 -
| |
| 3.9 x 10-3 -
| |
| surfaces 2.3 x 10-7 1.2 x 10-7 Lungs 8.6 x 10-3 .
| |
| 9.3 x 10-7 -
| |
| 4.7 x 10-7 -
| |
| 3.4 x 10-2 l
| |
| TABLE 46. COMMITTED DOSE EQUIVALENTS FOR INHALATION OF T1-204 FROM INCINERATOR EMISSIdNS .
| |
| Maximum Average Population Organ Individual (rem) Individual (rem) (Person-rem)
| |
| Conads 1.1 x 10-9 5.3 x 10-10 3.8 x 10-5 Breast 1.1 x 10-9 5.3 x 10-10 Red Marrow 3.8 x 10 1.1 x 10-9 5.4 x 10-10 3,9 x 10-5 Lungs 2.8 x 10-9 1.4 x 10-9 Stomach Wall 1.0 x 10-4 1.2 x 10-9 5.9 x 10-10 4.3 x 10-5 Small Intestine 1.1 x 10-9 5.3 x 10-10 3.8 x 10-5 Kidney 7.4 x 10-9 3.7 x 10-9 Remainder of 2.7 x 10-4 Body 1.1 x 10-9 5.3 x 10-10 3.8 x 10-5
| |
| ~~
| |
| April 10, 1987 54
| |
| )
| |
| | |
| I i
| |
| l TABLE 47.
| |
| COMMITTED DOSE EQUIVALENTS FOR INHALATION OF Ru-106 FROM INCINERATOR EMISSIONS Maximum Average Population Organ Individual (rem) Individual (rem) (Person-rem) ]
| |
| Class W Class Y Class W Class Y Class W Class.Y ,
| |
| Lungs 2.3 x 10-8 1.1 x 10-7 1.1 x 10 8 5.4 x 10 8 8.2 x 10-4 3.9 x 10-3 i
| |
| i TABLE 48. T COMMITTED DOSE EQUIVALENTS FOR INHALATION OF Pm-147 FROM INCINERATOR,,CNISSIONS 5
| |
| ' Maximum !
| |
| Average Population Individual (rem) Individual (rem) (Person-rem)
| |
| Organ Class W Class Y Class W Class Y Class W Class Y Red s Marrow ''3 x 10*9 2.6 x 10-9 -
| |
| 1.9 x 10-4 l Lungs 6.2 x 10*9 x 10-8
| |
| }
| |
| - Bone
| |
| .',.0 3.1 x 10-9 2.5 x 10-8 2.3 x 10*4 1.8 x 10-3 )
| |
| Surfaces 6,4 x 10'' -
| |
| 3.2 x 10-8 -
| |
| 2.3 x 10-3 2 I Liver- 1.7 x 10-8 -
| |
| 8.7 x 10-9 -
| |
| 6.3 x 10-4 i'
| |
| l 1 i
| |
| TABLE 49.
| |
| l l
| |
| COMMITTED DOSE EQUIVALENTS FOR INHALATION ;!
| |
| it l-OF Pb-210 FROM INCINERATOR EMISSIONS i Maximum Average Population Organ Individual (rem) i Individual (rem) (Person-rem; i Red Marrow 4.8 x 10-7 2.4 x 10-7 1.7 x 10-2 Bone Surfaces 7.l'x 10-6 3.5 x 10-6' 2.6 x 10-1
| |
| {
| |
| i Kidneys 9.1 x 10*7 4.6 x 10-7 3,3 x to-2 '
| |
| Liver 1.9 x 10-6 9.6 x 10"7 7.0 x.10-2 1 1
| |
| i i
| |
| April 10, 1987 55 i
| |
| a i
| |
| 4
| |
| | |
| l TABLE 50. COMMITTED DOSE EQUIVALENTS FOR INGES1 ION OF I
| |
| . 1.8 X 10-4 mci 0F Sr-90 Organ Dose Equivalent (rem) i fi - 0.3 fy - 0.01 ;
| |
| Red Marrow 0.13 4,3 x 10~3 Bone Surfaces I 0.zd 9.5 x 10-3 Upper Large Intestine )
| |
| 4.1 x 10-3 Lower Large Intestine -
| |
| 1.8 x 10-2 TABLE 51.
| |
| COMMITTED DOSE EQUIVALJNTS FOR INGESTION OF 6.2 X 10 5 mci O( I' 204 l
| |
| Organ Dose Equivalent (rem)
| |
| Gonads 1.5 x 10-4 Breast 1.5 x 10-4 i Red Marrow 1.5 x 10-4 Lungs 1.5 x 10-4 ;
| |
| Stomach Wall 2.1 x 10-4 ;
| |
| Small Intestine 1.5 x 10-4 l Kidneys 1.1 x 10~3 '
| |
| Remainder of Body 1.5 x 10-4 I
| |
| 4 i
| |
| Ap ril, 10, 1987 56
| |
| | |
| TABLE 52. COMMITTED DOSE EQUIVALENTS FOR INGESTION OF
| |
| . 8.7 X 10'8 mci 0F Ru-106 Organ Dose Fe, -- l e n t (rem)
| |
| Upper Large Intestine 8.0 x 10-6 Lower Large Intestine 2.3 x 10-5 TABLE 53. COMMITTED DOSE EQUIVALENTS FOR INGESTION OF 6.0 X 10-5 mci 0F Pb-210 Organ Dose Equivalent (rem)
| |
| Red Marrow 0.33 Bone Surfaces 4.8 Kidneys 0.62 Liver 1.4 l
| |
| i April 10, 1987 57 4
| |
| l l
| |
| l-
| |
| | |
| 3 6) CLASS C BETA GAUGES: TRANSM'SS10N TYPE '
| |
| 1 3.6.1) Device Description Devices,.of this type are used to measure thickness, density, or composition of materials on process lines. These transmission beta gauges are typically mounted permanently as opposed to -the more portable backscatter beta gauge;. Sr-90 is the source most often used in 'these devices. A number of gauges are excluded from this study by limiting the activity of the source to less than 20 mci (740 MBq). '
| |
| l The physical characteristics of these devices are very similar to those i l of the gamma gauges, except that more metal is probably needed for i
| |
| shielding in some of the gamma gauges.
| |
| 3.6.2) Scenario .<elopment i
| |
| As with the gamma gauges, these devices are meant to be j j
| |
| ' permanantly mounted on production lines until they are no longer of ;
| |
| use. For these reasons, the same probabilities assigned to the gamma gauges were assigned to the beta transmission gauges, and the resultant i probabilities for final devico .tatus aere identical (Figure 7).
| |
| i 3.6.3) Dose Assessment
| |
| ! 3.6.3.1) ' Intact' Source I'
| |
| External dose - maximum individual Table 6 shows dose equivalencs for exposer
| |
| * to point sources at I I
| |
| I the activity levels stated in section 3.6,1 for Sr-90 in contact for three hours. i External dose - realistic case As with the beta backscatter gauges, the dose equivalents for skin contact are very high even with a short exposure, and may produce ;
| |
| permanent effects to the skin should an exposure of more than a few minutes take place. Because the devices are designed so that the activity may be exposed, it is conceivable that someone not knowledgeable about the hazards might receive a significant exposure.
| |
| l Internal dose - maximum individual f
| |
| l As with other gauges, a fraction of 0.3 was arbitrarily assigned as the maximum amount of the source activity which might be inadvertently taken into the body. Table 54 lists dose equivalents for intake of 6 mci (222 MBq) of Sr-90.
| |
| Internal dose - realistic case '
| |
| l j
| |
| An intake value of 10-6 was assigned to Sr-90 as described in section 3.5.3.1. Dose estimates for intake of 2 x 10~b mci (7.4 x 10-4 4
| |
| MBq) of Sr-90 are listed in table 55.
| |
| April 10, 1987 58 i
| |
| | |
| CLASS C-2 BETA TRANSMISSION GAUGES Dewece remehe 0.3 h M W net C' '** *emame h ome h plee waheut 0.3 control
| |
| &t h 8M8d P Devtee semehe h acerego at h encontroemd meerego en ene Device Geneferred to unadherM Device me h passession af 0.938 NM emmahertmed hdh44ael 0.2 0.1 Devtos remehe h dheweed 8 833 **"
| |
| m SA m 0.3 0 De 0.00 doe 8er neesses 0.220 0.33 0.033 -
| |
| 8.1s 0.053 Trson 0.3 Devios decorded handler 0.1 *** 8.57 88-- 0.033 I 0 03 0.7 Consewet6a" 9.107 0.$ w hh 0.1 hd Device tnaried 88 * *0' h eenmary 0.240 de m eendfas Device Maname L- - __'h 0.033 Figure 7 Scenario development for device class C-2, beta transmission gauges
| |
| | |
| l 3.6.3.2) Dispersed Source i
| |
| 3.6.3.2.1) Incineration Based 'on the model of Buckley, the intake by the maximum individual th!. s near sn incinerator which receives 5 devices per year of type would be 7.0 x 10-8 mi (2.6 x 10-6 MBq). Dose estimates for l the maximum and average individual and the population near an !
| |
| I incinerator are listed in table 56.
| |
| 3.6.3.2.2) Metals Recycling As stated in section 2.3.2.2, a general result for Co-60 gauges is the only result afforded by the available data. See that section for a i discussion of this result. 1 3.6.3.2.3) Burial in Landfill The model of Buckley predicts that the intake at the withdrawal point near a landfill which receives one of these devices would be 0.147 mci (5.4 MBq). Dose estimates for this amount of activity are ,
| |
| listed in table 57. ! '
| |
| l i
| |
| l l
| |
| l l
| |
| l l
| |
| l April 10, 1987 60 i
| |
| I t
| |
| l
| |
| | |
| TABLE 54. I COMMITTED DOSE EQUIVALENTS FOR INGESTION OR INHALATION 1 0F 6 mci (2.2 x 108 Bq) 0F Sr-90 Organ Dose Eauivalent (rem)
| |
| Ingestion Inhalation f i -0.3 f i -0.01 Class D Class Y Red Marrow 4.2 x 103 1.4 x 102 7.3 x 103 -
| |
| Bone Surfaces 9.3 x 103 3.1 x 102 1.6 x 104 -
| |
| Upper Large Intestine -
| |
| 1.4 x 102 . .
| |
| Lower Large Intestine -
| |
| 5.8 x 102 . .
| |
| Lungs - - -
| |
| 6.4 x 104 i
| |
| TABLE 55. C0KMITTED DOSE EQUIVAL;::TS FOR INHALATION OR INGESTION OF 2 X 10-5 mci (7.4 X 10-0 MBq) OF Sr-90 l
| |
| Dose Equivalent (rem)
| |
| Organ Ingestion Inhalation-f i -0.3 f1 -0.01 Class D Class Y Red Marrow 1.4 x 10-2 4,7 x 19-4 2.4 x 10-2 .
| |
| Bond Surfaces 3.1 x 10-2 1,o x 10-3 5.4 x 10-2 .
| |
| Upper Large Intestine -
| |
| 4.5 x 10-4 - -
| |
| Lower Large Intestine -
| |
| 1.9 x 10-3 - -
| |
| Lungs 2.1 x 10*1 TABLE 56. COMMITTED DOSE EQUIVALENTS FOR INHALATION OF Sr-90 FROM INCINERATOR EMISSIONS Maximum Average Population Individual (rem) Individual (rem) (Person-rem)
| |
| Organ Class D Class Y Class D Class Y Class D Class Y Red Harrow 8.5 x 10-5 . ;.2 x 10-5 . 3,1 .
| |
| Bone Surfaces 1.8 x 10-4 9.4 x 10-5 -
| |
| 6.8 - I Lungs -
| |
| 7.4 x 10'4 -
| |
| 3.7 x 10-4 -
| |
| 27 April 10, 1987 61
| |
| | |
| TABLE 57.
| |
| COMMITTED DOSE EQUIVALENTS FOR INGESTION OF L 0.147 mci OF Sr-90 !
| |
| Organ Dose Equivalent (rem) f'.
| |
| i f t -0.vl '
| |
| j Red Marrow 100 3.5 Bone Surfaces 230 7.6 Upper Large Intestine -
| |
| 3.3 Lower Large Intestine -
| |
| 14 I
| |
| 4 I
| |
| I l
| |
| \
| |
| l s
| |
| l l
| |
| April 10, 1987 62
| |
| | |
| 3.7) CLASS D - CAS CHROMATOCRAPHS 3.7.1) Device Description Cas e,hromatographs are laboratory analytical instruments containing ionization sources in detector cells or electron capture detectors. Electron capture detectors are used to analyze the chemi--I composition of gas sample.. Ins ti am t ..t s may have interchangeable detector cells. Common nuclides and activities are Ni-63 with 20 mci (740 MBq) and H-3 with 1000 mci (37 GBq). Approximately 900 of the Ni-63 detectors are sold per year.
| |
| These devices are complex, expensive instruments used mainly in
| |
| ' laboratories by trained, knowledgeable personnel. Licensees were more aware of the !
| |
| other devices (NRC general license program and requirements than with many '
| |
| 1987). Vendor information about gas chromatographs appears to be more comprehensive, thereby improving knowledge of license regulations.
| |
| Several manufacturers make the sources almost impossible to access by ordinary means.
| |
| I 3.7.2) Scenario Development Pathways for gas chromatographs contain a very high probability of remaining in place without control (0.8), as shown in Figure 8. This I
| |
| value was high due to the assumption that unused gas chromatographs are likely to reLain in laboratorios. Their size and expense' combined with the awareness of regulations by licensees make this outcome more likely than any of the others.
| |
| devices A smaller probability is assigned to the ,
| |
| j j remaining in place without control (0.15) due to the i possibility that they may be stored for future use. Other pathways and final conditions have probabilities which are insignificant. Some gas chromatographs may be discarded or sold to salvage dealers but these probabilities are minimal given the characteristics of the devices and regulation awareness of the licensees, i
| |
| 3.7.3) Dose Assessment 3.7.3.1) ' Intact' Source External dose - maximum individual Neither H-3 nor Nt-63 has any significant photon emissions; their l beta energies are too low to penetrate the dead layer of the skin.
| |
| therefore, these nuclides pose no threat as external sources.
| |
| Internal dose - maximum individual These devices are designed so that the sources are extremely difficult to reach or remove. It would take a significant effort to expose any of the source activity. As with other devices in which the sources are not readily accessible, a fraction of 0.3 was arbitrarily
| |
| ; assigned as the maximum amount of the source activity which might be inadvertently taken into the body. Intake of 300 mci (11,000 MBq) of H-3 would result in a dose equivalent of 19 rem (0.19 Sv) to the soft April 10, 1987 63
| |
| | |
| . .-...- - ~. . . - - . - . .. ~
| |
| t CLASS D GAS CHROMATOGRAPHS !
| |
| Desessioname es he ptsee test no, mesme m ,,,, > pasem wane s eA }
| |
| coneel e.1s !
| |
| Destes Plasse m enorego Daviso reseems m- _n e.ss eterego en site i on we esassense **
| |
| 88 - %, eme !
| |
| . et e.00s '
| |
| beewidW hr - - -
| |
| tidentsasi s.s e.1 a.c mem,e e.ets e.e Dortum sees.orse ..s , , , , , , ,
| |
| entvego F*'d
| |
| , omve.ne.,macereed o.s
| |
| , . so , my e,s ,so,on es. ,,
| |
| ; eA S.eewege
| |
| ** "**=''s
| |
| ,,, e.ss e.e 3.3 s'ets pro.ucie e.ees 1,e s.
| |
| t os se see,.se me e, ai me - an
| |
| *= ''
| |
| ,, e.e as _ ar ca e .e.,
| |
| e.s l
| |
| Devkm este OAs
| |
| ""'''*8" %%
| |
| = e.eis
| |
| > piisenna.ery "z." " ' ' " " * " " . e..
| |
| Figure 8 Scenario c.;4 wii for device class D, gas chromatographs
| |
| | |
| 3 tissues of'the whole body.
| |
| ICRP 30 -assigns oxides, hydroxides, and carbides of nickel to .i j
| |
| inhalation class W and all other compounds to inhalation class D, with all compoun,ds having an fl of 0.05. Table 58 lists dose equivalent- l commitments'from ingestion or inhalation of 6 mci (222 MBq) of Ni-63.
| |
| j Internal dose - realistic case i
| |
| Based on the study by Wilmot, values of 10-4 were assigned to both H-3 and Ni-63 as intake fractions for the realistic case. For H-3, this would result in a dose equivalent of 6.3 x 10-3 rem (6.3 x 10-2 m3y) to ,
| |
| the soft tissues of the whole body. Table 59 lists dose estimates for i intake of 2 x 10*3 mci (7.4 x 100 Bq) of Ni-63. r 3.7.3.2) Dispersed Source 3.7.3.2.1) Incineration i
| |
| The model of Buckley predicts an intake by the maximum individual near an incinerator site of 7.0 x 10-5 mci (2.6 x 10-3 MBq) of H-3 and 7.0 x 10-8 mci (2.6 x 10-6 MBq) of Ni-63. This would result in dose equivalents of 4.4 x 10-6 rem (4.4 x 10-8 Sv), 2.2 x 10-6 rem (2.2 x 10-8 Sv), and 0.16 person-rem (0.0016 person-Sv) to the maximum individual, average individual, and population, respectively fror H-3.
| |
| Estimates for Ni-o3 are shown in table 60.
| |
| 3.7.3.2.2) Metals Recycling As stated in section 2.3.2.2, a general result for Co-60 gauges is the only result afforded by the available data. See that section for a discussion of this result.
| |
| 3.7.3.2.3) Burial in Landfill Based on the model of Buckley, the intake at the withdrawal point near a landfill receiving these sources woul.d be 3.5 mci (130 MBq) of H 3, and no Ni-63 will reach the withdrawal point. The intake of this amount of H-3 would result in a dose equivalent of 0.22 rem (0.0022 Sv).
| |
| 1 i
| |
| l April 10, 1987 i 65
| |
| | |
| l TABLE 58.
| |
| COMMITTED DOSE EQUIVALENTS FROM INGESTION OR INHAIATION OF 6 mci (2.2 x 108 Bq) 0F Ni-63 D_ose Equivalent (rem)
| |
| Organ Inhalation Ingestion Class D Class W Gonads 1.9 x 100 1.8 x 101 Breast 5.6 x 100 1.9 x 100 1.8 x 101 5.6 x 100 Red Marrow 1.9 x 100 1.8 x 101 -
| |
| Lungs 1.9 x 100 1.9 x 101 6.9 x 101 Stomach Wall 2.2 x 100 1.8 x 101 -
| |
| Small Intestine 2.9 x 100 1.8 x 101 -
| |
| Upper Large Intestine 8.0 x 100 1.9 x 101 -
| |
| Lower Large -
| |
| i Intestine 2.0 x 101 2.1 x 101 1.5 x 101 l Kidneys -
| |
| 1.3 x 101 -
| |
| i Thyroid -
| |
| 1.8 x 101 -
| |
| l Bone Surfaces ,
| |
| 1.8 x 101 -
| |
| l TABLE 59.
| |
| COMMITTED DOSE EQUIVALENTS FROM INGESTION OR INHALATION OF 2 x 10-3 mci (7.4 x 100 Bq) 0F Ni-63
| |
| ! Dose Equivalent (rem)
| |
| Inhalation Organ Ingestion Class D Class W l
| |
| Gonads 6.3 x 10-4 Breast 6.1 x 10-3 1.8 x 10-3 6.3 x 10-0 6.1 x 10-3 1.8 x 10-3 Red Marrow 6.3 x 10-0 6.1 x 10-3 l Lungs 6.3 x 10-4 6.4 x 10-3 2.3 x 10-2 l Stomach Wall 7.4 x 10-4 6.1 x 10-3 -
| |
| Small Intestine 9.6 x 10-4 6.1 x 10-3 -
| |
| Upper Large Intestine 2.7 x 10-3 6'.4 x 10-3 -
| |
| Lower Large Intestine 6.8 x 10-3 7,o x 10-3 5.0 x 10-3
| |
| ; Kidneys l 6.1 x 10-3 .
| |
| 1 Thyroid -
| |
| 6.1 x 10-3 .
| |
| Bone Surfaces -
| |
| 6.1 x 10-3 . t l
| |
| l l
| |
| l April 10, 1987 66 i
| |
| l I l
| |
| | |
| .- .. -- . - - . ~ .. .. - -
| |
| TABLE 60.
| |
| COHMITTED DOSE EQUIVALENTS FOR INHALATION t i
| |
| 0F Ni-63 FROM INCINERATOR EMISSIONS
| |
| \
| |
| * Maximum Average Population organ Individual (rem) Individual (rem) (Person-rem)
| |
| Class D Clasa V ' lass D Class W Class D Class W l
| |
| Gonads 2.1 x 10-7 6.4 x 10-8 1.1 x 10-7 3.2 x 10-8 7.7 x 10-3 2.3 x 10-3 '
| |
| l Breast 2.1 x 10-7 6.4 x 10-8 1.1 x 10-7 3.2 x 10-8 7.7 x 10-3 2.3 x 10'3 l
| |
| Red Marrow 2.1 x 10-7 -
| |
| 1.1 x 10-7 -
| |
| 7.7 x 10-3 -
| |
| ! Lungs 2.2 x 10-7 8.0 x 10-7 1.1 x 10-7 4.0 x 10-7 8.2 x 10-3 2.9 x 10-2 i
| |
| Thyroid 2.1 x 10-7 -
| |
| 1.1 x 10-7 -
| |
| 7.7 x 10-3 .
| |
| Bone Surfaces 2.1 x 10-7 1.1 x 10-7 Stomach 7.7 x 10-3 -
| |
| Wall 2.1 x 10-7 1.1 x 10-7 7.8 x 10-3 Small Intestine 2.1 x 10-7 -
| |
| 1.1 x 10-7 -
| |
| 7.8 x 10-3 -
| |
| Upper Large l
| |
| Intestine 2.2 x 10-7 -
| |
| 1.1 x 10-7 -
| |
| 8.2 x 10-3 -
| |
| Lower Large Intestine Kidneys 2.4 x 10-7 1.7 x 10-7 1.2 x 10-7 8.6 x 10-8 8.9 x 10-3 6.3 x 10-3 2.1 x 10-7 -
| |
| 1.1 x 10-7 -
| |
| 7.7 x 10-3 -
| |
| i I
| |
| l l
| |
| l l
| |
| l l
| |
| 1 l
| |
| i f
| |
| I April 10, 1987 67
| |
| | |
| 3.8) CLASS E 1 - HIGH T0XICITY X-RAY FLUORESCENCE ANALYZERS 3.8.1) Device Description
| |
| -X-Ray , fluorescence analyzers (XRF analyzers) with high toicity radionuclides are used in laboratories, on process lines, or in tield use to determine the elemental composition of samples. Radioactive sotuces interest, emit soft X-rays whi.h excite atoms in the material of which in turn emit other, characteristic X-rays. Nuclides common to these instruments include Am-241 with activity of 30 mci (1100 MBq) and Cm-244 with activity of 100 mci (3700 MBq).
| |
| Approximately 90 of these analyzers are distributed per year.
| |
| In accountability, this device is somewhat similar to the gas chromatograph. X-ray fluorescence analyzers are generally large, expensive, complex, and are used by trained individuals. These factors are responsible for high accountability. Four of six licensees surveyed by the NRC were aware of license regulations, kept records, and performed leak tests.
| |
| 3.8.2) Scenario Development Pathways for the XRF analyzers are similar to the gas chromato-graph pathways. Because they are complex, expensive, and used by trained individuals, it is more likely that they would remain in place if unused (0.7). These devices could be left in labo'atories r after they close or in plants no longer in operation. XRF analyzers may also be placed in storage and forgotten (0.2). Transfer of these devices may occur due to transfer of equipment among laboratories (0.05).
| |
| Because they can be portable and are used in the field, it is presumed that some of the instruments may be discarded to the environment I
| |
| (0.05).
| |
| Final conditions reflect some of the same probabilities mentioned in the previous paragraph. The highest probabilities occur for the device remaining in place without control (0.7) and remaining in storage (0.2). The highest category after these is for the device remaining in possession of en unauthorized individual (0.08). This results from the assumed prooability for transfer directly to an unauthorized individual (0.05) and from transfer of the device after it is discarded (0.03). Other final probabilities are very small as noted in Figure 9.
| |
| 3.8.3) Dose Assessment 3.8.3.1) ' Intact' Source External dose - maximum individual Table 6 shows dose equivalents for exposure to point sources at the activity levels stated in section 3.8.1 for Am-241 and Cm-244 for 20 weeks at 100 cm. The table also shows the dose equivalent to the skin from contact to the Am-241 source for three hours (Cm-244 has no beta emissions with energy sufficient to penetrate the dead layer).
| |
| April 10, 1987 68
| |
| | |
| 2 CLASS E-1 HIGH TOXICITY XRF ANALYZERS
| |
| -- 0.7 h Penes hus
| |
| * D e.e,.e h" h pan = =esua 0.7 centos Devarepanoeg h tturage Devka somehe b esteentemed 02 hemm Device Dane $ erred EU to esnadhertend Devita reename hdtvidual h poemenmaan et 0.00t asneutherland hdistesed 0.75 0.1 Device ,esnahe e,,,g,g 0.0 E2s
| |
| ; 9.0
| |
| / g dessor testate 0.001
| |
| .d d
| |
| , 01
| |
| ::=.
| |
| U 0.7 h-~
| |
| 0
| |
| / ..
| |
| : 0. . .
| |
| Canetveteen 0.002 0.5 Dovete song 0.0 18 W Dev6ce tiurned 0.003 h iand..
| |
| senae.ry h 0.01 Figure 9 Scenario development for device class E-1, high toxicity XRF analyzers
| |
| | |
| External dose - realistic case Values quoted in the NRC registry indicate that external exposure rates or would not exceed 10 mR/hr (2.6 uC/kg-hr) to either the whole body skin. These values were recorded very near to the devices, with the shutter open, and would fall off quickly with distance. The values listed for the maximum dose equivale...s could produce transitory effects to the skin (contact with nonencapsulated source), but due to the nature of the device enclosures, these sources would not be expected to be a significant external hazard.
| |
| Internal dose - maximum individual Both Am-241 and Cm-244 represent extremely high internal hazards because of their alpha emissions and, in the case of Cm-244, spontaneous fission events. ICRP 30 assigns all compounds of both of these elements to class W with an ft of 0.0005. As with other devices in which the sources are not readily accessible, a fraction of 0.3 was arbitrarily assigned as the maximum amount of the source activity which might be inadvertently taken in. Tables 61 and 62 show the committed dose equivalent predicted by the ICRP 30 models for ingestion of 30 percent of the activity associated with these sources. As with the Am-241 gamma gauges, the dose equivalents for intake of even small amounts of the activity are significant, and any tampering w'*h the sources which results in an intake could have serious consequences.
| |
| But unauthorized individuals are not likely to have access to these devices because they are usually effectively controlled, Internal dose -
| |
| realistic case Based on the values cited in the paper by Wilmot, intake fractions of 10-6 were assigned to Am-241 and Cm-244. Tablec 63 and 64 show dose estimates for intake of these amounts of activity.
| |
| 3.8.3.2) Dispersed source 3.8.3.2.1) Incineration The model of Buckley predicts an intake of 1 x 10-7 mci (3.7 x 10-6 MBq) of Am-241 and 3.5 x 10*7 mci (1.3 x 10-5 MBq) of Cm-244.
| |
| Tables 65 and 66 list dose equivalents corresponding to intake of these amounts of activity.
| |
| 3.8.3.2.2) Metals Recycling As stated in section 2.3.2.2, a general result for Co-60 gauges is the only result afforded by the available data. See that section for a discussion of this result.
| |
| 3.8.3.2.3) Burial in Landfill The model of Buckley predicts that no Am-241 will reach the withdrawal point. A value of K is not listed for Cm-244. It seems april 10, 1987 70 i
| |
| | |
| unreesonable to assign a valus of 1 because of the nature of the radionuclide. The value for Am-241 was assigned; the model again predicts that none of the activity will reach the withdrawal point.
| |
| l l
| |
| 1 l
| |
| l l
| |
| April 10, 1987 73 l
| |
| r
| |
| | |
| l TABLE 61.
| |
| COMMITTED DO W FOUIVALENTS FROM INCESTION OR INHALATION OF
| |
| , 9 mci (3.3 x 108 Bq) 0F Am-241 Dose Equ'calent (rem)
| |
| Organ Ingestion Inhalation Gonads 4.7 x 103 1.1 x 106 Red Marrow 2.8 x 104 6.7 x 106 Bone Surfaces 3.7 x 105 8.3 x 107 Liver 7.7 x 104 1.8 x 107 TABLE 62.
| |
| COMMITTED DOSE EQUIVALENTS FROM INCESTION OR INHAIATION OF 30 mci (1.1 x 109 Bq) 0F Cm-244 Dose Equivalent (rem)
| |
| Organ Ingestion Inhalation Gonads 7.3 x 103 1.8 x 106 Red Marrow 4.9 x 104 1.1 x 107 Bone Surfaces 6.0 x 105 1,4 x 198 Liver 1.4 x 105 3,3 x 107 TABLE 63.
| |
| COMMITTED DOSE EQUIVALENTS FROM INGESTION OR INHALATTION OF 3 x 10-5 mci (1100 Bq) of Am-241 Duse Equivalent (rem)
| |
| Organ Ingestion Inhalation Gonads 0.0016 3.6 Red Marrow 0.093 22 Bone Surfaces 1.2 280 I Liver 0.26 61 l
| |
| l
| |
| ! l l
| |
| t April 10, 1987 72 I
| |
| | |
| l 1
| |
| l i
| |
| I
| |
| ; TABLE 64.
| |
| 1 COMMITTED DOSE EQUIVALENTS FROM INGESTION OR INHALATION 0F 1 x 10-4 mci (3700 Bq) of Cm-244 Dose Equivalent (rem)
| |
| Organ Ingestion Inhalation i Gonads 0.024 5.9 I Red Marrow 0.16 37 Bone Surfaces 2.0 480 Liver 0.48 110 I
| |
| TABLE 65. COMMITTED DOSE EQUIVALENTS FOR INHALATION i
| |
| 0F Am-241 FROM INCINERATOR EMISSIONS Maximum Average Population Organ Individual (rem) 4 Individual (rem) (Person-rem)
| |
| ; Gonads 1.2 x 10-2 6.2 x 10-3 450 4
| |
| ~ Red Marrow 7.7 x 10-2 3,9 x 10-2 2800
| |
| ' Bone Surfaces 9.6 x 10-1 4.8 x 10-1 35000 Liver 2.1 x 10-1 1.1 x 10-1 7700 l
| |
| TABLE 66. COMMITTED DOSE EQUIVALENTS FOR INHALATION i
| |
| OF Cm-244 FROM INCINERATOR EMISSIONS Maximum Average Population i Organ Individual (rem)
| |
| . Individual (rem) (Person-rem) !
| |
| Gonads 2.1 x 10-2 1.0 x 10-2 750 Red Marrow 1.3 x 10-1 6.4 x 10-2 4700 Bone Surfaces 1.7 x 100 8.4 x 10-1 61000 I Liver 3.9 x 10-1 1.9 x 10-1 14000 E
| |
| April 10, 1987 73
| |
| | |
| 3.9) CLASS E MODERATE T0XICITY X-RAY FLUORESCENCE ANALYZERS 3.9.1) Device Description X-ray fluorescence analyzers (XRF analyzers) with moderate toxicity nuclides are similar to the E-1 class of XRF analyzers except J. a t taey employ different radionuslides. Nuclides included here are cd-109 with activity of 20 mci (740 MBq) and Fe-55 with activity of 100 mci (3700 MBq). .These XRF analyzers are used in laboratories, process lines, and in the field. Approximately 90 of these are distributed each year. The physical characteristics and accountability of these devices are similar to the E-1 class of XRF analyzers.
| |
| 3.9.2) Scenario Development Because of their similarity to the other XRF analyzers, the same probabilities were assigned to this class as to class E-1, and the resultant probabilities for the final device status are identical (Figure 10).
| |
| 3.9.3) Dose Assessment 3.9.3.1) ' Intact' source l
| |
| External dose - maximum individual Table 6 shows dose equivalents for exposure to point sources of Cd-109 at the activity levels stated in section 3.9.1 for 20 weeks at 100 cm. A specific gamma constant for Fe-55 could not be located, so I
| |
| one was calculated, including all photons capable of penetrating the skin dead layer (Fe-55 has several very weak L-shell X-rays of about 0.6 kev as well as K-shell X-rays of about 6 kev). Table 6 thus also lists an estimated dose equivalent for exposure to . 100 mci Fe-55 source at 100 cm for 20 weeks. Neither nuclide has beta rticles with energy sufficient to penetrate the skin dead layer.
| |
| External dose - realistic case Values quoted in the NRC registry for dose equivalent rates near these devices were all below 5 mrem /hr (0.005 mSv/hr) with the shutter closed. With the shutter open, however, the values were as high as 26 mrem /hr (0.26 mSv/hr) whole body and 163 mrem /hr (1.63 mSv/hr) to the skin for Cd-109 and 55 mrem /hr (0.55 mSv/hr) whole body and 5200 mrem /hr (52 mSv/hr) to the skin for Fe-55. If an individual were to i work with the device often, opening and closing the shutter, he could l exceed the 500 mrem /yr (0.005 Sv/yr) limit for the general public for
| |
| ; an exposure time of 20 hours r.. t the cd-109 source or 10 hours near l the Fe-55 source. Skin dose equivalents under these situations are predicted to be 3.3 rem (0.033 Sv) for Cd-109 and 52 rem (0.52 Sv) for
| |
| . Fe-55.
| |
| 1 April 13, 1987 74 4
| |
| | |
| CLASS E-2 MODERATE TOXICITY XRF ANALYZERS h p4.co but not h uas f.ei piece-wahoeg 0.7 contal m 0.2 t-Devka kansferred 0 05 te m "***
| |
| h as 0.00s 0.73 0.1 Devka sosnahe in storage et U l E8 Devkarecovere. 0.3 Devkm meeered by enomer 0.2s esswego yard y seereen hetweeuse E0 %
| |
| s4 m m,g, 0.01
| |
| ; 0.01
| |
| =
| |
| O, 0.7
| |
| / 0,
| |
| .0.
| |
| Conotuction 0.002 0.5 a0
| |
| : m. w / D k.
| |
| e seur . n ,, am seneres
| |
| . 04 Figure 10 Scenario development for devk:e class E-2, moderate toxicity XRF analyzers
| |
| | |
| l Internal dose - maximum individual !
| |
| For. internal'' exposures, ICRP 30 assigns oxides and hydroxides of l
| |
| .Cd-109 to class Y, sulphides, halides, and nitrates to class W, and all others to class. D. l All compounds are assigned an ft of 0.05. For I Fe-55, .ICRP 30 assigns oxides, hydroxides, and halides to class W, and l
| |
| all others to class- D. All compounds are assigned an f of 0.1. As '
| |
| i with o.her devices in which the so.rces are not readily accessible, a fraction of 0.3 was arbitrarily' assigned as the maximum amount of the source activity which might- be inadvertently taken into the body. ;
| |
| Tables 67 and 68 list dose estimates for intake of this fraction of the maximum source activity listed in table 1. I s
| |
| Internal dose - realistic case values of 10-6 were assigned to cd-109 and 10-4 to Fe-55 for intake of material in the realistic case. Tables 69 and 70 list dose estimates corresponding to these levels of activity.
| |
| 3.9.3.2) Dispersed source 3.9.3.2.1) Incineration The 'model of Buckley predicts an intake by the maximum individual near an incinerator which receives 5 devices of this type per year of 7.0 x 10-8 mci (2.6 x 10-6 MBq) of Cd-109 and.3.5 x 10-7 mci (1.3 x 10-5 MBq) of Fe-55. . Tables 71 and 72 list dose estimates for the maximum and average individual and population near an incinerator.
| |
| 3.9.3.2.2) Metals Recyling
| |
| , As stated in section 2.3.2.2, a general. result for Co-60 gauges is
| |
| 'the-only result afforded by the available data. See that section for a )
| |
| discussion of this result.
| |
| 3.9.3.2.3) Burial in Landfill i The model of Buckley predicts an intake of 2.8 x 10-4 mci (1.0 x j 10-2 MBq) of Cd-109 and 0.027 mci (1 MBq) of Fe-55. Tables 73 and 74 list dose estimates for ingestion of this amount of activity.
| |
| 1 April 13, 1987 76
| |
| | |
| TABLE 67.
| |
| C0KMITTED DOSE EQUIVALENTS FROM INGESTION AND INHALATION
| |
| , OF 6 mci (2.2 x 108 Bq) 0F Cd-109 Dose E uivalent (rem)
| |
| Organ Inhalation Ingestion Class D Class W Class Y Kidneys 9.1 x 102 8.7 x 103 2.4 x 103 7.5 x 102 Liver 1.6 x 102 1.6 x 103 4.7 x 102 .
| |
| Lungs -
| |
| 3.3 x 102 1.7 x 103 Lower Large Intestine 1.0 x 102 . . .
| |
| TABLE 68.
| |
| COHMITTED DOSE EQUIVALENTS FROM INCESTION AND INHALATION OF 30 mC1 (1.1 x 109 Bq) OF Fe-55 Dose Equivalent (rem)
| |
| Inhalation Organ Ingestion Class D Class W-Gonads 1.1 x 101 5.7 x 101 2.0 x 101 Breast 1.1 x 101 5.7 x 101 1.9 x 101 Red Marrow 1.1 x 101 5.7 x 101 Lungs 2.0 x 101 1.1 x 161 5.7 x 101 1.2 x 102 Small Intestine 1.3 x 101 - -
| |
| Upper Large Intestine 1.9 x 101 5.7 x 101 -
| |
| Lower Large Intestine 3.3 x 101 6.0 x 101 3.1 x 101 Liver 3.7 x 101 1.9 x 102 6.4 x 101 Spleen 6.2 x 101 3.1 x 102 1,1 x 102 Remainder of Body -
| |
| 5.7 x 101 -
| |
| i l~
| |
| l i
| |
| April 10, 1987 77
| |
| | |
| TABLE 69.
| |
| COMMITTED DOSE EQUIVALENTS FROM INCESTION OR INHAIATION
| |
| , OF 2 x 10-5 mci (740 Bq) 0F Cd-109 Dose Equivalent (rem)
| |
| Inhalation Organ Ingestion Class D Class W Class Y Kidneys 3.0 x 10-3 2.9 x 10-2 8.1 x 10-3 2.5 x 10-3 Liver 5.5 x 10-4 5.2 x 10-3 1.6 x 10-3 .
| |
| Lunge Lower Large 1.1 x 10-3 5.8 x 10-3 3.4 x 10-4 - - -
| |
| Intestine TABLE 70.
| |
| COMMITTED DOSE EQUIVALENTS FROM INCESTION OK INHALATION OF 0.01 mci (3.7 x 105 Bq) of Fe-55 Dose Equivalent (rem)
| |
| Inhalation Organ Ingestion Class D Class W Conads 4.1 x 10 3 1.9 x 10-2 6.7 x 10-3 Breast 3.7 x 10-3 1.9 x 10-2 6.7 x 10-3 Red Marrow 4.1 x 10-3 1.9 x 10-2 6.7 x 10-3 Lungs 3.7 x 10-3 1.9 x 10-2 6.7 x 10-3 Small Intestine 4.4 x 10-3 - -
| |
| Upper Large Intestine 6.3 x 10-3 1.9 x 10-2 .
| |
| Lower Large Intestine 1.1 x 10-3 2.0 x 10-2 1,9 x 19-2 Liver 1.3 x 10-2 6.3 x 10-2 2.1 x 10-2 Spleen 2.1 x 10-2 1.0 x 10-1 3.5 x 10-2 Remainder of Body -
| |
| 1.9 x 10-2 .
| |
| April 10, 1987 78
| |
| )
| |
| | |
| i TABLE 71.
| |
| COMMITTED DOSE EQUIVALENTS FOR INHALATION .
| |
| OF Cd-109 FROM INCINERATOR EMISSIONS Maximum Avarage Individual (rem; Population Organ Class D Individual (rem) (Person rein)
| |
| Class 3 Class D ;
| |
| Kidneys 1.0 x 10 '? 5.0 y 10-3 Liver 1.8 x 10-3 3.7 x 100 9.1 x 10-6 6.7 x 10-1 i
| |
| Maximum Average Population Organ Individual (rem) Individual (rem) !
| |
| Class W Class Y Class W (Person-rem)
| |
| Class Y C1tss W Class Y Lungs Kidneys 3.9 x 10-6 2.0 x 10-5 1.9 x 10-6 1.0 x 10-5 1,4 x to 1 7.3 x 10-1 2.8 x 10-5 8.7 x In-6 1.4 x 10-5 4,4 x to-6 1,0.x 100 !
| |
| Liver 5.4 x 10-6 -
| |
| 2.7 x 10-6 3.2 x 10-1 2.0 x 10-1 -
| |
| TABLE 72. i COMMITTED DOSE EQUIVALENTS FOR INHALATION OF Fe-55 FROM INCINERATOR EMISSIONS Maximum Average Population Organ Individual (rem) Individual (rem) {
| |
| Class D Class W (Person-rem) '
| |
| Class D Class W Class D Class W Conads Breast 6.7 x 10 7 2.3 x 10*7 3.3 x 10-7 1.2 x 10-7 2.4 x 10-2 8.4 x 10 3 6.6 x 10-7 2.2 x 10-7 3.3 x 10-7 1.1 x 10-7 2.4 x 10-2 8.0 x 10 3 i Red LungsMarrow6.76.7 x 10-7 2.3 x 10-7 3.3 x 10-7 1.2 x 10-7 2,4 x 10-2 g,4 x to-3 x 10-7 1.4 x 10-6 3,3 x to-7 7,1 x to-7 2.4 x 10-2 5.2 x 10-2 Upper Large Intestine 6.7 x 10-7 -
| |
| 3.3 +. 10-7 Lower Large 2.4 x 10-2 .-
| |
| Intestine 6.9 x 10-7 3.6 x 10-7 3.5 x 10-7 1.8 x 10-7 Liver 2.2 x 10 6 7.5 x 10-7 1.1 x 10-6 3,7 x to-7 2.5 x 10-2 1.3 x 10-2 8.0 x 10-2 2.7 x 10-2 Spleen Remainder 3.6 x 10-6 1.2 x 10-6 1.8 x 10-6 6.1 x 10 7 1.3 x 10-1 4.4 x 10-2 of Body 6.7 x 10-7 -
| |
| 3.3 x 10-7 2.4 x 10 2 .
| |
| April 10, 1987 79
| |
| | |
| TABLE 73.
| |
| CORMITTED DOSE EQUIVALENTS FOR INCESTION OF 2.8 X 10~' mci OF Cd-109 Organ Dose Equivalent (rem)
| |
| Lower Large Intestine 4.6 x 10-3 Kidneys Liver 4.3 x 10-2 7.7 x 10-3 !
| |
| 1 TABLE 74.
| |
| CORMITTED DOSE EQUIVALENTS FOR INCESTION OF 0.0274 mci OF Fe-55 Organ Dose Equivalent (rem)
| |
| Gonads Breast 1.1 x 10 2 i Red Marrow 1.0 x 10-2 Lungs 1.1 x 10-2 1.0 x 10-2 Small Intestine 1.2 x 10-2 Upper Large Intestine 1.7 x 10-2 Lower Large Intestine 3.0 x 10-2 Liver Spleen 3.4 x 10-2 5.7 x 10~2 I
| |
| April 10, 1987 go
| |
| | |
| 3.10)
| |
| CLASS F - CALIBRATION OR REFERENCE SOURCES - Cs-137, Co-60 ,
| |
| Ra-226, Sr-90 3.10.1) Device Description This~c' manufacturer lass includes sources often supplied by the detector cr as and are calibratio./ used to atandard:.
| |
| analytical check instrument performance in the field are susceptible ...all in size, these sources 3 with to loss. Nuclides commonly used as calibration sources (0.15 activities include Am-241, 0.01 mci (0.37 MBq), Ra-226, 0.004 mci MBq), Cs-137, 0.10 Sr-90, 0.001 mCL (0.037 mci (3.7 MBq), Co-60, 0.01 mci (0.37 MBq),
| |
| Covered MBq), and plutonium, 0.005 mci (0.185 MBq).
| |
| under this Americium-241, section are such devices regulated under 10CFR31.5.
| |
| regulated 10CFR70.19, are covered under 10CFR31.8, and Pu-239, regulated under as class 1 and K, respectively. No estimates are available on the number of these sources distributed per year.
| |
| Detector sources are uced in instruments. laboratories to calibrate These sources are used by knowledgeable personnel, inventoried periodically, and are usually kept in a locked cabinet.
| |
| This system allows
| |
| , for a close check on the sources. However, these small sources are also taken out into the field to be used as calibration standards, which increases their susceptibility to loss and theft.
| |
| 3.10.2) Scenariu Development Pathway probabilities environments were estimated taking into account the two believed to in which these sources are used (Figure 11). The pathway but not in use be the most probable is for the device remaining in place (0.5).
| |
| This accounts for those sources used in laboratories or stored for field purposes which remain in place even if the lab is shut down, unused, or abandoned. The device also may remain in storage (0.1) without use in the laboratory.
| |
| of Due to the field use some sources, discarded in the trash higher probabilities are assigned to the device being (0.20 total). A smaller probability (0.15 total) or discarded to the environment was estimated for the transfer of the source to an unauthorized individual (0.05).
| |
| Final condition results place or in storage. Othat remain the same for the device being in the various pathways. As conditions change as they are affected by discarded to the trash or the environment,a result of these detector sources bei in a sanitary landfill (0.12). some devices will be buried A higher probability now occurs when the device remains in the possession of an unauthorized individual (0.12) as several pathways affect this final result.
| |
| These detector sources also may remain unrecovered in the environment (0.14) due to the field use of these sources. Small probabilities appear in the remaining conditions construction products).(storage at the salvage yard and consumer or April 13, 1987 81 i
| |
| 1
| |
| | |
| CLASS F CALIBRATION OR REFERENCE SOURCES
| |
| -- u h pense hus est muse f m peaceeme.m s.s caneet r
| |
| isecenerW &8 seerage en ese De,v.co t,enecerre. 8.85
| |
| .,_,,e, b"*'"****
| |
| cn w on
| |
| .e .
| |
| ce, *= Deutme:
| |
| m,v
| |
| / *** 83
| |
| =- ..
| |
| he *ege yar.
| |
| . sa. ""
| |
| o.,,_e, m O as ***"*"
| |
| e.s ar
| |
| .ias tr ca.*== a.o r.
| |
| 0.5 r
| |
| o as to estrage
| |
| =, oe m sandery 8 323 wa Dr.us em
| |
| -e susne.sse a n.
| |
| Figure 11 Sci i.fiv development for device class F, cattwalion or reference sources
| |
| | |
| . . . - - .- . - - . - _ ~ . _ _ - - - - .. -. . . . _ . - _ . _ . . . - _ - .
| |
| 3.10.3) Dosa Assessmant 3.10.3.1) ' Intact' source i
| |
| External dose - maximum individual Table 6 shows dose equivalents for exposure to point t I
| |
| I
| |
| ! the activity levels staced in ecticn 3.10.1 for Ra-226 . Cs-137, sources at and Sr-90 for 20 weeks at Co-60, I these sources are small 100 cm or in contact for three hours. Because and designed to be portable, longer contact s
| |
| I times could these cases,be envisioned if the source were to be kept in a pocket. In some of the beta energy would be attenuated by the person's clothing before reaching the skin. Dose equivalents received from gamma emissions could be considerably higher than those listed, as the distance between the source and the person would be less, and the
| |
| : exposure rate changes in proportion to the square of the distance.
| |
| External dose - realistic case ._
| |
| In most cases, 3
| |
| individuals who find a check source would only be exposed to the beta to them for a short period of time, and would rarely be exposed
| |
| ; emissions, because the sources are usually enclosed. The sources source are labeled as being radioactive, so most people would turn the one over to a local authority or discard it. Dose equivalents for a (0.05 hour Sv), contact time to an encapsulated source would be at most 5 rem 4
| |
| 2 tor Cs-137.
| |
| Internal dose - maximum individual Because of the size of the sources, it is not inconceivable that someone case might inadvertently ingest the entire source. Indeed such a
| |
| [ is discussed in section 3.14.3.1. However, the amount of material
| |
| ; which organsmight wouldbe freed from the source and available for uptake into body
| |
| ] maximum fraction not usually of 0.3 be 100% of that ingested. For this reason, a 3 was again assigned as the maximum amount of free activity which might be ingested. For inhalation, the assumption used l
| |
| i inhaledin .was previous sections that 0.3 would be the maximum that might be dislodge the again used, as it would also take a significant effort to source activity. Tables 75-78 list dose estimates for intake of 30% of the maximum source activity, as listed in table 1.
| |
| i Internal dose - realistic case On the basis of the recommendations of Wilmot, fracticaa of 10 6 i
| |
| )
| |
| were assigned to assigned to Ra-226, Cs-137, and Sr 90 and a fraction of 10'4 was '
| |
| of ingestion orCo-60. Tables 79-82 list dose estimates for these amounts 4
| |
| inhalation.
| |
| 3.10.3.2) Dispersed source
| |
| ' 3.10.3.2.1) Incineration 4r The of 1.4 x 10-11model of Buckley predicts an intake by the maximum individual mci (5.2 x 10-10 MBq) of Ra-226, 3.6 x 10-10 mci sl.3 x '
| |
| April 13, 1987 83 h
| |
| | |
| q 10-8 3.5 gg ) og x 10-12 mcics.g37, 1,9 x 10-11 mci (7.0 x 10-10 MBq) of co-60 (1.3 x 10-10 estimates for the maximum MBq) of Sr-90. Tables 83-86 list the dose and average individual and the po near an incinerr.cor which receives 5 of these sources per year.pulation 3.10.3.2.2) hetalsReeveling As stated in sectian 2.3. .2, a general tne only result afforded by the available result for Co-60 gauges is data.
| |
| discussion Of this result. See that section for a 3.10.3.2.3) Burial in Landfill 4
| |
| i The model of Buckley for landfill leaching predicts that 7.6 x 10-5 mci (2.8 x 10'3 MBq) of Ra-226 and 7.2 x 10-6 mci (2.7 x 1{
| |
| of Sr-90 will reach the withdrawal point, radionuclides will and that neither of the other radioactive decay. reach the withdrawal point before removal by Tables 87 and 88 list the dose equivalents correspondin5 to ingestion Sr-90. of these levels of activity of Ra-226 and l
| |
| 1 April 13, 1987 84
| |
| | |
| 1 i
| |
| ~
| |
| i 3 :
| |
| TABLE 75.
| |
| COMMITTED DOSE EQUIVALENTS FROM INGESTION OR INH 1 f **
| |
| OF 0.0012 mci (4.4 x 104 Bq) 0F Ra-226 .
| |
| 1-i 4
| |
| 1
| |
| ]
| |
| Dose Equivalent (rem) }
| |
| Organ Ingestion Inhalation ,
| |
| Gonads i
| |
| Red Harrow 4.1 x 10'l -
| |
| ! 2.7 x 100 Bone Surfaces .
| |
| Lungs 3.0 x 101 3.4 x 101 -
| |
| 7.1 x 101 l
| |
| 1 TABLE 76. ,
| |
| : COMMITTED DOSE EQUIVALENTS FROM INGE1]IONjf- INHALATION j _OF 0.03 m N (1.11 : 106 E-) 0F Cs-131
| |
| ', Organ Dose Equivalent (rem) a Ingestion Inhalation 1
| |
| Conads i
| |
| Breast 1.6 x 100 9.8 x 10-1
| |
| }
| |
| Red Marrow 1.3 x 100 8.7 x 10-1 i
| |
| Lungs 1.4 x 100 9.2 x 10-1 j' Thyroid 1.4 x 100 9.8 x 10-1 1.4 x 100 8.8 x 10-1 i Bone Surfaces 1.4 x 100 Small Intestine 1,6 x 100 8.8 x 10 1 Upper Large Intestine 1,9 x 100 4 1.6 x 100 1,o x 100 Lower Large Intestine 1.6 x 100 1,9 x 100 5 Remainder of Body 1.7 x 100 1,1 x 100 i
| |
| i I
| |
| l l
| |
| April 10, 1987 85
| |
| | |
| TABLE 77.
| |
| COMMITTED DOSE EQUIVALENTS FROM INGESTION OR OF 0.003 mci (1.11 x 105 Bq) 0F Co-60 Dose Equivalent (rem)
| |
| Ingest on ~
| |
| Inhalation
| |
| -Organ f i-0.05' f 3 -0.3 Class W Class Y Gonads 3.6 x 10-2 g,o x to-2 Breast 4,4 x to-2 3,3 x 100 1.2 x 10-2 5.7 x 10-2 4,7 x 10-2 Red Marrow .
| |
| Lungs 1.4 x 10 2 6.1 x 10-2 4,7 x to-2 .
| |
| 9.6 x 10-3 5.6 x 10-2 4,o x 10-1 Small-Intestine 4.0 x 10-2 9,t x to-2 .
| |
| Upper Large .
| |
| Intestine 6.3 x 10-2 1.1 x 10~1 -
| |
| Lower Large -
| |
| Intestine 1.2 x 10~1 1.6 x 10*1 Liver '9.1 x 10-2 -
| |
| 2.6'x 10-2 1,4 x 10-1 1,o x 10-1 Remainder of .
| |
| Body
| |
| '2.3 x 10-2 9.6 x 10-2 8.9 x 10-2 -
| |
| TABLE 78-.
| |
| COMMITTED DOSE EQUIVALENTS FROM INGESTION OR INHAI OF 0.0003 mci (1.1 x 104 Bq) 0F Sr-90 Organ Dose Equivalent (rem)
| |
| Ingestion Inhalation !
| |
| f i -0.3 f i-0.01 Class D Class Y Red Marrow 2.1 x 10-1 7.1 x 10-3 3.7 x 10-1 Bone Surfaces 4.7 x 10*1 1.6 x 10-2 g,1 x 10-1 Upper Large .
| |
| Intestine -
| |
| 6.8 x 10-3 .
| |
| Lower Large .
| |
| Intestine -
| |
| 2.9 x 10-2 .
| |
| Lungs -
| |
| j 3.2 x 100 j i
| |
| l April 10, 1987 86
| |
| | |
| TABLE 79.
| |
| COMMITTED DOSE EQUIVALENTS FROM INGESTION OR INHA
| |
| ,, OF 4 x 10-9 mci (0.15 Bq) 0F Ra-226 Organ Dose Ec o alent (rem)
| |
| Ingestion Inhalation Gonads 1.4 x 10 6 Red Marrow 8.9 x 10-6 ,
| |
| Bone Surfaces 1.0 x 10-4
| |
| ' Lungs' 1.1 x 10-4 2.3 x 10-4' TABLE 80.
| |
| COMMITTED DOSE EQUIVALENTS FROM INGESTION OR INHA OF 10-7 mci (3.7 Bq).0F Cs-137 Organ Dose. Equivalent (rem)
| |
| Ingestion Inhalation Conads Breast 5.2 x 10 6 3,3 x 16-6 Red Marrow 4.4 x 10-6 2.9 x 10-6 Lungs 4.8 x 10-6 3,1 x to-6 Thyroid 4.8 x 10-6 3.3 x 10 6 4.8 x 10-6 2.9 x 10-6 Bone Surfaces 4.8 x 10 Small Intestine 2.9 x.10 6 5.2 x 10-6 3,4 x to 6 l Upper Large Intestine 5.2 x 10-6 Lower Large Intestine 3.3 x 10-6 5.2 x 10-6 3,4 x to-6 Remainder of Body 5.6 x 10-6 3.5 x 10 6 l
| |
| l April 10, 1987 87
| |
| ,. \
| |
| t .. . ._- .!
| |
| | |
| TABLE 81.
| |
| COMMITTED DOSE EQUIVALENTS FROM INGESTION OR IN
| |
| ,, OF 10-6 mci (37 Bq) 0F Co 1
| |
| ;. I i Dose Equivalent (rem)
| |
| Organ Ingestica Inhalation.
| |
| f i-0.05 ft -0.3 Class W Class Y Gonads 1.2 x 10-5 2.7 x 10-5 1.5 x 10-5
| |
| -Breast 4.1 x 10-6 1.9 x 10-5 1.6 x 10-5 Red Marrow 4.8 x 10-6 2.0 x 10 5 1.6 x 10-5 Lungs 3.2 x 10-6 1.8 x 10-5 1,3 x 10-4 1.2 x 10-3 Small Intestine 1.3 x 10-5 3.0 x 10-5 .
| |
| Upper Large Intestine 2.1 x 10-5 3.6 x 10-5 .
| |
| Lower Large Intestine 4.1 x 10-5 5.2 x 10-5 3,o x 10-5
| |
| . i Liver .
| |
| 8.5 x 10-6 4.8 x 10-5 3,4 x 10-5 Remainder of Body 7.8 x 10-6 3.2 x 10-5 3,o x 10-5 TABLE 82.
| |
| COMMITTED DOSE EQUIVALENTS FROM INCESTION OR INHALA OF 1 x 10'9 mci (0.037 Bq) 0F Sr-90 Dose Equivalent (rem)
| |
| Ingestion Inhalation Organ f i -0.3 f i-0.01 Class D Class Y Red Marrow 7.0 x 10-7 2.4 x 10-8 1,2 x to-6 Bone Surfaces 1.6 x 10-6 5.2 x 10-8 2.7 x 10 .
| |
| Upper Large -
| |
| 2.3 x lu "
| |
| Intestine- - -
| |
| Lower Large -
| |
| 9.6 x 10-8 Intestine . .
| |
| Lungs - - -
| |
| 1.1 x 10-5 l
| |
| l l
| |
| f April 10, 1987 88 l l 4
| |
| )
| |
| | |
| . . . ~ . - - . . . . . . .- -- . - . .. , _ - . ~ . . ~ . - - - . . . . . . . . . - - . - . . - . . . .
| |
| TAB"LE 83.
| |
| COHMITTED DOSE' EQUIVALENTS FOR INRALATION OT Ra-226 FRO': INCINERAT" " MISSIONS Maximum Average Organ Population Individual (rem) Individual (rem) ,
| |
| (Person-rem). {
| |
| Lungs 8.2 x 10-7 Bone Surfaces 4.1 x 10-7 3.0 x 10-2 i
| |
| 3.9 x 10-7 2.0 x 10-7 1.4 x 10-2 TABLE 84.
| |
| COMMITTED DOSE EQUIVALENTS FOR INHALATION OF Cs-137 FROM INCINERATOR EMISSIONS Maximum I Organ Average Population Individual (rem) Individual (rem) (Person-remi j Gonads 1.2 x 10-8 !
| |
| Breast 5.9 x 10-9 4.3 x 10'4 '
| |
| 1.1 x 10-8 5.3 x 10-9 Red Marrow 1.1 x 10-8 3.8 x 10'4 Lungs 5.6 x 10-9 4.1 x 10-4 1.2 x 10-8 5.9 x 10-9 ,
| |
| Thyroid 1.1 x 10-8 4.3 x 10*4 !
| |
| Bone Surfaces 5.3 x 10-9 3.9 x 10''
| |
| 1.1 x 10-8 5.3 x 10-9 !
| |
| Small Intestine 1.2 x 10-8 3.9 x 10'' l Upper Large 6.1 x 10-9 4.5 x 10-4 '
| |
| Intestine 1.2 x 10-8 6.1 x 10*9 Lower Large 4.4 x 10-4 ;
| |
| Intestine 1.2 x 10 8 6.1 x 10 i Remainder of 4.5 x 10-4 i Body 1.3 x 10*8 I' 6.4 x 10*9 4.7 x 10-4 r
| |
| i I !
| |
| l April 10, 1987 89
| |
| | |
| j 1'
| |
| A l
| |
| TABLE 85.
| |
| COMMITTED DOSE EQUIVALENTS FOR INHALATION k l
| |
| OF Co-60 FROM INCINERATOR EMISSIONS
| |
| : 3. .
| |
| 1 i
| |
| i Maximum Average Population
| |
| ; Organ. Individual (rem) Individual (rem)
| |
| Class W- Class Y Class W (Person-rem)
| |
| Class Y Class W Class Y Gonads 2.8 x 10-10
| |
| : i. Breast 1,4 x 10-10 .
| |
| 1,o x 10-5 I 2.9 x 10-10 .
| |
| 1.5 x 10-1 I Red Marrow '2.9 x 10-10 1.1 x 10-5 .
| |
| Lungs 1.5 x 10-10 .
| |
| ),1 x 10-5 ..
| |
| 2,5 x 10-9 2.4 x 10-8 1.3 x 10-9 1.2 x 10-8 9.2 x 10-5 8.7 x 10-4 Lower Large I
| |
| Intestine 5.8 x 10 10 (
| |
| 2.9 x 10-10 -
| |
| 2.1 x 10-5
| |
| ' Liver 6.5 x 10-10 .
| |
| 3,2 x 10-10 .
| |
| 1 Remainder -
| |
| 2.4 x 10-5 .
| |
| 4 of Body 5.6 x 10-10 4 -
| |
| 2.8 x 10-10 -
| |
| 2.0 x 10-5 -
| |
| i TABLE 86.
| |
| COMMITTED DOSE EQUIVALENTS FOR INHATATION d, -0F Sr 90 FROM INCINERATOR EMISSIONS
| |
| ?
| |
| 1 Maximum Average
| |
| ;. Population
| |
| ; Organ Individual (rem) Individual (rem)
| |
| Class D Class Y Class D (Person-rem)
| |
| Class Y Class D Class Y -'
| |
| 3 Red Marrow 4.2 x 10-9
| |
| : j. Bone 2.1 x 10-9 -
| |
| 1.5 x 10-4 -
| |
| Surfaces 9.4 x 10-9
| |
| $ Lungs 4.7 x 10-9 -
| |
| 3.4'x 10-4 3.7 x 10-8 -
| |
| 1.9 x 10-8 . 1,4 x 10-3 i
| |
| l l
| |
| 1 5
| |
| 4 i
| |
| i i
| |
| 1 i
| |
| i e
| |
| April 10, 1987
| |
| ! 90 i
| |
| i 1
| |
| | |
| TABLE 97.
| |
| COMMITTED DOSE EQUIVALENTS FOR INGESTION OF 7.6 X 10-5 mci 0F Ra-226
| |
| 'O r"gan Dose Equivalent (rem)
| |
| Gonads Red Marrow 2.6 x 10-2 1.7 x 10~1 Bone Surfaces 1.9 x 100 TABLE 88.
| |
| COMMITTED DOSE EOUIVALENTS FOR INGESTION OF 7.2 X 10-6 mci OF Sr-90 Organ Dose Equivalent (rem) f - 0.3 f - 0.01 Red Marrow 5.1 x 10-3 Bone Surfaces 1.7 x 10-4 1.1 x 10-2 3,7 x 10-4 Upper Large Intestine Lower Large Intestine 1.6 x 10-4 6.9 x 10-4 i
| |
| l I l
| |
| 1 l
| |
| I 4
| |
| l April 10, 1987 91
| |
| | |
| 3.11) CLASS C SELF-LUMINOUS DEVICES 3.11.1) Device Description i !
| |
| This signs, em'ergency c, lass of device includes lighted warning signs such as exit light sources, Lighted gunsights and reference safety markers, and light wands.
| |
| i l activity are also ir.clude d . photometric standards with very low I are The principal nuclides in these devices
| |
| _to 1700 H-3 with mci an activity (62.9 GBq), of 5000 mci (185 CBq), Kr-85 with activity up
| |
| , MBq). and C-14 with an activity of 0.10 mCL (3.7 Approximately 50,000 of these devices are distributed per year.
| |
| of the several devices included in this classification, self- t luminous the provide exit signs are most frequently cited and discussed. They will t frame !
| |
| light source for of reference for discussion of this category. The these signs consist of phosphor-coated tubes filled }
| |
| with tritium. They l are cased in plastic and can be ordered with or without an aluminum frame. As with the self-luminous aircraft signs, i i
| |
| manufacturers life or disposing recommendof returning exit signs at the end of their'useful j
| |
| useful life may be ten, them fifteen,in a licensed radioactive waste site. The '
| |
| of sign. One manufacturer stated or twenty years depending on the type user, serial that records are kept of the end number of the signs, and date of manufacture. The signs are also labeled according to NRC regulations with the hazardous symbol, survey by UL sign, activity level, and address for return. However, the and that the NRC found that the placement of labels was inconsistent not as the label could not always be found. Recordkeeping was also consistent as the manufacturer stated. Seventy-six percent of j licensees surveyed reported that they kept the required records, t including record of purchase, installation, and transfer.
| |
| signs. The NRC survey found widespread problems with transfer of exit Vendors licensees and theysell the signs to electrical distributors as general resell the signs.
| |
| found to be unaware of regulations (14 The electrical distributors were could not make their customers awareofof16 surveyed) and therefore regulations. This resale transfer track of signs makes (NRC it impossible 1987). for vendors and regulatory bodies to keep 3.11.2) Scenario Development The probabilities assigned to the pathways were based on the labeling recordkeeping of exitby signs, their improper transfer by general licensees, licensees, appeal to lack of awareness of regulations, their (Figure 12). the general public, and their apparent dollar value if sold Initial these include transferring the devicepathways which show higher probabilities based on (0.35), to an unauthorized individual being discarded remainingtointrash place but not in use (0.20), storage (0.15), and (0.15). The outcome is also affected for these same reasons with a high probability that the devices will remain in the possession of unauthorized individuals (0.58) because of improper and recovered transfer among licensees and from the devices being discarded or sold to a salvage dealer who resells them. It is assumed they would be attractive to the public and that salvage dealers April 10, 1987 92
| |
| | |
| CLASS G-1 SELF-LUMINOUS DEVICES
| |
| -- u h Pin.es
| |
| .- haft met D.w he pt 0.2 y eesususeus ,
| |
| l r
| |
| S.15 Destos pemmeg be es. age Dowte mnetes
| |
| < to esteuneeped O.I5 storego ese she D.w.co . --- 8.33 le ti un.autena wedges W, J bi; as 0 50t estausherased
| |
| #'8***=d gg y --
| |
| e 4.d Oswene discarM Osutas by ens ser sac.es.ed 4.25 08S .. . .
| |
| to Wooh hege yerei j toevensal e.e, f ,,,,,,,
| |
| m '
| |
| g,ygg N 0.33 3,34 0a S.000 0 803 y P*88 D.
| |
| .- - : :=
| |
| .3 -
| |
| Y, . , ,
| |
| .81 w 8.13 Coneewcean'' O.006 9.S "'"""
| |
| Dowks sold 0'IO esmewage tBeeser Dewace tourted w mi.,, e.ees landfie Dewste remarie i urwecewared e, 0.01 artweerrnent Figure 12 Scenano development for device class G-1, self-luminous devices
| |
| , _ , _ _ ____.__m - - - - " - - ' ' ' ' ' ' - ' ' - - - _ _ _ _ _ _ _ _ _ _ . . . - - - - - - - - - - - - - - - - - - - - - - - - - - - ^ - ' ^ ~ -
| |
| | |
| would maks a greater profit by rasale than by matals rocycling. Metals recycling also consists of depends on whether the sign is encased in aluminum or plastic only. The other significant probabilities include the device remaining remaining in in place without control (0.20) and the device due uncontrolled storage on site (0.15). These outcomes are to
| |
| * l'ack of regulation awareness, abandoned buildings, and inconsistent labeling. Other outcomes had lower probabilities.
| |
| 3.11.3) Dose Assessment 3.11.3.1) ' Intact' Source External dose - maximum individual levels Dose equivalents for exposure to point sources at the activity stated at 100 cm are in section 3.10.1 for H-3, Kr-85, and C-14 for 20 weeks hours for C-14 shown in table 6 for Kr-85 and for contact for three H-3 and C-14 do not present external photon hazards, and Kr-85 disperse sources if the would activity were not be a contact hazard because the Kr-85 wo released.
| |
| External dose - realistic case As neither of the maximum dose equivalents would produce any observable effects in the individual, estimates for a more r-alistic situation would also not be of concern for acute effects. The limit for the general public for whole body exposure (500 mrem) might be exceeded if a Kr-85 source were kept for a substantial period of time at 100 cm distance (more than 8 days) and the occupational limit for dose equivalent to the C-14 source itself extremities (75 rem / year) could be exceeded if the were contacted for more than 5 hours, but this is less likely than the case of exposure to Kr-85.
| |
| Internal dose - maximum individual As withofother fractions devices in which the source is not normally exposed, 0.3 of the H-3 and C-14 activity were assumed to be ingested or inhaled in the maximum case. This would result in dose equivalents and 0.062 rem to the whole body or any organ of 94 rem (0.94 Sv) for H-3 the source were(0.62 mSv) for C-14. For Kr-85, it was assumed that if tampered with, 100% of the activity would be released into the smallest size room (100 m) 3 listed in the ICRP 30 dose estimate tables. Table 89 lists the dose equivalent rates to the organs listed in that report for release of this amount of activity.
| |
| Internal dose - realistic case l
| |
| Based on the values given by Wilmot, intake values of 10-4 were assigned to H-3 and C-14 and a value of 10*1 was assigned to Kr-85 (although in dose this represents a release, not an intake). This would result mSv) equivalents for H-3 to the whole body or any organ of 0.031 rem (0.31 lists dose equivalentsand 2.1forx 10-5 remm3 the 100 (2.1 x 10 7 Sv) for C-14. Table 90 room for exposure to 10% of the maximum activity listed in table 1 for Kr 85.
| |
| April 10, 1987 94
| |
| | |
| 3.11.3.2) Dispersed Source 3.11.3.2.1) Incineration The model 10-2 MBq) of H-3 and 7.0 x'10.10of Buckley predicts an intake of 3.4 x 10''
| |
| maximum individual. For mci (2.6 x 10 8 MBq) of C-14 by t ha .
| |
| concentration of 1.6 x 10-8Kr-85,
| |
| .C1 3 (5 9 x 10~7the model predicts a maximu annual dose estimates for expo /m MBq/m3
| |
| ). Table 91 lists the maximum which and average individual and population near an incin receives 5 of these devices per year. The intakes listed for H 3 would result x 10-5 rem in dose (1.1 commitments of 2.2 x 10-5 rem (2.2 x 10*7 Sv), 1.1 x 10-7 the maximum individual,Sv), average and 0.80 person-rem (0.0080 person Sv) to respectively. individual, 'and population, commitments of 1.4 10-12 Sv), and The intake x 10-9 rem listed 1 for C-14 would result in dose 5.3 x 10-5p(er.4 x 10 Sv), 7.2 x 10-10 11 son-rem (5.3 x 10-7 person-Sv) rem (7.2 x maximum individual, average individual, and population to the
| |
| , respectively.
| |
| 3.11.3 2.2) Metals Reeveling the only result afforded by the available dataAs -
| |
| gaugesstated is in se discussion of this result. .
| |
| See that section for a 3.11.3.2.3) Buttal in Landfill \
| |
| H-3 point.
| |
| The0.0025 and modelmci of Buckley for a landfill predicts 17 5 mci (0.0925 MBq) .
| |
| q of (648 MB correspond No groundwater pathway is hypothesized for Kr-85of C 14 will reach to dose equivalents (0.051 mSv), . These intakes of 1.1 rem (0.011 Sv respectively, to the'whole body or any orga)n.and 0.0051 rem i
| |
| i April 10, 1987 95
| |
| | |
| TABLE 89.
| |
| DOSE EQUIVALENT RATES FOR IMMERSION IN 1700 mc (6.3 x 1010 Bq) 0F Kr-85 IN A M3 ROOM i Organ Dose Equivalent (rem /hr) !
| |
| Conads 1 Breast 7.5 x 10-4 '
| |
| Red Marrow 6.9 x 10-4 Lungs 8.2 x 10-4 Bone Surfaces 6.0 x 10~4 Stomach Wall 8.8 x 10''
| |
| Kidneys 6.1 x 10-4 Liver 5.7 x 10-4 Spleen 5.3 x 10-4 Adrenals 6.3 x 10~4 Skin 5.6 x 10"'
| |
| Lens of the eye 2.9 x 100 1.3 x 10~3 TABLE 90. \
| |
| DOSE EQUIVALENT RATES FOR IMMERSION IN 170 mci (6.3 x 109 Bq) 0F Kr-85 IN A 100 M3 ROOM.
| |
| i Organ Dose Equivalent Rate (rem /hr) !
| |
| i Conads 1 Breast 7.5 x 10-5 Red Marrow 6.9 x 10-5 Lungs 8.2 x 10-5 Bone Surfaces 6.0 x 10-5 Stomach Wall 8.8 x 10-5 Kidneys 6.1 x 10-5 Liver 5.7 x 10-5 Spleen 5.3 x 10-5 l
| |
| Adrenals 6.3 x 10-5 Skin 5.6 x 10-5 2.9 x 10-1 Lens of the eye 1.3 x 10-4 l
| |
| d April 13, 1987 96
| |
| | |
| TABLE 91.
| |
| ANNUAL DOSE EQUIVALENTS FOR INHALATION OF Kr-85 FROM INCINERATOR EMISSIONS l 1
| |
| Maximum Average i Organ Population !
| |
| Individual (rem) Individual (rem) (Person-rem) i Gonads 2.7 x 10-7 Breast 1.4 x 10-7 1.0 x 10'2' 2.4 x 10-7 1.2 x 10-7 Red Marrow 3.0 x 10-7 8.6 x 10-3 Lungs 1.5 x 10-7 1.1 x 10-2 2.3 x 10-7 1.1 x 10-7 Bone Surfaces 3.2 x 10-7 8.2 x 10'3 Stomach Wall 1.6 x 10-7 1.2 x 10-2 2.3 x 10-7 1.1 x 10-7 Kidneys 2.0 x 10-7 8.2 x 10-3 Liver 1.0 x 10'7 7.3 x 10-3 2.0 x 10-7 1.0 x 10-7 ;
| |
| Spleen 2.4 x 10-7 7.3 x 10'3 Adrenals 1.2 x 10-7 8.8 x 10-3 2.1 x 10-7 1,o x 10-7 7,7 x 10-3 Skin 2.5 x 10-5 Lens of the eye 3.1 x 10-7 1.2 x 10-5 9,0 x 10-1 1.5 x 10-7 1.1 x 10-2 l
| |
| I i
| |
| April 10, 1987 97
| |
| | |
| 3.12) CLASS C-2 SELF-LUMIN0US DEVICES IN AIRCRAFT 3.12.1) Device Description Devices of this type include lighted warning signs which are permanently mounted.
| |
| are H-3 with activities The sources most commonly found in these devices up to 5000 activities of up to 300 mci mci (185 GBq) and Pm-147 with devices are 'll CBq). " ore than 30,000 of these presently installed in aircraft with six or more in large commercial 1987). planes and four to six in the smaller commercial planes (NRC A
| |
| (NCRP 1977).potential Thefor internal hazard exists if the device is damaged life of an aircraft exit sign as stated by one manufacturer is five years.
| |
| At least one manufacturer of these devices highly recommends returning theAssigns disposal. a to the manufacturer as the first alternative for devices second alternative, the licensee may dispose of the manufacturerin a licensed radioactive waste disposal facility. The the serial tries to keep track of each sign by knowing the end user planes are number of the sign, and the date of manufacture. However, and transferredfrequently sold and their interior parts are often removed that although a to another location. The survey done by the NRC found records of receipt,largethey number couldof licensees (14 out of 15) kept proper month. Transfer records only access records dating back one accurate records ( \'RC 19 8 7 ) were not well kept; only 10 out of 15 kept final The survey group's main concern was the other disposition of the signs if planes are stripped before resale to countries.
| |
| The devices may or may not end up in a licensed radioactive waste area or returned to the manufacturer.
| |
| 3.12.2) Scenario Development Probabilites in the pathways for leading from initial events (Figure 13) are highest individual (0.25) and the for device being transferred to an unauthorized (0.25). the device being sold to a salvage dealer high The probability of transfer to an unauthorized individual is anotherdue to the fact that aircraft are transferred from one company to appear to and devices from one plane to another. The device itself would have some value for a salvage dealer to resell to other individuals. The also high (0.15). probability of the device being placed in storage is to the trash (0.20) are considerable, Total probabilities for the device being disca returned to the manufacturer as required. as the devices will not always be initial pathways are lower. Probabilities for other place but not in use These include the device remaining in (0.05) and the device being discarded to the environment (0.10 total). These were considered to be less likely because most airplanes do not sit idle for long periods of time and because the be discarded. nature of these devica.s is such that they would not often i The
| |
| ! final status probabilities are highest for the device remaining in the possession of an unauthorized individual (0.47) because of the high probabilities assigned to such pathways as transfer to an unauthorized individual, recovery by another individual, and April 10, 1987 98 l
| |
| l l
| |
| l
| |
| | |
| CLASS G-2 AIRCRAFT SELF-LUMINOUS DEVICES mu g .
| |
| m 8.1s o,
| |
| .o-*-
| |
| , - - = a
| |
| =,=>
| |
| - is o.*.. , us
| |
| -m. .o m
| |
| : o. - -
| |
| a e ese
| |
| ,w 0.75 0.3 o.vkm,w m.
| |
| o
| |
| _ ,, ta o,,
| |
| .,. us u h e.,.g. .e 0.114
| |
| * hlu.I
| |
| *" g,,,
| |
| " ~~
| |
| -- ~~~ "
| |
| = .
| |
| %=w,w S.g D'88'"*8" 9.67 us asi Nai u ==a=''* ""
| |
| o us
| |
| = =* s=
| |
| ; .,, e454 we
| |
| """"": ==
| |
| Figure 13 Sces& dig development fordevice class G-2, aircraft self-luminous devices
| |
| | |
| trcnsfer from tha salvage dsaler. Other final condition probabilities are much lower for this device and do not appear to be as important.
| |
| 3.12.3) Dose Assessment 3.12.3.1) ' Intact' Source ,
| |
| Ex*.ernal dose - maximum .ndivid al Table 6 shows dose equivalents for exposure to point sources at the activity levels stated in section 3.12.1 for Pm-147 for 20 weeks at l 100 cm or in contact for three hours. i External dose - realistic case maximum The case, external photon dose equivalent for Pm-147 is very low for the but This indicates the contact dose from beta emissions is very high.
| |
| thatbroken public if they are not the devices open.will not be an external hazard to the Contact to the Pm-147 sources could produce observable skin damage in a short period of time.
| |
| Internal dose - maximum individual i
| |
| As witha other accessible, devices in which the activity is not easily fraction of 0.3 of the maximum source activity was arbitrarily assigned taken into the body. as the maximum amount which might be inadvertently ,
| |
| Intake of this much activity would result in a dose body equivalent of 94 rem (0.94 Sv) to the soft tissues of the whole MBq) offor H-3. Table 92 lists dose estimates for intake of 90 mci (3300 Pm-147 Internal dose - realistic case From the study of Wilmot, an intake value of 10-6 was assigned to Pm-147 and a value of 10-4 was assigned to H-3. The dose equivalent from intake of this much H-3 would be 0.031 rem (0.31 mSv). Table 93 lists dose estimates for intake of 3 x 10-4 mci (0.011 MBq) of Pm-147.
| |
| 3.12.3.2) Dispersed Source 3.12.3.2.1) Incineration The model of Buckley predicts an intake of 3.4 x 10-4 mci (0.013 MBq) o f.'
| |
| of H-3 and 1.0 x 10-6 mci (3.7 x 10-5 MBq) of Pm-147. The intake H-3 would result in a dose equivalent to the soft tissues of the
| |
| '< hole body of 2.2 x 10-5 rem (2.2 x 10-7 Sv), 1.1 x 10-5 rem (1.1 x 10-7 Sv), and 0.80 person-rem (0.008 person-Sv) individual, for the maximum average individual, and population, respectively. Table 94 lists corresponding values for Pm-147.
| |
| 3.12.3.2.2) Metals Reevling As stated in section 2.3.2.2, a general result for Co-60 gauges is the only result afforded by the available data. See that section for a April 10, 1987 100
| |
| | |
| l 1 l
| |
| l 4 i y l
| |
| discussion of this result. !
| |
| I l
| |
| : 3.12.3.2.3) Burial in Lqndfill
| |
| ~
| |
| l The inodel of Buckley for the landfill predicts that none of the Pm-147 H-3 will will reach thethe withdrawalpoint.
| |
| point, and that 17.5 mci (648 MBq) of 3 reach withdrawa' {'
| |
| l dose equivalent of 1.1 rem (0.011 Sv). The H-3 intake would result in a l
| |
| l l
| |
| l I
| |
| l l
| |
| l l
| |
| 1 i
| |
| l l
| |
| i l
| |
| i l l 1
| |
| 1 l
| |
| l l
| |
| l April 10, 1987 101
| |
| | |
| TABLE 92.
| |
| COMMITTED DOSE EQUIVALENTS FROM INCESTION OR 90 mci (3.3 x 109 Bq) 0F Pm-147 Dose Equivalents (rem)
| |
| Organ Ingestion
| |
| .thalation ,
| |
| Class W Class Y Upper Large
| |
| -Intestine 3.7 x 102 .
| |
| Lower Large .
| |
| Intestine 1.1 x 103 -
| |
| Red Marrow -
| |
| 2.7 x 103 Lungs -
| |
| 3.2 x 103 Bone Surfaces 2.6 x 104 Liver 3.3 x 10* -
| |
| 9.0 x 103 -
| |
| TABLE 93.
| |
| COMMITTED DOSE EQUIVALENTS FROM INGESTION OR INHAL 3 x 10-4 mci (1.1 x 104 Bq) 0F Pm-147 Dose Equivalent (rem)
| |
| Organ Inhalation Ingestion Class W 7 Class Y Upper Largc 1.2 x 10-3 -
| |
| Intestine Lower Large 3.6 x 10-3 .
| |
| Intestine _
| |
| Red Marrow Lungs 9.1 x 10-3 .
| |
| Bone Surfaces 1.1 x 10-2 8.5 x 10-2 Liver 1.1 x 10-1 -
| |
| 3.0 x 10-2 -
| |
| 1 April 13, 1987 102 i
| |
| i
| |
| | |
| TABLE 94.
| |
| COMMITTED DOSE EQUIVALENTS FOR INHALATIuN OF Pm-147 FROM INCINERATOR EMISSIONS Maximum Average Populat, ion Organ Individual (rem) Individual (rem)
| |
| Class W Clsss Y (Person-rem)
| |
| Class '' Class Y Class W Class Y Red Harrow 3.2 x 10-5 Lungs 1.6 x 10-5 -
| |
| 1.2 -
| |
| Bone 3.7 x 10-5 3.0 x 10-4 1.9 x 10-5 1.5 x 10-4 1.4 11 Surfaces 3.9 x 10-4 -
| |
| 1.9 x 10-4 -
| |
| 14 Liver 1.0 x 10-4 -
| |
| 5.2 x 10-5 3.8 -
| |
| I i
| |
| April 10, 1987 103
| |
| | |
| 3.13)
| |
| CLASS H - ANALYTICAL INSTRUMENTS WITH CALIBRATION S 3.13.'l) Device Description This a"c1' ass of device mainly includes liquid scintillation counters although included. few mass spectrometers and ion mobility detectors are also mci (1.5 The device usually contains Cs-137 with an activity of 0.040 use 15 mciMBq) and is well shielded inside a cabinet. Some devices may (555 MBq) Ni-63 sources.
| |
| analytical laboratories These devices are used in analytical equipment. toApproximately set and test600 theofcalibration of the these devices are distributed per year.
| |
| Liquid scintillation to the gas chromatographs and XRF analyzers. counters are quite similar in many They are large, complex, expensive one laboratory instruments licensee- among twenty-three used by trained personnel. All but surveyed (eight used liquid scintillation counters) were aware of regulations, performed leak tests, and kept proper records (NRC 1987).
| |
| . is not a major problem with these instruments. Compliance with regulations 3.13.2) Scenario Development Pathways for these device classes (Figure 14). devices appear similar to the three previous probably stay in a laboratory Because these laboratory instruments would no longer in use, the pathway to remaining in place but Because personnel are; more not in use was weightest the most heavily (0.8).
| |
| aware of regulations and because these instuments are expensive, unlikely. transfer to unauthorized individuals would be as some instrumentsA probability of 0.1 was assigned to placement in storage, may be sold are stored for later laboratory use. The device value. to a The highestsalvage dealer (0.1) since many of its parts are of without control and remaining in uncontrolled storageSmaller on site. final p; final probabilities resulted for the device remaining in the possession ,
| |
| of an unauthorized -individual (0.01), remaining in storage at the
| |
| * salvage yard (0.03), being included in a consumer product (0.01) or construction material (0.02), or being buried in a landfill (0.03). t 3.13.3) Dose Assessment 3.13.3.1) ' Intact' Source I
| |
| External dose - maximum individual Table 6 shows dose equivalents for exposure to point sources at the activity 100 cm or levels stated in section 3.13.1 for Cs-137 for 20 weeks at contact for three hours. As previously demonstrated, Ni-63 poses no external hazard.
| |
| External dose - realistic case The external photon dose rates near one of these devices would be
| |
| ~I April 10, 1987 104 l
| |
| 1 i
| |
| | |
| . _ __m .. . _ _ _. _. m m .m . ... _ _ . .. , , . . _ . _ _ _ m _ . . _ _ __
| |
| CLASS H ANALYTICAL INSTRUMENTS W/ CALIBRATION SOURCES Dorko mmoine 0.8 hplaco urehend 0.8 no.,
| |
| Devkapieced h esorage
| |
| , Devtco remahe h tecentemed &S einese en ,ene Device tensferred to unauthorized bdivkluel Devka ramene
| |
| ; b poseeensen of 0.01 unautherimed bdividual 0.5 0.1 W 0.0 Device mmeine Devka diecerded h to ereen by enospie, hdevidues 0.5 0.3 s*
| |
| yerd
| |
| . > &O 0.0
| |
| .o .
| |
| mete 0.0 0.10 33 0.2 0.03 0.010 products W h eded 0.1 ****
| |
| to erwhenmerd 0.67 2 0.0 0.0 0.0 0.3 0.7 Cenetuction 0.020 0.S """"
| |
| covkm seks 0.10 88** *9*
| |
| heter h eennery 0.03 landfig unrecovered to 8.0 erwoorsnent Figure 14 Scenario development for device class H, analytical instruments w/ calibration sources t
| |
| __________ _ _ _ _m.__ _ - - - - - - - - - - _ _ _ _ _ ____ __ ____ ___--_ - - - -- - - - - - - - - - - - - - - - - - - -
| |
| | |
| insignificant, activity because- the rate for ~the unanclosed source at this level was only 51 mram (0.51 mSv) for a 20 week exposure. The beta- emissions and the- source would not which exposed, be a hazard unless the unit were broken into because the, is unlikely in a realistic scenario, inaccessible, manufacturers of the:e units take care to make the source Internal dose - maximum individual As witha- other devices in which the activity is not easily accessible, fraction of 0.3 of the maximum source activity was
| |
| (
| |
| arbitrarily taken intoassigned as the maximum amount which might be inadvertently the body. Tables 95 and 96 lists dose estimates for intake of 0.012 mci (0.44 MBq) of Cs-137 or 4.5 mci (170 MBq) of Ni-63. s Internal dose - realistic case The study of Wilmot suggests an intake value of 10-6 In the absence of a value for Ni-63, a value of 10~4 for Cs-137.
| |
| Tables 97 was assigned. ;
| |
| the maximum source activity, as listed in table 1.and 98 list dose est '
| |
| 3.13.3.2) Dispersed Source I 3.13.3.2.1) Incineration !
| |
| The 1 near an model of Buckley predicts an intake by the maximum individual incinerator 1.5 x 10 10 mci (5.6 xwhich10*9 receives MBq) for 5 of these devices per year to be Cs-137 and 5.2 x 10-8 mci (1.9 x 10-6 MBq) for NL-63. Tables equivalents for the maximum and average individual and population.99 3.13.3.2.2) Metals Recyling As stated in section 2.3.2.2, a general result for Co-60 gauges is the only result afforded by the available data.
| |
| discussion of this result. See that section for a 3.13.3.2.3) Burial in Landfill .
| |
| The- landfill model proposed by Buckley predicts that neither Cs-137- nor Ni-63 will reach the withdrawal point before they are removed by radioactive decay.
| |
| i l
| |
| l 1
| |
| -j April 10, 1987 106
| |
| | |
| _ _m ... _. . _ . _ . . _ . _ - . _ . . _ . _ . . - _ . - . . _ . - _ . _ _ . .. .. . _ ... ___..._ . _ _ .. _ --..
| |
| e
| |
| - TABLE 95.
| |
| COHMITTED DOSE EQUIVALENTS FOR INGESTION OR INHA r
| |
| 0F 0.012 mci (4.4 x 105 Bq) 0F Cs-137 I Organ Dose Equivalent (rem)
| |
| Ingestion Inhalation Gonads Breast- 6.2 x 10-1 3.9 x 10~1 5.3 x 10-1 3.5 x 10~1 Red Marrow 5.8 x 10-1 Lungs 3.7 x 10-1 Thyroid 5.8 x 10-1 3l9 x 10-1 '
| |
| 5.8 x,10-1 r Bone Surfaces- 3.5 x 10-1 5.8 x 10-1 3.5 x 10~1 Small Intestine 6.2 x 10-1 !
| |
| Upper Large Intestine 4.0 x 10-1 6.2 x 10-1 4. 0 x 10 [
| |
| Lower Large Intestine 6.2 x 10-1 [
| |
| Remainder ~ of Body- 4.0 x 10-1 6.7 x 10~1 4.2 x 10-1 i s
| |
| TABLE 96.
| |
| l COMMITTED DOSE EQUIVALENTS FROM INGESTION OR INRALAT !
| |
| OF 4.5' mci (1.7 x 108 Bq) 0F Ni-63 '
| |
| Dose Equivalent (rem)
| |
| Organ Inhalation Ingestion Class D i Class W .
| |
| Gonads Breast 1.4 x 100 1.4 x 101 4.2 x 100 t
| |
| t 1.4 x 100 1.4 x 101 4.2 x 100 Red Marrow 1.4'x 100 l 1,4 x tol '
| |
| Lungs .
| |
| 1.4 x 100 1.4 x 101 5.2 x 101 Stomach Wall 1.7 x 100 1,4 x 101 Small Intestine 2.2 x 100 Upper Large Intestine 1.4 x 101 -
| |
| 6.0 x 100 1,4 x 191 i Lower Large Intestine 1.5 x 101 Kidneys 1.6 x 101 1.1 x 101 '
| |
| Thyroid 1.4 x 101 -
| |
| Bone Surfaces 1.4 x 101 -
| |
| {
| |
| 1.4 x 101 - '
| |
| l- -
| |
| l l
| |
| l l i
| |
| ; O April 10, 1987 107 i
| |
| i l'
| |
| i
| |
| | |
| t TABLE 97.
| |
| COMMITTED DOSE EQUIVALENTS FROM INCESTION OR 0F 4 x 10-8 mci (1.48 Bq) of Cs-137 i i
| |
| i Organ Dose Equivalent (rem) '
| |
| Ingestion Inhalation j Conads Breast 2.1 x 10-6 1.3 x 10-6 i
| |
| Red Marrow 1.8 x 10-6 1.2 x 10-6 [
| |
| Lungs 1.9 x 10-6 1,7 x in 6 '
| |
| Thyroid 1.9 x 10-6 1.3 x 10-6 Bone Surfaces 1.9 x 10-6 1.2 x 10 6 {
| |
| I 1.9 x 10-6 1.2 x 10-6 Small Intestine 2.1 x 10-6 ,
| |
| Upper Large Intestine 1.3 x 10-6 >
| |
| Lower Large Intestine 2.1 x 10-6 1.3 x 10-6 2.1 x 10-6 1.3 x 10*6 i Remainder of Body 2.2 x 10 6 1,4 x to-6 t
| |
| o TABLE 98. .
| |
| COHMITTED DOSE EQUIVALENTS FROM INCESTION OR INHA OF 1. 5 x 10 3 mci (5.5 x 104 Bq) of Ni-63 Dose Equivalent (rem)
| |
| Organ Inhalation Ingestion Class D Class W Conads Breast 4.7 x 10-4 4.6 x 10-3 1.4 x 10-3 Red Marrow 4.7 x 10 ' 4.6 x 10-3 1.4 x 10'3 Lungs 4.7 x 10'4 4.6 x 10-3 -
| |
| Stomach Vall 4.7 x 10'' 4.8 x 10'3 1.8 x 10-2 5.6 x 10-4 4.6 x 10-3 Small Intestine 7.2 x 10 4 Upper Large Intestine 4.6 x 10-3 -
| |
| 2.0 x 10-3 4.8 x 10-3 Lower Large Intestine 5.1 x 10*3 Thyroid 5.3 x 10"3 3.7 x 10-3 Bone Surfaces 4.6 x 10-3 -
| |
| Kidneys 4.6 x 10-3 -
| |
| 4.6 x 10-3 -
| |
| ) April 13, 1987 108 l
| |
| j
| |
| | |
| TABLE 99.
| |
| COMMITTED DOSE EQUIVALENTS FOR INHALATION OF Cs-137 FROM INCINEPATOR EMISSIONS Maximum Average Organ Population Individu . (rem) Individual (rem) (Person-rem)
| |
| Conads 4.8 x 10-9 Breast 2.4 x 10-9 1.7 x 10-4 4.2 x 10-9 2.1 x 10-9 Red Marrow 4.5 x 10-9 1.5 x 10-4 Lungs 2.2 x 10-9 1.6 x 10-4 4.8 x 10-9 2.4 x 10-9 Thyroid 4.3 x 10-9 1.7 x 10-4 Bone Surfaces 2.1 x 10-9 1.6 x 10-4 4.3 x 10 9 2.1 x 10-9 Small 1.6 x 10-4 Intestine 4.9 x 10-9 Upper Large 2.5 x 10-9 1.8 X 10-4 Intestine 4.9 x 10-9 Lower Large 2.5 x 10-9 1.8 x 10-4 Intestine 4.9 x 10-9 Remainder of 2.5 x 10-9 1.8 x 10-4 Body 5.3 x 10-9 2.6 x 10-9 1.9 x 10-4 TABLE 100.
| |
| COMMITTED DOSE EQUIVALENTS FOR INHALATION OF Ni 63 FROM INCINERATOR EMISSIONS Maximum Average Population Organ Individual (rem) Individual (rem)
| |
| Class D Class W Class D (Person-rem)
| |
| Class W Class D Class W Conads Breast 1.6 x 10-7 4.8 x 10-8 7,9 x 10-8 2. 4 x 10 ''l 5.8 x 10-3 1.6 x 10-3 Red Marrow 1.6 xx 10*7 1.6 10*7 4.8 x 10-8 7,9 x 10-8 2.4 x 10'8 5.8 x 10-3 1.6 x 10-3 7,9 x 10-8 Lungs -
| |
| 5.8 x 10-3 -
| |
| Thyroid 1.7xx10-7 1.6 10-7 6.0 x 10-7 7.9 x8.410-8x 10-8 3.0 x 10-7 6.1 x 10-3 2.2 x 1 Bone -
| |
| 5.8 x 10-3 -
| |
| Surfaces 1.6 x 10-7 -
| |
| 7.9 x 10-8 -
| |
| 5.8 x 10-3 Stomach Wall 1.6 x 10-7 -
| |
| 8.0 x 10-8 Small -
| |
| 5.8 x 10-3 -
| |
| Intestine 1.6 x 10-7 -
| |
| 8.0 x 10-8 Upper Large 5.8 x 10-3 -
| |
| Intestine 1.7 x 10-7 -
| |
| 8.4 x 10-8 1 Lower Large 6.1 x 10-3 -
| |
| {
| |
| Intestine Kidneys 1.8xx10-7 1.6 10-7 1.3 x 10-7
| |
| . 7,99.2 x 10-8 6.4 x 10-8 6.7 x 10-3 4.7 x 10-3!!
| |
| x 10-8 -
| |
| 5.8 x 10-3 -
| |
| O April 10, 1987 109
| |
| | |
| I 3.14) l CLASS I - CALIBRATION OR REFERENCE SOURCES - Am-241 j 3.14.1) Device Description These devices are identical in nature to 3.10, except that section those described in separate section in this they contain Am 241. They are covered in a report because they are regulated under a separate .section in the Code of Federal Regulations Source activities may be as high as 0.005 mci (0.185 MBq).(see table 1). l j
| |
| 3.14.2) Scenario Development Because the sources 3.10, the same set of are identical in nature to those in section scenario (Figure 15). probabilities are assumed to apply 3.14.3) Dose Assessment 3.14.3.1) ' Intact' Source External dose - maximum individual Table 6 shows dose equivalents for exposure to point sources at the activity levels stated in section 3.14.1 for Am-241 for 20 weeks at 100 cm or in contact for three hours. i External dose - realistic case l Both were very of the dose estimates for -external exposure to these sources low in the maximum casti, and will not be significant in a realistic situation.
| |
| I Internal dose - maximum indf*,1 dual !
| |
| As discussed for activity was assumed tothe besources free in section 3.10, 0.3 of the source :
| |
| Table if the entire source was swallowed. '
| |
| Am-241. 101 lists dose estimates for intake of 0.0015 mci (0.056 MBq) of Internal dose - realistic case
| |
| * From the study of Wilmot, a intake value of 10-6 xAm 241. Table 102 lists dose estimates for intake of 5 x 10-9was assigned to i 10-7 MBq) of Am-241. mci (1.85 ceramic In an incident in which a technologist inadvertently swallowed a disk containing 0.0028 mci (0.10 MBq) of Am-241, Smith et al. 1 (Smith 1983) found that the disk passed through the gastrointestinal system with almost dose equivalent no loss of activity from the disk. They estimated a assumed that to the lower large intestine is 17 rem (0.17 Sv), and system, which only would1.6 x 10-7 mci (5.8 Bq) were absorbed into the body result in a dose equivalent of i3 rem (0.13 Sv) to the bone surfaces, based on the ICRP 30 model.
| |
| This incident did not involve a source with characteristics similar to the ones commonly used g April 10, 1987 110
| |
| | |
| _ ._ ,. . . _ . . . _.. - -&- ' '~ ^ ' ' '
| |
| CLASS I CAllBRATION OR REFERENCE SOURCES (Am-241)
| |
| Devko sustehe S.5 h panes ense not h een Duvue susnesia h pasa..as.es e.S
| |
| ==*ee a
| |
| DeveneM
| |
| " " '8"
| |
| < 2 g,
| |
| eserese en%
| |
| Dev6te teneferred 0.0 S se esnaushertoeg hav4Wesel h 'uneen h; _ . af S.119 NI'*d heirtshsef
| |
| /
| |
| _ o ,se,, ,,,,,,,, e.coi " '""***
| |
| Devine g,,,,,,, rec,omered 0.si 0.3 -- sest totrash ,,,,,,,y,,,
| |
| hderkhsef m ; 8 848 gy, 388'898 emeter essesse D.OM ? 0.33 0.06 0.2 0 011 r ,
| |
| F" 88830
| |
| _ .e.e,es. 0.5 te z ' - --
| |
| J.'r:::r .1 ---
| |
| Oa 2
| |
| g ,4, S.M S ,,
| |
| .135 0.7 Consevettes' a.S '***"'''' S.0874 Device sete 0.0 to estrage doeter Devtee tourhd h aanmary 0.123 lendret Devka sumahe $
| |
| n_ _C h 9.135 erwesenent Figure 15 Scenario development for device class I, calitration or reference sources (Am-241)
| |
| | |
| for datector decrease in calibration, but the case study does demonstrate that soma the intestinesthe radiation dose can occur if the source passes through without much of the activity leaching off, as would be expected in most cases.
| |
| 3.14.3.2) bispersed Souree 3.14.3.2.1) Incineration The 10-9 MBq) model of Buckley predicts an intake of 3.5 x 10-11 mci 1.3 x 103 lists dose estimates for the maximum and averare i population from intakes corresponding to this es ite .
| |
| 3.14.3.2.2) Metals Recyling As stated in section 2.3.2.2, a general result for Co-60 gauges is the only result afforded by the available data.
| |
| discussion of this result. See that section for a 3.14.3.2.3) Burial in Landfill The model of Am-241 will reach the withdrawal pcint.Buckley for the landfill predicts that n
| |
| ? April 10, 1987 112
| |
| | |
| TABLE 101.
| |
| COMMITTED DOSE EQUIVALENTS FROM INCESTION OR IN _
| |
| 0F 0.0015 mci (5.6 x 100 Bq) 0F Am-241 Organ Dose Equivalent (rem)
| |
| Ingestion ;
| |
| Inhalation '
| |
| Gonads 7.7 x 10~1 1.8 x 102 Red Marrow 4.7 x 100 Bone Surfaces 1.1 x 103 Liver 6.1 x 101 1.4 x 104 '
| |
| 1.3 x 101 3.0 x 103 TABLE 102.
| |
| COMMITTED DOSE EQUIVALENTS FROM INGESTION OR INHALA '
| |
| I 0F 5 x 10-9 mci (1.85 x 10'7 Bq) 0F Am-241 Organ Dose Equivalent (rem)
| |
| Ingestion Inhalation Gonads 2.6 x 10 6 5.9 x 10-4 Red Marrow 1.6 x 10-5 Bone Surfaces 3.7 x 10-3 Liver 2.0 x 10-4 4.6 x 10-2 4.2 x 10 5 1, o x 10-:2.
| |
| TABLE 103. :
| |
| COMMITTED DOSE EQUIVALENTS FOR INHALATION OF Am-241 FROM INCINERATOR EMISSIONS Maximum Average Organ Population Individual (rem) Individual (rem) (Person-rem)
| |
| Conads 4.1 x 10-6 Red Marrow 2.1 x 10-6 0,15 2.6 x 10-5 1.3 x 10-5 o,94 Bone Surfaces 3.2 x 10-4 Liver 1.6 x 10-4 12 .
| |
| 7.1 x 10-5 3.5 x 10-5 2.6
| |
| )
| |
| ?
| |
| April 10, 1987 113 l
| |
| l
| |
| | |
| 3.15) CLASS J - SMALL QUANTITIES OF SOURCE MATERIAL Included in this class are two items in the NRC registry and both i stainless steel and arefor are calibration standards logging tools. The sources are encased in e used attached to the end of a long cable. Nuclides and'ty'pical activities are natural uranium, 0.005 mci (0.185 MBq) .
| |
| and Th-232, 0.005 mci (0.185 MBq). l Companies which receive this material to produce these sources are limited time, by no and 10CFR40.22 to reception of 15 pounds of material at any one have been placed on disposal up to this activity limit.more than 150 !
| |
| The sources themselves represent minimal external hazards.
| |
| External from intake dose of rates near these encapsulated sources are very low. Risks the sources or damage the activity if released through machining to produce chemical to an existing source are higher from the be locatedtoxicity whichthan from the radiotoxicity. No published data could assessment, so a would help with either scenario development or dose classes, was not performed, detailed analysis, as was done for the other device i r
| |
| l l
| |
| l l
| |
| l
| |
| .; April 13, 1987 114
| |
| | |
| 3.16) CLASS X - CALIBRATION OR REFERENCE SOURCES - Pu-239 3.16.1) Device Description These sections 3.10devices and 3.14, are except identical thatin nature to those described in covered in they contain Pu 239. They are a separate section in this report because they are covered in a separate section in the Code of Federal Regulations (see table 1) ,
| |
| Source activitiep may be as high as 0.005 mci (0.185 MBq). .
| |
| 3.16.2) Scenario Development Because the sources 3.10 and 3.14, are identical in nature to those in section apply (Figure 16). the sac.e set of scenario probabilities are assumed to 3.16.3) Dose Assessment 3.16.3.1) ' Intact' Source External dose - maximum individual
| |
| ; Table 6 shows dose equivalents for exposure to point sources at the activity 100 cm or inlevels contactstated in section 3.16.1 for Pu-239 for 20 weeks at for three hours. '
| |
| External dose - realistic case The dose low for the estimate maximum for case, external exposure to these sources is very situation. and will not be significant in a realistic I Internal dose - maximum individual As with the activity was assumed sources to bein free section 3.10 and 3.14, 0.3 of the source i Table if the entire source was swallowed.
| |
| Pu-239. 104 lists dose estimates for intake of 0.0015 mci (0.056 MBq) of 1
| |
| Internal dose - realistic case .
| |
| From the study of Wilmot, a intake value of 10-6 xPu-239. Table 105 lists dose estimates for intake of 5 x 10-9was assigned to 10*7 MBq) of Pu-239. mci (1.85 3.16.3.2) Dispersed Source 3.16.3.2.1) Incineration The 10-10 model of Buckley predicts an intake of 1.7 x 10'11 mci 6.4 x MBq) of Pu-239 from incineration of five sources per year.(Table 106 lists dose estimates for the maximum and average individual and the population from inhalation of this amount of activity. j
| |
| ; i
| |
| ,y April 13, 1987 115 i j
| |
| | |
| .., .__. ,.m. _ _ .. .. . _ _ ._ _ _ _ . .
| |
| r CLASS K CALIBRATION OR REFERENCE SOURCES (Pu-239)
| |
| Dewke vemesia 0.S bP88" W W h 8888 h poem wesed 0.S ceneet Derke pieced El i h eterage Devem somehe h esneenevaard &l eterage en ene Devke tenskered 0.05 to sanauthertrod bdry4 dust Devuwievnee e
| |
| ; h possesoami et 0.110 emesshariard bdividusa 0.99 0.1
| |
| * 0.001 Devko e Devre somehe Devko discarded by enouis, 0.01 0.3 *
| |
| ,,,, ,,, i
| |
| ' 0.149 O.064
| |
| * 88888' '
| |
| E884888 0 0.0438 D.evice.inocerded 0.1 hchintetor
| |
| -- 0.67
| |
| .t.. . 1. 0.
| |
| .135 0.7 Construction g,gg, 0.5 r '
| |
| Devko sold 0.0 to se8vege 8B**W Device tourted h senaery 0.123 landtag De,*e -ee,e h 0.135 Figure 16 Scenario development for device class K, calibration or reference sources (Pu-239)
| |
| | |
| l 3.16.3.2,2) Metals Recyling As stated in section 2.3.2.2, a general result for Co-60 gauges is the only result afforded by the available data.
| |
| discussion of this result. See that section for a 3.16.3.2.3) Burial in Landfill The model of Buckley for the landfill predicts an intake at the withdra'al point of 9.07 x 10-6 mci (3.4 x 10-0 MBq) of Fu-239. Table 107 lists dose estimates for ingestion of this amount of activity.
| |
| 'I 1
| |
| I l
| |
| i s
| |
| 1, I
| |
| i i
| |
| I 3 )
| |
| 1 4
| |
| ~'
| |
| April 10, 1987 117 1
| |
| il S
| |
| | |
| TABLE 104. COMMITTED DOSE EQUIVALENTS FOR INGEST OF 0.0015 mci (0.0555 MBq) 0F Pu-239 Dose Equivalent (rem)
| |
| Ingestion Organ f i-1 x 10-4 Inhalation f i-1 x 10-5 Class W Class Y Gonads 0.14 0.014 180 Red Marrov 0.89 0.089
| |
| \
| |
| Bone Surfaces 1100 420 12 1.2 Liver 14000 5300 2.4 0.24 Lungs -
| |
| 2900 1200 Upper Large -
| |
| 1800 Intestine -
| |
| 0.094 Lower Large Intestine -
| |
| 0.29 - -
| |
| TABLE 105. COMMITTED DOSE EQUIVALENTS FOR INGESTION O
| |
| . OF 5 x 10-9 mci (1.85 x 10-7 MBq) 0F Pu-239
| |
| . Dose Equivalent (rem)
| |
| Ingestion Organ f i-1 x 10-4 Inhalation f t-1 x 10-5 Class W Class Y Gonads 4.8 x 10-7 Red Marrow 4.8 x 10-8 5.9 x 10-4 3.0 x 10-6 3,o x to-7 3,7 x 10-3 1,4 x 10-3 Bone Surfaces 3.9 x 10-5 3,9 x to-6 Liver 8.1 x 10-6 4.6 x 10-2 1,3 x 19-2 Lungs -
| |
| 8.1 x 10-7 9.8 x 10-3 3.9 x 10-3 s
| |
| Upper Large 5.9 x 10-3 Intestine -
| |
| 3.1 x 10-7 -
| |
| i Lower Large -
| |
| Intestine -
| |
| 9.8 x 10-7 . ,
| |
| i 1
| |
| -1 April 13, 1987 118
| |
| | |
| d TABLE 106. COMMITTED DOSE EQUIVALENTS FOR INHALA ._
| |
| OF Pu-239 FROM INCINERATOR EMISSIONS 1
| |
| Maximum Organ Average Population Individual (rem) Individual (rem)
| |
| Class W Class Y (person rem)
| |
| Class W Class Y l Class W Class Y Gonads 2.1 x 10-6 Red Marrow 1.0 x 10-6 .
| |
| o,o77
| |
| : 1.3 x 10 5 4.9 x 10-6 .
| |
| # Bone 6.5 x 10-6 2.4 x 10-6 0.47 0.18 Surfaces 1.6 x 10~4 6.1 x 10-5 Liver 3.4 x 10-5 1,3 x to-5 8.0 x 10-5 3,o x 10-5 5.8 2.2 l Lungs 1.7 x 10-5 6,5 x 10-6 1.2 0.47 2.1 x 10-5 -
| |
| 1.0 x 10-5 ,
| |
| o,77 TABLE 107. COMMITTED DOSE EQUIVALENTS FOR INCES' I 9.07 x 10-6 mci (3.4 x 10-4 MBq) 0F Pu-239 Organ Dose Equivalent (rem) f 1-1 x 10'4 fy-1 x 10-5. 1 Gonads Red Marrow 8.7 x 10'4 8.7 x 10-5 )
| |
| 5.4 x 10-3 5.4 x 10-4 Bone Surfaces 7.0 x 10-2 !
| |
| Liver 7.0 x 10-3
| |
| ; 1.5 x 10-2 1.5 x 10-3 !
| |
| Upper Large $
| |
| i Intestine l 5.7 x 10'4 Lower Large i Intestine 1.8 x 10-3 1
| |
| I C
| |
| 1 l
| |
| l April 13, 1987 119
| |
| : 6) Summary and Coaclusions a Although small am,ount theofresults data, generated through this analysis were based on are evideht. some conclusions about public health concerns and internal radiation field irradiation. The former concern relates to ex the source. from the radioactive source in a device or contact to Internal the device, in which irradiation may occur from improper handling of activity may be released from the source, or intake of radioactivity because of the device which has been dispersed over a wide area metals recycling, or buried inadvertently in a being incinerated, passed through landfill, with radioactive material into public drinking water supplies. leaching of the Exposure to the radiation field was not a significant concern for most 20 weeks at 100 cm. even under a worst case assumption of exposure for of the devices, Devices for which exposures under this assumption i were (classgreater B), than 500 mrem (0.005 Sv) include all types of gamma gauges X-ray fluorescence analyzers (class E-1 and E-2), and self-luminous devices containing Kr-85 (class cases, contact G-1). In almost all to localized radiation doses the radioactive source for three hours will produce )
| |
| I general public. For many devices, which are inappropriate for members of the and these dose the source is extremely inaccessible, equivalents would chromatographs, not be expected (e.g., gas sources). In other analytical instruments with calibration or reference through a built-in mechanism, and significact radiation doses cad n occur if the device is manipulated by persons unaware of the radiation hazards (see table 6).
| |
| I Internal radiation doses worst case situation were derived for several situations. A material by inhalationwas defined as intake of 30% of the radioactive or ingestion. More realistic numbers were generated 10-4 by assuming a much smaller fraction, usually between 10-6 and Estimates were also generated for inhalation of radioactive material if a device were incinerated, and for ingestion of radioactive material situations,which may have leached from a landfill site. For these and activity level of the radionuclide concerned.the were results varied gre Significant estimates and, derived for inhalation of Am-241, Ra-22L , Cm-244, Pm-147, Pu-239, to a Cs 137, lesser extent, Po-210, Pb-210, Fe-55, Ni-63, Sr-90, Co-60, of Am-241, and Cm-244, Cd-109. High dose equivalents were estimated for ingestion Pm-147, Fe-55, Sr-90, Co-60, Cs-137, and and, to a lesser extent, Po-210, Pb-210, Cd-109.
| |
| of material Dose equivalents from intake Ra-226, and Cm-244 from the plume of an incinerator were highest for Am-241, were significant for Ingestion of radioactive material from landfills l Ra-226 from the and Sr-90. The model used for transport matrix landfill for somepredicted nuclides,very whichlong holdup times in the landfill soil Am-241, Cm-244, or resulted in the prediction that no significant hazards Pm-147 would be ingested. These nuclides were all if ingested, and would be of concern if the migration rates were significantly faster than predicted by this model.
| |
| 3 April 13, 1987 120
| |
| | |
| 1 i
| |
| Although more recycling of literature was availeble for consequences of metal i radionuclides than for any other !
| |
| quantitative. information listed related to Co-60.category, the only Therefore, this analys' was performed only for Co-60 in gamma gauges. 3 Inferences based on availabl'e members li ce ra ture showed that dose equivalents received by i most likely not exceed 500 mrem /yr (0.005 Sv/yr)of the genera in most cases.
| |
| l l
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| l I
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| l l
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| I 1
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| I l
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| l l
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| l April 13, 1987 121
| |
| | |
| Re ferences Andrews, Could A.
| |
| 1963. Mexican Co 60 Radiation Accident. _ Isotopes and Radiation Technology. Vol. 1, No. 2, pp. 200-201-Brekke, David D., Robert R. Landolt, and Neil J. Zimmerman, 1985.
| |
| Measurements of Effluent Radioactivity During the Incineration of Carcasses Containing Radioactive Microspheres. Health Physics, Vol. 48, No.3, pp. 339-341.
| |
| Bush, D. and R.S. Hundal, 1973. The Fate of Radioactive Materials Burnt
| |
| .in an Institutional Incinerator. Health Physics, Vol. 24, No. 5, pp.
| |
| 564-568.
| |
| Classic, Kelly, Gary Cross, Radionuclides in Ash from the Incineration of Animals. Healthand R Physics, Vol. 49, No. 6, pp. 1270-1271.
| |
| CFR 1986. Code of Federal Regulations. Energy. Title 10. U.S.
| |
| Government Printing Office.
| |
| Cross W.
| |
| Beta-Ray H. Ing, Dose N. Freedman, and J. Mainville, 1982. Tables of Distributions in Water. Air, and Other Media.
| |
| Chalk River Nuclear Laboratories, AECL-7617, Chalk River, Ontario.
| |
| IAEA 1973 Safe Handling of Radionuclides.
| |
| International Atomic Energy Agency Safety Series No. 1. IAEA, Vienna.
| |
| ICRP 1978. _ Limits for Intakes of Radionuclides by Workers.
| |
| International Commission on Radiological Protection Report No. 30, Parts 1-3 with supplements. Pergamon. Press, New York, NY.
| |
| Jacobs 1968. Sources of Tritium and Its Behavior Upon Release to the Environment.
| |
| U.S. Atomic Energy Commission, Division of Technical Information Extension, Oak Ridge, Tennessee.
| |
| Kocher, D.C. and K.F. Eckerman, 1986. Electron Dose-Rate Conversion Factors For External Exposure of the Skin From Uniformly Deposited Activity on the Body Surface. Draft. Research sponsored by DOE.
| |
| Lubenau, Joel 0.,
| |
| and Donald A. Nussbaumer 1986. Radioactive Contamination 4, pp. 409-425. of Manufactured Products. Health Physics, Vol. 51, No.
| |
| Metal Statistics 1986.
| |
| Fairchild Publications, New York, NY. Metal Statistics 1986. American Metal Mar NCRP 1971.
| |
| Basic Radiation Protection Criteria. National Council on Radiation MD, Protection and Measurements Report No. 39. NCRP, Bethesda, NCRP 1972. Protection Against Radiation from Brachytherapy Sources.
| |
| National Council on Radiation Protection and Measurements Report
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| ? April 10, 1987 122
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| | |
| No. 40. NCRP, B2thesda, MD.
| |
| NCRP 1977. Radiation Miscellaneous Sources. Exposure From Consumer Products and Measuremen,ts Report No. National Council on Radiation Protection and
| |
| : 56. NCRP, Washington, D.C.
| |
| NCRP 1981. Managemert of Persons Accidentally Contaminated with Radionuclides.
| |
| Measurements Report No.65. National Council on Radiation Protection and ,
| |
| NCRP, Bethesda, MD.
| |
| NRC 1983 AgreementThe United States Nuclear Regulatory Commission and the States.
| |
| Regulatory Commission, Washington, D.C. Licensing Statistics and Other NRC 1987. Steve document, draft. Baggett, Nuclear Regulatory Commission. Internal NUREG 1980. Environment Assessment Radioactive Material. of Consumer Products Containing K. Nicholaw, J. D.W. Buckley, R. Belanger, P. Martin, Swenson, of Science Applications, Inc. Prepared for the ,
| |
| U.S. Nuclear Regulatory Commission, Washington, D.C. NUREG/CR-1775.
| |
| NUREG 1985.
| |
| Radioactive Materials Into the United States.The Q. Kwan, E. Vierzba, A.
| |
| Feasibility of De R. Bee, J. Gordon, U.S. Nuclear Regulatory Commission.Wallo, Aerospace Corporation. . Prepared fo NUREG/CR-4357.
| |
| NUREG 1986.
| |
| Incident. The Auburn Steel Company Radioactive Contamination F. Bradley, L. Cabasino, R. Kelly, A. Awai, G. Kasyk, U.S. Nuclear Regulatory Commission, Washington, D.C. NUREG-ll88, Peters, Hans-Jurgen 1985.
| |
| The Hazards of Scrap. North American 1
| |
| Scrap, A Supplement to Metal Bulletin Monthly. March, pp. 113-117.
| |
| Ricks, Robert C.,
| |
| Response William L. Beck, and James D. Berger, 1981.
| |
| to Radioactive Materials Transport Accidents. Dept. of Transportion, p. 10/6. DOT /RSPA/MTB-79/8. l Smith, L.R. , P. A. Sullivan, J. Laferriere, E. Cumming, and D. Demis 1983.
| |
| 2.85 Intake uCi and subsequent fate of a ceramic particle containing of Am-241: c case study. Health Phys 44(4):329-334-Unger, L.M. and D.K. Trubey, 1981. Specific Gamma-Ray Dose Constants for Nuclides Important to Dosimetry and Radiological Assessment.
| |
| Oak Ridge National Laboratory, ORNL/RSIC-45, Oak Ridge, Tennessee.
| |
| USDHEU 1970. Radiological Health Handbook. United States Department of Health, Education, and Welfare, Rockville, Maryland.
| |
| UNSCEAR 1982. Ionizing Radiation:
| |
| Sources and Biological Effects. i United Nations Radiations. Scientific United Nations, Committee New York, NY. on the Effects of Atomic J April 10, 1987 123
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| | |
| Webb, C. , B. Wilkins, A. Wrixon 1975. Assessment of the hazard to the public from ceramic anti-static brushes containine. Po-210 in the form of microspheres.
| |
| Board, U.K. NRPB R36, National Radiological Protection Wilmot, Edwi'n L.
| |
| Commercial Spent1981. Transportation Accident Scenarios for Fuel. Sandia National Laboratories, Albuquerque, New Mexico. SAND 80-2124. 1 I
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| I l
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| l l
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| I I
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| .. April 10, 1987 124
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| _. . -. ..__ ~ ~ . - - . - - - . . . - . - ~ . . - - .. - - . -- - .-
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| t APPENDIX. DESCRIPTIONS. !
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| l This. Appendix !
| |
| considered contains brief descriptions of the devices-mentioned. in this report with typical nuclides and activities !
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| l.
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| A-1 STATIC ELIMINATORS: ,
| |
| 31.3) HAND-HELD / PORTABLE /SMALL BRUSHES (10 CFR t i
| |
| The 3MinCompany " Static Master" devices using Po-210 (138 day half-life) I have 'no salvagemicrosphere value. encapsulation are small size, low cost, and
| |
| ' dislodged from the Studies have shown that microspheres can be i foil source, but normally would not create an internal hazard unless microspheres are damaged. Activities when j distributed are r up to 0.50 mci (18.5 MBq). Devices are supposed to be returned after a year when activity is less than 0.080 mci (3.0 MBq).
| |
| A-2 STATIC ELIMINATORS OR DETECTORS:
| |
| IN EQUIPMENT OR PROCESS LINE j (VERY HICH TOXICITY SOURCES) (10 CFR 31.5) '
| |
| in equipment or on production lines (conveyer in air ducts, roller systems) belts,These or io etc. !
| |
| charge. Detectors are used to sense and measure static !
| |
| (3700 MBq), Typical nuclides and activities include Po-210, up to 100 mci Am-241, 0.0005 !
| |
| (0.0185 MBq). mci (0.0185 MBq), and Ra-226, 0.0005 mci '
| |
| i A-3 STATIC ELIMINATORS OR DETECTORS:
| |
| IN EQUIPMENT OR ON PROCESS LINE
| |
| .(LOW T0XICITY SOURCES) (10 CFR 31.5)
| |
| Same description as A-2 but H-3, 250 mci (9250 MBq) and Kr-85, 2 mci (74 MBq). typical nuclides and activ B CAMMA CAUCES: (10 CFR 31.5)
| |
| MBq) For arethis included. study these devices with activities less than 20 mci (740 Normally these are permanently mounted on the process line to measure thickness, level, ;
| |
| potential for density, etc.
| |
| damage in the normal work environment. Common nuclides There is with typical activitfes are: Co-60, 10 mci (370 MBq), Cs-137, 20 mci (740 MBq), Am-241, 20 mci (740 MBq), and Ra-226, 10 mci (370 MBq).
| |
| C-1 BETA CAUCES: BACKSCATTER TYPE. (10 CFR 31.5)
| |
| These !
| |
| devices are widely used in. production facilities to monitor process lines or measure thickness, density, composition, etc. Devices }
| |
| are sometimes There is permanently mounted but are of ten portable and hand-held. ;
| |
| nuclides potential for damage in the normal work environment. Common with typical activities are: )
| |
| , T1-204, 0.10 mci (3.7 MBq), Sr-90, 0.025 mci (0.925 MBq),
| |
| { 0.050 (1.85 MBq), C-14, 0.050 mci (1.85 MBq), Ru-106, 0.025 mci (0.925 MBq), Pm-147, i
| |
| and Pb-210, 0.010 mci
| |
| ~4
| |
| ; April 10, 1987
| |
| ! A.1
| |
| | |
| (0.37 MBq).
| |
| C-2 BETA CAUCES:
| |
| TRANSMISSION TYPE (10 CFR 31.5)
| |
| MBq) For are t,hi,s study these devices with activities less than 20 mci (740 included.
| |
| the process line. These Theredevices are typically mounted permanently on environment. Sources consist is potential for damage in normal work to 20 mci (740 MBq). almost exclusively of Sr-90 sources of up D
| |
| CAS CHROMATOCRAPHS (10 CFR 31.5)
| |
| Devices consist of ionization sources in laboratory analytical instruments containing detector Instruments may have cells or electron capture detectors.
| |
| Sr-90 or Ra-226, but interchangable detector cells. Older units used They are subject to there damagearedue no examples to currently in the registry.
| |
| cleaning. chemical environment or during Common nuclides MBq) and H-3, 1000 mci (37 GBq).
| |
| with activities are: Ni-63, 20 mci (740 E-1 s X-RAY FLUORESCENCE ANALYZERS:
| |
| 31.5) VERY HICH T0XICITY SOURCES (10 CFR i
| |
| in the These analytical instruments using radioactive sources may be used laboratory monitor a productionor are designed to be portable for. field use or to line.
| |
| They are supplied with source installed.
| |
| Separate source / detector modules may be on hand.
| |
| shielded in normal use. Common Source is well Am-241, 30 mC1 (1100 MBq) and Cm-244, 100 nuclides mci (3700with activities include:
| |
| MBq).
| |
| E-2 X RAY FLUORESCENCE ANALYZERS:
| |
| 31.5) MODERATE T0XICITY SOURCES (10 CFR Same Cd 109, 20 mci description as F-1, but common nuclides with activities are:
| |
| (740 MBq) and Fe-55, 100 mci (3700 MBq).
| |
| F SOURCES FOR CHECKING DETECTOR OPERATION OR CALIBRATION ANALYTICAL REFERENCE SOURCES (10 CFR 31. 5)
| |
| These are used tosmall sources, often supplied by the detector manu'facturer, check performance in the field or as, calibration or analytical standards.
| |
| or theft. Their small size makes them susceptible to loss Source could be damaged in work environment. Typical nuclides with activities are:
| |
| 0.10 mci (3.7 MBq), Ra-226, 0.004 mci (0.15 MBq), Cs-137, l (0.037 MBq). Co 60, 0.01 mci (0.37 MBq) and Sr-90, 0.001 mci l
| |
| 0-1 1 SELF-LUMINOUS DEVICES (10 CFR 31.5) 1 sources,Lighted warning safety signs markers, which include exit signs, emergency light and !
| |
| included are light wands are in this class. Also photometric standards. very low activity lighted gunsights or reference l Common nuclides with activities in are: H-3, 5000 mci (185 GBq), Kr-85, 1700 mci (62.9 CBq), and C-14, 0.10 mci (3.7 0 April 10, 1987 A.2
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| | |
| MBq).
| |
| C-2 SELF-LUMINOUS DEVICES IN AIRCRAFT (10 CFR 31. 7)
| |
| Lighted signs and safety markers are permanently mounted to minimize .the ft or vandalism.
| |
| (185 CBq) of activity and Pm-147 wit).Nuclides include H-3 with up to 5000 mci an activity of 300 mci (11 GBq).
| |
| H ANALYTICAL SOURCES (10 CFRINSTRUMENTS 31.5) CONTAINING SHALL CALIBRATION Cs-137,These 0.040are usually mci liquid scintillation counters with built in (1.5 MBq) or Ni 63, 15 mci (555 MBq) reference sources but detectors. may also include a few mass spectrometers or ion mobility well shielded. Devices are used in an analytical laboratory. Sources are I
| |
| SOURCES FOR CHECKING DETECTOR OPERATION OR CALIBRATION ANALYTICAL REFERENCE SOURCES (10 CFR 31.8)
| |
| Same description as class F, but these include Am-241, 0.005 mci (0.185 MBq).
| |
| J SHALL QUANTITIES OF SOURCE MATERIAL (10 CFR 40.22)
| |
| Only standards for twologging items are in the registry and both represent calibration tools.
| |
| Nuclides and activities include: Th-232, 0.005 mci (0.185 MBq) and natural uranium, 0.005 mci (0.185 MBq).
| |
| K SOURCES FOR CHECKING DETECTOR OPERATION OR CALIBRATION AND ANALYTICAL REFERENCE SOURCES (10 CFR 70.19)
| |
| Same descritpion (0.185 MBq). as class F, but these include Pu-239, 0.005 mci
| |
| .; April 10, 1987 A.3
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| 4 '
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| D
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| \(
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| s' s
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| % ICF INCOR POR ATED
| |
| .' , , . . ( . . . -< ) 9 l
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| MEMORANDUM DECEMBER 16, 1990 TO: Stephen Baggett, NMSS/NRC FRON: Craig Dean RE: Draft Report on Survey of 10 CFR 531.5 General Licensees Enclosed is a draft of the subject report. It represents the preliminary results of the survey effort, in several respects:
| |
| i
| |
| * We have not yet prepared a summary of the report or final '
| |
| conclusions and recommendations. Several additional data analysis tasks must be completed before we can feel confident of what the final " message" will be, and we will also want to discuss those findings with you. We are still collating information, such as l corrections to data taken from quarterly reports on types of I devices held by survey respondents, that i:: not easily amenable to simple printouts. Similarly, we are still culling information' that was presented in written supplements to the questionnaires, in letters, etc. Some of this may change slightly the picture presented from the tabulated yes/no answers.
| |
| * Several data tables that will eventually form appendices to the report still contain numerous duplications, due to different ways 1 of spelling firm names, a'iresses, etc., and have not been included in this draft. We have given the general sense of such l data in the text, and will supply you with printouts as soon as I the cleanup of the data is completad.
| |
| * Estimates of the cost per survey respondent will be affected by the on-site visits that we are currently conducting. In addition, i this draft reports data that are about 7 to 10 days old. Instead of calculating the cost per survey for this draft, we will provide you with a memorandum in the near future and will incorporate i' final cost information into the final draft.
| |
| l We are looking forward to your comments on this preliminary discussion l of results. A number of additional analytic points may occur to you as you read this draft. Please let me know at your convenience of any changes or additions that you would like us to make in future drafts. :
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| @tM T gp. Mi
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| : .'' l e, :
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| ie !
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| 4 I. I
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| ! DRAFT REPORT ON SURVEY OF l
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| ) GENERAL LICENSEES LWDER !
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| 1 10 CFR 31.5 l i
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| l l
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| ]
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| 1 1
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| j Prepared by C.M. Dean, M. S. Lawrence, H. Lester
| |
| )i ICF Incorporated December 1990 l
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| i i
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| i l Prepared for U.S. Nuclear Regulatory Commission nz w sv is n e c.
| |
| | |
| s' NOTICE This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, c; maiumes an legal liabil.ty or respc:sibility F any third party's use, or the results of such use, of any information, spparatus, product, or process disclosed in this report, or represents that its use by any such third party would not infringe privately owned rights.
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| DRAFT REPORT ON SURVEY OF l GENERAL LICENSEES UNDER l 10 CFR 31.5 1
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| 4 i
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| Manuscript Completed: Dece:nbar 1990 Prepared by C.M. Dean, M.S. Lawrence, H. 14 ster ICF Incorporated Fairfax, Virginia j
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| 4 Prepared for Division of Licensing j
| |
| Office of Nuclear Material Safety and Safeguards '
| |
| U.S. Nuclear Regulatory Commission Washington, D.C. 20555 NRC FIN D 2554-0 I
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| 1 l
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| _. _ _ _ _ . _ _ . _ . - - - ~. .__
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| l 4 . ,
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| s.
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| AESTRACT The NRC for several years has been studying the regulatory framework for
| |
| * the licensing of the possession and use of certain measuring and gauging devices containing nuclear materials. In particular, the NRC has evaluated reports of incidents involving devices that are held under general licenses.
| |
| ThL report describes the procedures that were followed and the results that i
| |
| were obtained from a survey of general licensees under 10 CFR $31.5. A sample of general licensees from all non-Agreement States for three categories of devices -- gauges, analytic instruments, and self-powered exit lights -- was contacted by mail with a questionnaire designed to obtain information about the respondents' knowledge of the regulatory requiraments for general licensees and their practices and procedures concerning maintenance, testing, and disposition of the generally licensed devices. The response rates for the survey were between 85 and 95 percent, depending on the type of device.
| |
| Although a high proportion of the general licensees displayed knowledge of the regulatory requirements and compliance with them, a significant number of l discrepancies were identified.
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| 4 4
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| ==SUMMARY==
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| b
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| ) [To be provided]
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| I l l
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| _ ..- ._ =_. __. . ._ ~ _ _ . __ _ _ _ _ _ _ . - _ _ _ _ ._
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| . D :
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| !r . !
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| a 4
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| CON 1ENIS i
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| l l
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| i l
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| 4 ABSTRACT............................................................... iii .
| |
| . SUltlARY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . [ To be p rovided]
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| I i
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| | |
| ==1.0 INTRODUCTION==
| |
| ..................................................... 1
| |
| , l i 1.1 PURPOSE OF REP 0RT.......................................... 1 i i
| |
| i
| |
| | |
| ==1.2 BACKGROUND==
| |
| ................................................. 1 I 4
| |
| 1 1.3 DATA S0URCES............................................... 2 1.4 LIMITATIONS................................................ 2 1
| |
| 4 1.5 SC0PE...................................................... 3 j
| |
| 2.0 SURVEY DESIGN AND QUESTIONNAIRES................................. 4 i
| |
| , i 1 j 4
| |
| 2.1 SURVEY POPUIATION, FRAME , ' AND SAMPLE . . . . . . . . . . . . . . . . . . . . . . . 4 l 2.2 SURVEY QUESTIONNAIRES...................................... 10 ' i 3.0 SURVEY IMPLEMElfrATI0N............................................ 12 i ;
| |
| 3.1 PREPARATION AND NAILING OF SURVEY QUESTIONNAIRES................................... 12 a
| |
| 3.2 TRACKING PROCEDURES........................................ 12 .
| |
| 1 i i 3.3 DATA COLLECTION AND MAINTENANCE PROCEDURES................. 13
| |
| ) 3.4 POLIDWUP ACTIVITIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 h
| |
| 2 4.0 RESULTS.......................................................... 18 i- !
| |
| 4.1 RESPONSE RATES BY CATEGORY OF LICENSEE
| |
| ; AND BY NRC REGI0N.......................................... 18 I 4.2 SURVEY FINDINGS: AWARENESS OF REGULATORY ;
| |
| ; REQUIREMENTS AND COMPLIANCE WITH 10 CFR $31.5.............. 24 !
| |
| 4.2.1 GENERAL LICENSEES FOR ANALYTIC DEVICES...............
| |
| 25
| |
| + 4.2.2 GENERAL LICENSEES FOR GAUGES......................... 33 j i 4.2.3 GENERAL LICENSEES FOR TRITIUM-POWERED 4
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| 1 EKIT.,%
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| e : -
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| SIGNS........................................... 41
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| )
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| 4 5
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| (
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| 1 a
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| 9
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| f f
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| CONTENTS (CONTINUED)
| |
| | |
| ==5.0 CONCLUSION==
| |
| S AND RECONNENDATIONS. ......... be provided)
| |
| .........[To APPENDIX A: SURVEY QUESTIONNAIRES APPENDIX 5: SURVEY COVER LETTER APPENDIX C: RESPONSES TO COMMONLY-ASKED HOTLINE QUESTIONS APPENDIX D: DATA TABLES [To be provided]
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| 4
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| 1 5
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| ;e
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| )
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| | |
| ==1.0 INTRODUCTION==
| |
| | |
| 1.1 Purpose of Report This report describes the procedures that were followed and the results that were obtained from a mail survey conducted for the Office of Nuclear Materials 3afety and Safeguards (NMSS) by ICF Incorporated (ICF) during 1990.
| |
| The survey involved NRC general licensees under 10 CFR Part 31.5. A sample of general licensees from all non-Agreement States for three categories of devices containing nuclear materials licensed under Section 31.5 -- gauges, analytic devices, and self-powered exit signs -- was contacted by mail with a j questionnaire designed to obtain informatica about the respondents' knowledge i of the regulatory requirements for general licensees and their practices and procedures concerning maintenance, testing, and disposition of the generally licensed devices. This report presents the results of the survey.
| |
| | |
| ===1.2 Background===
| |
| NRC has obtained information in the past about situations in which the i
| |
| regulatory requirements pertaining to general licensed materials have not been satisfied.
| |
| This information has come primarily from inspection findings and l
| |
| from reports from general licensees of incidents involving devices containing i
| |
| radioactive isotopes. Such situations have included failures to conduct leakage tests and tests of the device on-off mechanisn and indicator, if any, at the specified intervals; failure to maintain required records; failure to 4
| |
| comply with reporting requirements; failure to comply with the requirements concerning transfers and custody of devices; and failure to comply with labeling requirements. Since about 1982, NRC has been collecting information ,
| |
| and conducting studies of the general license regulatory framework. Between 1984 and 1986, NRC conducted an extensive study and prepared a report on this subj ect. The NRC report concluded that there were several areas of safety concern due to inadequate accountabil'ty of the devices and frequent lack of awareness on the part of the users of the contents and requirements of the i
| |
| : pertinent regulations, particularly the rules on transfers, disposal, and )
| |
| i recordkeeping.
| |
| In 1989, NRC contracted with ICF Incorporated to conduct another study of general licensees in the form of a mail survey. The survey was undertaken to accomplish at least two goals:
| |
| (1) To obtain additional information about current general licensees.
| |
| Survey respondents were firms that had obtained one or more 1
| |
| devices during.the period 1985 to 1990.
| |
| (2) To investigate whether mail survey techniques could.usefully support NRC in informing gevaral license:es of their responsibilities and deterr.ining the extent to which general 1!.bensees know, understand, and comply with the requirements in 10 CF2 ! 31.5.
| |
| | |
| 4 2
| |
| 1.3 Data sources The survey frame was orovided to ICF by NRC from quarterly reports l
| |
| eibnitted by specific licensees dur..ig the period iron about 1985 to early 1900. '
| |
| These period.
| |
| the reporting quarterly reports list firms to which devices are shipped during 1
| |
| In addition to the firm ncemos and addresses of general licensees provided on quarterly reports, informatian about members of the survey frame was also obtained by ICF through telephone calls and through library research. l This information was developed after the initial distribution of the survey, when it became clear that certain names and/or addresses taken from the quarterly reports were no longer correct. ICF also conducted telephone I followup to determine the correct name of the person who was serving as the radiation safety officer or other firm contact for the device, in cases when the name and address of the firm appeared to be correct but no sur;vey response d
| |
| had been obtained. Finally, survey respondents provided addit!nnal
| |
| ; information on survey forms about firm names, addresses, and contacts.
| |
| Information about the devices held by the general licensees was obtained initially from quarterly reports. The survey questionnaire requested that such information be verified and corrected if necessary by the respondent, and general licensees themselves provided information about the types and serial numbers of devices in their possession through survey responses. In a number of cases, respondents did correct the information on the questionnaire. t l.4 Limitations Quarterly reports did not always contain complete correct addresses for firms to which devices were shipped. In addition, the reports only identified the firm to which a device was shipped initially; if the firm then transhipped the device the quarterly report did not indicate the ultimate recipient of the i
| |
| device. Finally, quarterly reports were in some cases unclear about whether they were describing devices that had been shipped to respondents or shipped back from respondents to specific licensees. In a number of cases, respondents reported that the device listed on the questionnaire had been
| |
| ; returned to the vendor.
| |
| The list of potential respondents initially furnished by NRC did not
| |
| , include the names of 500 exit si 5n licensees, 1000 analytic device licensees, and 1500 gauge licensees, which was the goal for the final size of the survey.
| |
| A supplementary sample therefore was chosen. Some members of the supplementary sample were chosen in error ' rom Agreement States, and at the request of NRC those potential respondents were not followed up or included in the results. The fin &1 sample size was reduced by subtraction of these potential respondents.
| |
| Some potential respondents were vendors or electrical contractors who had shipped the devices to ultimate users. In some cases the vendor or contractor had no information about the ultimate destination and/or current
| |
| | |
| s 4 /
| |
| 3 condition of the device. In other cases, Ehe radiation safety officer was not familiar with all of the devices listed for a particular organization. This i
| |
| was particularly a problem eith large organization- such as universities, i with oiffuse organization structures and rapid turnover of personnel.
| |
| j Some potential respondents were impossible to locate, either because they were out of business, had left to forwarding address, or for other l reasons could nor be traced. A number of responses were received in which the 1 respondent reported that the firm had been sold, entered bankruptcy, or had closed, and no remaining personnel were familiar with the disposition of the device.
| |
| i A small group of potential respondents refused to cooperate. In general, however, response problems occurred due to lack of information rather than unwillingness to supply information.
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| 1.5 Scope This report discusses the following:
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| : 1. Surve,' Design and Questionnaires (Chapter 2.0) I
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| : 11. Survey Implementation (Chapter 3.0) iii. Findings of the study with respect to the general licensees '
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| authorized in 10 CFR 531.5 (Chapter 4.0); and iv. Overall general conclusions and recommendations (Chapter 5.0). ;
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| | |
| ,u *
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| , 2.0 SURYEY DESIGN AND QUESTIONNAIRES '
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| This section describes the initial design of the survey and the design and contents of the survey questionnaires. It first discusses the national population of general licensees, the survey frame (i.e., those members of the pc,.lstion from which the survey sample was chosen),, and the survey sample.
| |
| Next it discusses the development and contents of the survey questionnaires, including how the questions were chosen and how the questionnaires were tailored to reflect the different characteristics of the different types of general licensees included in the sample.
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| 2.1 Survey Population, Frame, and Sample The Office of Nuclear Materials Safety and Safeguards is responsible for administering the NRC's programs concerning the possession and use of nuclear materials. Licenses for possession and use of byproduct material constitute approximately 5,000 of the approximately 8,000 materials licenses currently administered by NRC (the 29 Agreement States administer another 16,000 licenses).1 As 10 CFR $30.31 provides, che NRC's licenses for byproduct materials can take one of two forms: (1) the license can be a specific license issued to individually-named persons or organizations upon applications filed pursuant to the regulations in Part'30 and Parts 32 through 35 of 10 CFR, or (2) the license can be a general license, effective without the filing of an application with the Commission or the issuance of licensing documents to a particular person. As the 1989 NRC Annual Report explains:
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| General licensees include individuals or organizations that ;
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| become licansees (without contacting the NRC) when they receive a byproduct source or a device containing a byproduct source from a specific licensee. These include certain measuring, gauging, illuminating, and controlling devices containing byproduct material with radioactivity ranging from microcuries to several curies. . . . General licensees are expected to be able to use the devices safely by following simple instructions -- without having radiological safety training or experience -- because safety is built into the devices.
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| , Altogether, the population of general licensees numbers about 30,000 in non-Agreement States, using abcut 400,000 devices, with about 60,000 general licensees in Agreement States with about 800,000 devices.
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| NRC currently has issued about 17 general licenses. Seven pertain to the transportation, storage, and installation of radioactive material; ten allow the use of particular devices or materials. The survey frame was comprised only of general lice,nsees under the general license established by 10 CFR $31.5.
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| 1 United States Nuclear Regulatory Commission,1989 Annual Reoort, NUREG-1145. Vol. 6, July, 1990, p. 74
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| l 5
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| Section 31.5 governs:
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| : I
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| { Any byproduct material contained in devices designed and manufactured for the purpose of detecting, j measuring, gauging, or controlling thickness, density, j 4
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| level, interface location, radiation, leakage, or 1 j
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| qualitecive or quant.itative chemical composition, or !
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| for producin light or an ionized atmosphere. l The identity of such general licensees is reported to NRC by the l t
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| specifically licensed manufacturers of the devices containing the byproduct '
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| material.
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| Manufacturers specifically licensed by NRC provide quarterly reports to NRC; Agreement State manufacturers or distributors must notify NRC !
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| when they sell generally licensed devices in non-Agreement States. The sampling frame for the survey was comprised of quarterly reports submitted by l 3
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| the manufacturers listed in Exhibits 2-1, 2-2 and 2-3, who produce a large !
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| 4 proportion of all such devices, for the periods indicated in the exhibits. l 4
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| The potential survey respondents were selected by NRC from the quarterly !
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| j reports submitted by specific licensees. The quarterly reports are provided I to NRC as proprietary business information under a claim of confidentiality, l
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| * Therefore, the names of potential survey respondents chosen from those reports were treated as confidential information, and maintained under conditions of security, both in hard copy form nd in computer records. Names of individual y
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| )
| |
| respondants are not included in this report. ,
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| Separate samples were selected for each of the three major categories of respondents, with the goal of identifying samples of the following sizes:
| |
| Cauges 1500 respondents l Analytic Devices 1000 respondents 1
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| Self-powered lights 500 respondents.
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| l Initial review of the selected potential respondents revealed several factors that required a supplemental sample to be selected. These factors
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| ; included:
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| j
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| * Duplication of firas; Missing addresses that could not be obtained in a timely manner; Other factors indicating that a potential respondent would not be J
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| appropriate, inclu m g the following:
| |
| T (1) The possibility that the devices were installed on aircraft, and therefore correctly reported to NRC under section 31.xx of 10 CFR. In addition, because such devices would present particular problems of tracking and identification, they vere removed from the sample.
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| 1
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| 9 6
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| 1 4
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| EERISIT 2-1 l SPECIFIC LICENSEES SUFFLYING DEVICES TO
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| ; GAUGE RESPONDENTS Firn Name Dates Covered
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| ; Combustion Engineering 1989-1990 5
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| Ohmart 1985, 1989-1990 Rosemount 1989 Industrial Dynamics 1987-1989 1
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| Sentrol 1985-1989 MPSI Technologies 1985-1988
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| ; Molins Richmond, Inc. 1989 4
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| Delphi. Instruments 1988-1990 AEG-Telefunken 1985
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| ; RMD, Inc.
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| 1989
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| ! Heuft 1989-1990
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| , Honeywell 1988 i
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| ' AccuRay 1985-1988 Texas Nuc. lear 1985-1988 Kay Ray 1985-1989 Berthold Systems 1989-1990 Intergrated Industrial Systems 1989 i Harrel 1989 j United Technologies 3 1988-1989 '
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| Ahlstrom 1988-1989 CMI International 1987-1990 1 Ronan 1989 NDC Systems 1987-1989 Fife 1985-1986 l Dosimeter Corporation 1985-1990 Victoreen 1989
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| ; Fairchild Weston 1985-1988
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| ; Data Measurement Corporation
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| ' 1988-1989 Aeonic Systems 1988-1990 Gamma Instruments 1988-1989 Peco Controls 1988-1990 Loral Control Systems 1989
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| ! Measurex 1985-1986 Barber Colman 1986-1989
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| : TCI 1988-1990 UPA Technology 1985
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| i . ;
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| l i . l I
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| l l 7 j
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| EERIBIT 2-2 SPECIFIC LICENSEES SUPPLYING DEVICES T0 i
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| ANALYTIC DEVICE RESPONDENTS Firm Name Dates Covered NRD 1989 Chrompack 1988 Fischer Technology 1988-1989 i Beckman 1989-1990 i Perkin-Elmer 1987-1990 l Varian 1988-1989 Packard Instrument 1989-1990 Hewlett-Packard 1989-1990 TSI 1989 Vesco 1988 Panametrics 1988-1990 PCT 1988-1989 Shimadzu 1989 CSI 1985-1989 ASOMA Instruments 1989 I Valco Instruments 1988-1989 l S-Cubed 1988-1989 !
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| ICN Micromedic Systems 1988-1989 l Tracor Instruments 1988-1989 3M Electrical Specialties Division 1988-1990 O
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| Ge
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| | |
| . . _ . - . - -. _ -. .-.- ... - .. .~ - . - - - _ - - - . ..-. ..
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| a i
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| I
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| ; 8 4
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| i j EERIBIT 2-3 l SPECIFIC LICENSEES SUFFLYING DEVICES TO 4
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| ] SELF-F0VERED LIGHTS LICENSEES l j Fira Name Daten covered a
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| Stusser Electric 1989 Self-Powered Lighting 1989
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| ; JMJ Sales 1986-1989 Electric Distributors 1989 i Safety Light 1988-1989 i
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| Isolite 1989 i SRB Technologies 1989 Shielded Source 1989 l
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| I ,
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| i 1 1
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| 1 1
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| i t
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| i
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| 4 1 -
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| 9 (2) The possibility that the devices were installed on military ships, or shipped to military installations for other purposes.
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| * Indications that the number of firms initially identified was not sufficient to reach the goals for the size of the samples, even if other factors had not intervened.
| |
| A significant difference between the quarterly reports for gauges and analytic devices and the reports for self-powered lights was the number of such devices that could be included in a shipment. In most cases, gauges and analytic devices were described individually on the reports, with separate serial numbers or other identifiers. In contrast, the quarterly reports for i self-powered lights sometimes described shipments of several lights of the same type, without other identifying characteristics. A typical report of this type might identify the firm to which the lights were shipped and the date shipped, give a model number, and then indicate that 24 such lights were shipped on the same day.
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| In some cases, potential respondents would have been required to provide
| |
| , information about large numbers of self-powered li5 hts (For one firm, over one j hundred such responses would have been required; others would have been required to make several dozen responses.) Because c h a heavy burden would'
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| ~
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| ! have significantly affected the likelihood of response, an additional sample }
| |
| of potential respondents for self-powered lights was chosen, and some of the previously chosen potential respondents with the largest numbers of devices were replaced by members of the supplemental sample with significantly fewer i devices.
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| ]
| |
| 'The supplemental samples were selected by ICF staff from quarterly
| |
| , reports provided by NRC. Both quarterly reports previously used and additional quarterly reports were used. In some cases, quarterly reports from earlier years (e.g., 1985-1988) were used to supplement the quarterly reports initially used by NRC (which dated primatily from 1989).
| |
| A number of firms located in Agreement States were mistakenly included in the supplemental sample. At the request of NRC, these potential respondents were dropped as soon as they were identified. Because almost all firms included on the quarterly reports were already part of the sample,
| |
| ; however, the firma deleted from the sample could not be replaced.
| |
| The final sample sizes, following both supplementations and deletions, were as follows:
| |
| i Cauges 1178 respondents Analytic Devices 872 respondents Self-powered lights 483 respondents.
| |
| The number of devices attributable to these respondents, however,
| |
| | |
| . l l .-
| |
| 10 equalled or exceeded the goals for the survey.
| |
| Data concerning each respondent were entered into one of three databases, depending on the respondent type. Data included firm name and address; type of device, including serial number and amount and type of radionuclide, if available; and name and address of the vendor of the device.
| |
| Data entry began in early May 1990. In addition, data were entered into an AlmPlus database for use in addressing and managing the survey mailings.
| |
| 2.2 Survey Questionnaires Survey questionnaires were designed to include both fira-specific and device-specific information, which was to be verified by the respondent, and questions to be answered by the respondent.
| |
| 'The questions included in the survey were of two types. Questions were based on the set of questions used by NRC in its study conducted between 1984 and 1986. In most cases, those questions could be answered by a "yes" or "no" response. When additional information, of a supplementary nature, was necessary, the questionnaire was designed to include space and instructions for narrative responses. Finally, " skip patterns" were included to ensere that related sets of questions were addressed properly. Copies of the questionnaires for gauges and analytic devices, and for self-powered lights are included in Appendix A.
| |
| For each respondent, a single fira-specific set of questions was prepared, with preprinted information on the firm name, address, contact person, phone number, and principal business. This section of the questionnaire also contained questions about knowledge of the general license conditions. Each respondent was assigned a separate respondent identification number, which was printed on the questionnaire forms.
| |
| For respondents with gauges and/or analytic devices, a second device-specific set of questions was prepared, with preprinted information on the device, for each device indicated on the cuarterly reoorts to be in the nossession of the reneral licensee. Thus, a licensee with a single gaugs would receive one device-specific questionnaire; a licensee with three gauges would receive three such questionnaires. When possible, each device was identified by type and by serial number. When such identifiers could not be obtained from the quarterly reports, separate devices ware identified by number (i.e., one of one, one of two, etc.).
| |
| For respondents with self-powered lisSts, a different type of questionnaire was prepared. Such respondents received only one device-specific questionnaire, irrespective of the number of lights they had been shipped of the same type. When a respondent had received more than one type of self-powered light, a different device-specific questionnaire was prepared for each type. Each device-specific questionnaire also included an entry giving the number of devices shipped and the date of the shipment.
| |
| | |
| ~
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| i i
| |
| i 11 i In addition to the survey questionnaires, a cover letter was prepared i
| |
| for inclusion in the survey packages. Following review by NRC, the letter was printed on NRC letterhead. The cover letter explained the purpose of the rarvey, gave general instructions for .esponding .. the survey, explained the confidential nature of responses, and encouraged all respondents to return the survey. A copy of the cover memorandum is included in Appendix 3.
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| 1 1
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| 1 1
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| i i
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| l l 1 i
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| i I
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| l 1
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| l !!
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| J i
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| j i
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| d 9
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| l
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| , 3.0 SURVEY IMPLEMENTAT;.c This section describes the steps taken to implement the survey of general licensees.
| |
| It describes how the survey questionnaires were prepared and the contents of the packages that were sent to potential respondents; how r..ponses and nonresponses were trackeu; procedures for data collection and maintenance; and followup activities.
| |
| 1 l l 3.1 Preparation and Mailing of Survey Questionnaires Letters containing questionnaires with proprinted fira-specific and device-specific information for each respondent, a cover letter, a return j envelope, and a copy of 10 CFR $31.5 were prepared under secure conditions in a survey preraration room. Firm and device-specific infonmation was obtained ,
| |
| , from a master database prepared from the contents of the quarterly reports.
| |
| Address information was duplicated in an AlmPlus database, and address labels were printed out from a that database. Address labels were printed in quadruplicate. One was attached to the envelope for each respondent, one to certified mail forms attached to the envelop., one to certified mail return i receipts (" green cards"), and one to the appropriate entry in the respondent tracking notebook. Following preparation of each letter, it was reviewed by a different member of the survey staff as a check for the following:
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| e
| |
| ' to ensure that the respondent name and address matched the mailing labels; to ensure that the firm identification number matched on all {
| |
| ! device-specific questionnaires; e
| |
| to ensure that proper number of device-specific questionnaires were included; and e
| |
| to ensure that the package was complete.
| |
| An entry was made for each questionnaire package in a questionnaire tracking notebook, organized by respondent number, at the time the package was prepared. A copy of the appropriate mailing label was placed in the notebook under the correct respondent identification number. The date the package was checked, the identity of the checker, and the date the package was mailed were all entered as those actions took place. .,
| |
| Mailing of questionnaires was done in three groups, by type of respondent, beginning in mid-June. The first group of surveys to be mailed was self-powered lights, followed by gauges and then by analytic devices. The group approach was adopted to enable the printing of the survey forms and the preparation of the questionnaires ano survey packages to be spaced more evenly over time.
| |
| 3.2 Tracking Procedures Survey questionnaires were tracked in respondent notebooks and in the AlmPlus databases. The notebooks and databases were retained in the survey
| |
| | |
| r 13 preparation room. As certified mail receipts were received, the fact that the receipt had been sent back was entered into the respondent notebook and the )
| |
| . receipts were filed by survey respondent identification number. This procedure enabled survey staff to verity whether the survey package had been j
| |
| , accepted at the address to which it was sent, and by whom. i 1
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| A number of survey packages were returned as undeliverable. That fact 1 was entered into the AlmPlus survey tracking database. Special codes were used to indicate the reason for return (e.g., moved, left no forwarding !
| |
| address; no such address; addressee unknown, etc.) or action taken (e.g., l i
| |
| address changed to new address and questionnaire remailed).
| |
| l 3.3 Data Collection and Maintenance Procedures i As completed survey responses were received, they were opened in the l survey preparation room and logged in to the appropriate respondent notebook l at the entry corresponding to that survey package, LJ indicated by the respondent identification number. Periodically thereafter, survey responses j were entered in the appropriate survey results database. In addition, an '
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| entry indicating that the survey questionnaire had been returned was made in
| |
| , the AlmPlus survey tracking database. All returned survey questionnaires were filed by respondent number and retained in the survey prepai stion room.
| |
| The survey cover letter and the return envelopes both requested respondents to return completed surveys to the address of the survey {4l contractor, in Vienna, Virginia. In some cases, however, questionnaires that I could not be delivered to the respondent and/or completed questionnaires were mailed instead to the NRC Headquarters building in Rockville, Maryland. In such cases, the questionnaires were periodically transferred by messenger to ICF.
| |
| All survey response data were double key entered into the appropriate databases. The databases were maintained on computers located in the secure ,
| |
| survey preparation room and were backed up on a daily basis. A separate copy I of the databases was also maintained on a Bernoulli disk.
| |
| Certain respondents included letters with the questionnairt4, or substituted letters for the questionnaires. In such cases, the letter was reviewed by the preparation room staff and either treated as a response or set ;
| |
| aside for special attention. All letters were filed by respondent identification number and retained in the preparation room.
| |
| 3.4 Followup Activitier In addition to the initial mailing, a number of followup activities were undertaken to obtain the highest possible response rate and, whenever possible, to obtain information concerning the ultimate disposition of.the device even if it had been transferred or disposed of by the original general licensee. These followup activities, which are described in the following l subsections, included a wave II mailing, use of a coll-free telephone hotline
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| | |
| i l
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| i 14 to respond to questions by respcndents, telephone followup initiated by ICF, and on-site visits.
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| Vave II Mailing.
| |
| Approximately 40-60 days after completion of the first wave mailing, a second mailing was made to all respondents who had not returned a completed questionnaire. This mailing included a slightly stronger cover letter and copies of all other documents (i.e., a survey form, return envelope, and copy of relevant portions of 10 CFR Part 31) that had been included in the first-wave mailing. This mailing was also sent by certified mail, return receipt requested. The addresses used for the Wave II mailing were corrected, when possible, with the results of the Wave I mailing and from other sources, described in this section.
| |
| Hotline Responses to Inquiries Prior to the Wave I mailing, ICF established a dedicated telephone " hot line" and provided the number for that hotline in the cover memorandum that accompanied all survey questionnaires. This hotline was given a toll free 800 number. Software converted the single hotline into four incoming lines of the ICF telephone system, each of which was attached to a message storage and retrieval unit. ICF assigned staff to clear the " voice message boxes" and respond to incoming telephone calls on a daily basis. ICF sought to respond to telephone calls within one day, although this was not always accomplished, l depending on the volume of incoming calls and the ability to call back and reach the respondent. Frequently, several calls back and forth would be placed before the respondent's questions could be answered. A phone log was kept of all calls placed to the hotline and the disposition of the inquiry.
| |
| Periodic summaries were prepared of questions asked and answered most frequently. Those questions are described in Exhibit 3-1, which provides a sample of questions asked by potential respondents for the tritium exit sign survey. Similar questions were asked by respondents for the other two surveys.
| |
| Telephone Followup ICF performed three types of followup activities by telephone in addition to the hotline service. First, a number of outstanding addresses were checked by telephone; second, as the period of time for responses came to a close, ICF contacted every outstanding potential respondent by telephone and sought to elicit a response from them; finally, in a number of cases, new
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| | |
| 't 15
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| ' EINIBIT 3-1 1 TYPICAL QUESTIONS ASEF.D BY EXIT SIGN RESPONDENTS i
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| , 1. We are a distributor, and do not keep signs in stock but ship them immediately to the customer, are we a general licensee and are we required to fill out the survey?
| |
| : 2. We can't locate the sign--it is somewhere in the
| |
| ; city--are we required to track it down and obtain the information?
| |
| : 3. The proper contact person is out of town, or information is difficult to j find--can we have an extension of time to prepare the questionnaire?
| |
| : 4. What is a copy of the Gener.1 License?
| |
| l' S. We just took over this facility, and the signs were installed by a contractor; are we required to maintain records and supply information about them?
| |
| : 6. Is the label clearly visible if it is 1 1/4" wide?
| |
| : 7. The listed contact person is no longer an employee of the facility.
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| j 4
| |
| Should he/she still fill out the survey? il !
| |
| : 8. We have a broad-scope license, are we still required to fill out the
| |
| ; survey?
| |
| : 9. No serial number is given, and we can't locate which sign (s) we should
| |
| ; be providing information about.
| |
| : 10. The questionnaire must have been sent to us by mistake, we have no records of ever having received any signs.
| |
| : 11. We sent the sign back for disposal; therefore, we no longer have it, do we still have to fill out the questionnaire?
| |
| : 12. Is this survey mandatory?
| |
| | |
| . I I 1 16 I 1
| |
| survey questionnaires were distributed to respondents and returned by them using telephone facsimile equipment. Each of these activities is described below. ,
| |
| Address checking was initiated when a survey package was returned by the postal service as undeliverable because the addressee was not at the indicated
| |
| } address or because of a problem with the address. Calls were made to the
| |
| ! telephone information service for that city (in some cases for several different area codes associcted with the general location) and inquiries were made about the telephone number of the potential respondent. If the l 4
| |
| respondent was listed, an inquiry was also made about the address. In some cases, address information was obtained from the telephone information service. If address information was not available from that source, the
| |
| ! listed number was called and if the firm was reached new address information was obtained from the contact.
| |
| Approximately six weeks before the scheduled end of the survey, a list of all outstanding potential respondents was generated and a comprehensive telephone campaign was initiated to contact each of the firms listed. A j
| |
| protocol was developed to ensure that the person best suited to respond to the i
| |
| questionnaire was eventually contacted. Thus, once the firm itself wa:
| |
| reached, the following persons were requested, in order:
| |
| * Radiation Safety Officer. In general, if a radiation safety i
| |
| ' officer had been designated by the firm, they were knowledgeable !,
| |
| concerning the devices possessed by that firm. In some cases, 3
| |
| however, the radiation safety officer had not received the survey questionnaire because it had gone to another person in the firm
| |
| ) and had not been passed on properly to the Rso. In other cases, j
| |
| the RSO confirmed that the form had been received but that for
| |
| .some reason it had not yet been returned.
| |
| ; e Personnel Officer. If no RSO was available, most firms required all information about personnel and their duties to be obtained from the personnel office.
| |
| t
| |
| . Safety Officer. If no RSO was available, the next person contacted, through the personnel office, was usually the safety officer. Safety officers frequently were aware of devices and responsible for preparing survey responses.
| |
| Facilities / Maintenance Office. If no RSO or SO was available, information about devices could sometimes be obtained from the i
| |
| facilities / maintenance office. In particular, the location of self-powered exit signs was frequently something that those personnel were familiar with or could determine.
| |
| e Management.
| |
| For small firms, cases in which other personnel could j
| |
| not be contacted or were uncooperative, or when referred by other personnel, contacts were sometimes made with firm management.
| |
| | |
| 17 Thsir responses was sometimes that they were personally not familiar with all the facts concerning the devices, but that they would determine the aporopriate member of their staff to respon' to the questionnaire.
| |
| i During the course of the comprehensive followup, information from the same firm concerning responsibilities for preparing a survey response and whether or not such a response had been prepared sometimes varied significantly dspending upon who was contacted.
| |
| j
| |
| ~ The telephone followup frequently contacted a person who would ask for a 4
| |
| survey questionnaire to be resent, stating that the initial form had been lost or discarded. On several occasions, respondents asked that the questionnaire be sent by telephone facsbaile, and promised to respond immediately back by facsimile. This approach was extremely useful; however, it required that new 4
| |
| ' survey questionnaires be printed very quickly so that they could be sent in a timely manner. Although most were sent the same day, some required additional time to prepare.
| |
| ) On-Site Followup A limited number of site visits were performed when survey responses i
| |
| were not received from a significant number of respondents in a particular
| |
| ) geographical area. These site visits were arranged in advance, and in the course of making such arrangements a number of additional survey responses
| |
| , were generated. Survey teams made from 4 to 6 visits per day. In some (
| |
| 4 circumstances even with advance planning the necessary site representatives were not present and the device (particularly exit signs) could not be located by plant personnel. In other cases, filled out survey forms were obtained --
| |
| one of which had been delayed by review by the plant legal staff for about four months. In general, site visits added marginal response rates but were less productive than the other follo..tp activities described above.
| |
| 4 a
| |
| | |
| 4.0 RESULTS ;
| |
| This section presents the results obtained from the mail portion of the i survey of general licensees. It first provides the response rates, by category of licensee. It 'Len deteils the finding = that can be derived fron
| |
| ; the survey responses. Supplementary data are presented in Appendices to this report.
| |
| 1 4.1 Response Rates by Category of Licensee and by NRC Region i,
| |
| The response rates for all categori s of general licensees included in i
| |
| 4 the survey were at or above the rates normally obtained by mail surveys.1 i Exhibit 4-1 presents the response rates, by category of device and by Region, for each of the three broad types of devices included in the survey and each of the five NRC regions. As Exhibit 4-1 indicates, the response rates
| |
| ' nationwide for both analytic device licensees and gauge licensees currently t i exceed 90 percent, as do the response rates by Region for almose all Regions. 1 2
| |
| Response rates nationwide for self-powered light licensees are somewhat lower, at about 80 percent in total and for almost all Regiona.2 4
| |
| In addition to the formal questionnaire responses received, a number of firms responded by letter. Approximately 40 letters must be added to the !
| |
| 1 totals presented in Exhibit 4-1.
| |
| Finally, followup activities are likel,, if response rates continue at about their current rate, to elicit at least 100 t
| |
| additional responses.
| |
| : l
| |
| : 1' 1' Several hypotheses can be advanced for the ~ high response rates. They include: I e
| |
| A st.rong tradition of responsiveness on the part of NRC licensees to Commission and Staff inquiries; l
| |
| 4 1
| |
| 1 D.A. Dillman, Mail and Telenhone Surveys, Wiley-Interscience, 1979, p.
| |
| 51 suggests that response rates for mail surveys are usually lower than for
| |
| ! surveys conducted by telephone or personal interview, ranging from 60 to 75
| |
| ' percent for surveys of the general public using lengthy questionnaires to 85 percent or higher for homogenous samples, 2 l Additional analysis is necessary to determine the number of devices l
| |
| , that are reflected in this response rate. In some cases, the absence of only one respondent can significantly reduce the number of devices being reported upon. (For example, several potential respondents for exit signs submitted letters or survey questionnaires in which they stated that they were electrical supply companies, and therefore had transshipped all of the numerous signs delivered to them. In another case, a potential respondent ,
| |
| advised that the plant to which 58 analytic devices had been shipped was I operated under contract to the US Deparement of Energy, and therefore the questionnaires were not completed.) I
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| )
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| i
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| | |
| i -
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| 4 .
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| . 19 4
| |
| A EXHIBIT 4-1
| |
| [ INTT7IM SURVEY PESPONSE RAT''
| |
| +
| |
| BY CATEGORY OF GENERAL LICENSEE AND BY NRC REGION i
| |
| j i
| |
| Analytic Region 1 Region 2 Region 3 Region 4 Region 5 Total Nonaggreement Received 368 -
| |
| 54 306 47 10 785 Percent 89.101 94.74% 90.27% 90.381 90.91% 90.021 Total 413 57 339 52 11 872 Gauge Region 1 Region 2 Region 3 Region 4 Region 5 Total Nonaggreement Received 373 111 505 74 5 1068 Percent 90.311 90.241 90.66% 93.67% 83.33% 90.66%
| |
| Total 413 123 557 79 9
| |
| 6 1178 Tritium Region 1 Region 2 Region 3 Region 4 Region 5 Total Nonaggreement Received 140 32 163 18 24 377 Percent 77.781 60.001 79.901 81.821 64.861 78.05%
| |
| Total ' 180 40 204 22 37 483 12/08/90
| |
| | |
| 1 20 Statements in the questionnaire cover memorandum and in' followup l telephone calls that nonres onse could result in an on-site inspection of the device; l
| |
| * Intensive mail and telephone followup to the initial survey 4 mailing; l
| |
| )
| |
| * Clear instructions and questionnaires that mada response j relatively easy; ,
| |
| l Availability of a telephone hotline, so that respondents' , i questions could be answered in a timely manner; '
| |
| Assistance provided to general licensees by vendors of the devices; and A relatively long period of time for responses and followup.
| |
| Additional research would be necessary to determine if the response i rates are attributable to these or other factors.
| |
| i
| |
| [ Part IIA of each questionnaire requested the survey respondents to 4
| |
| supply information concerning the principal business conducted at the 3 l
| |
| facility. For analytic device and gauge respondents, 13 specified < )
| |
| altarnatives and an "other" category were provided; self-powered light i respondsats were given 9 alternatives and an "other" category. The descriptiens of principal business activities are summarized in Exhibits 4-2,
| |
| ] 4-3, and 4-4.
| |
| 1 I
| |
| l
| |
| | |
| 's 21 EIRIBIT 4-2 PRINCIPAL BUSINESS AC'IVITIES SPECIPIED BY ANALYTIC DEVICE RESPONDENTS Tyne of Activity Nn=her of Resoondents Percent Agriculture 3 0.4 Canning / Bottling 2 0.3 Chemical 35 5.1 Food 15 2.2 Foundry 1 0.1 Laboratories 274 39.7 Mining 1 0.1 Petroleum 17 2.5 Power-Coal Feed 2 0.3 Pulp and Paper 2 0.3 State / Local Government 17 2.5 Steel 2 0.3 Waste Treatment 14 2.0 Other* 309 44.2
| |
| * The business activities in the "other" category are currently j being tabulated. They include aerospace, electrical products, printed circuit board, instrumentation, medical electrical equipment, automotive parts, and pharmaceutical manufacturing.
| |
| | |
| d i
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| * i 1
| |
| i t
| |
| 4 22 b
| |
| 4 EIRISTT 4-3 PRINCIPAL BUSINESS ACTIVITIES SPECIFIED BY "
| |
| GAUGE RESPONDENTS i
| |
| $ Tyne of Activity Nn=her of Resoondents Percent
| |
| ! Agricultura 0 0.0 l Canning / Bottling 39 4.6 j Chemical 74 8.7 Food 41 4.8 l Foundry 1 0.1 Laboratories 18 2.1 Mining 64 7.5 Petroleum 41 4.8 Power-Coal Feed 21 2.5 Pulp and Paper 138 16.2 State / Local Government 3 0.4 Steel 73 8.6 Waste Treatment 4 0.5 O ther* 333 39.2 l l
| |
| The business activities in the "other" category are currently being tabulated.
| |
| | |
| l j ..
| |
| * i ..
| |
| 1 i
| |
| l 23 l
| |
| 4
| |
| : EIHIBIT 4-4 l 4 '
| |
| PRINCIPAL BUSINESS AC*IVITIES SPECIFIED BY 1
| |
| SELF-POWERED LIGHT RESPONDENTS 1
| |
| il Tyne of Activity Number of Resnondents Percent Aviation 8 0.3 Entertainment 6 1.8 1
| |
| ! Food and Lod5 ing 21 6.4 2
| |
| Hospital 13 4.0 Laboratory 5 1.5 i Manufacturing
| |
| * 66 20.2 l
| |
| Office Building 24 7.3 l g
| |
| School 38 11.6 l l State / Local Government 9 2.8 1
| |
| ; Other* 136 41.6 1
| |
| * The business activities in the " manufacturing" and "other" category are currently being tabulated.
| |
| }
| |
| i k
| |
| i f-d Y
| |
| m
| |
| | |
| i 24 4.2 Survey Findings: Awareness of Regulatory Requirements and Compliance ,
| |
| with 10 CFR 831.5 This section discusses the survey results for each of the three i categories of general licensees included in the survey. The questionnaires i
| |
| were designed to elicit information that paralleled the basic responsibilities i
| |
| and requirements of general licensees, and therefore included questions i
| |
| involving the following:
| |
| * Awareness of rerulations and accountability. Under 10 CFR 1
| |
| 531.5(a) a general license is issued to commercial, educational, and medical institutions, to individuals in the conduct of their business, and to federal, state or local governments to acquire, receive, possess, use or transfer byproduct material contained in the covered devices. The specific licensee who manufactures and i
| |
| distributes the device must supply a c:py of the general license to each person to whom the device containing byproduct material is transferred.
| |
| In practice, the pertinent sections of 10 CFR Part 31 are transferred by specific licensees and constitute notice to i
| |
| the general licensees of the general license terms and conditions.
| |
| The survey questionnaire thus inquired whether the respondents had a copy of the general license, and whether the general licensee had specified a company representative who was responsible for l knowing the general license conditions and ensuring compliance l
| |
| ; with them.
| |
| Conn 11ance with testina and maintenance reouirements. Part 31 requires leak testing for certain categories of devices, depending
| |
| , on the amount and type of byproduct material that they contain.
| |
| For those devices that must be leak tested, Part 31 also contains i specific requirements concerning the recording of test results and s retention of those records. In addition, Part 31 requires i j servicing of on-off mechanisms and retention of records of I
| |
| : servicing. Finally, any events involving failure of shielding. l on-off mechanisms or indicators, or the release of radioactive l material must be reported within 30 days to NMSS.
| |
| l e Comn11mnce with transfer and di nosal recuirements. Under 10 CFR Part 32, firms with a specific distribution license are required to report all transfers quarterly within 30 days after the end of the quarter to NMSS, and must maintain the transfer records for five years. The 5eneral licensees to whom the tranefer occurs are prohibited by 10 CFR 531.5(c)(5) from abandoning the device, by 531.5(c)(7) from exporting the device, except in accordance with specified procedures, by 531.5(c)(8) from transferring the device, except to a person holding a specific license and with notification to NMSS within 30 days, and by 531.5(c)(9) from transferring the device to another general licensee, except under certain specified conditions. Finally, Part 31 contains specific
| |
| | |
| , .. l l 25 requirements concerning receipt and transfer records and their retention, and 10 CIR Part 20 includes requirements for reporting i
| |
| of theft or loss of devices as well as reporting of exposures ~or incidents involving devices.
| |
| The survey results are reported in subsections 4.2.1, 4.2.2, and 4.2.3
| |
| ' for general licensees for analytic devices, gauges, and self-powered lights i
| |
| respectively. The responses of general licensees are provided in the following subsections on a question-by-question basis, unless otherwise noted.
| |
| In general, questions allowed either a yes or no answer, and did not include an opportunity for the respondent to indicate " don't know" or a similar equivocal response. Responses are reported as yes or no, and an additional entry (DNR for "did not respond") is included to account for survey forms returned with the relevant question left blank. In addition, the tabulated results do not reflect inquiries made by respondents over the telephone hotline, or inquiries made to NRG directly or to specific licensees. In some cues, therefore, tabulated responses reflect the results of inquiries made by the respondents before completing the questionnaire, and may indicate a higher j
| |
| level of knowledge of regulatory requirements than the respondent actually
| |
| * possessed before receiving and responding to the survey. When hotline records indicate a high level of uncertainty on the part of respondents, those re ults are noted.
| |
| l 4.2.1 General Licensees for Analytic Devices ,
| |
| I' The results reported below are based on two categories of responses --
| |
| fira-specific responses and device-specific responses. Overall, 783 analytic
| |
| ! device respondents had returned forms at the time the tabulations were l calculated. Firm-specific responses may not total 783, ho ever, because in some cases the form may have been returned with one or more questions unanswered. In addition, 783 separate firms are not represented by the i 5
| |
| responses. A smaller number of firms is represented, since in some cases different divisions or locations of the same firm were sent separate survey questionnaires.
| |
| l The results pertain to more than 785 devices, since in a number of cases '
| |
| the same respondent possessed more than one device at the same site and was sent more than one device-specific questionnaire. The number of devices being reported upon varies from question to question, but is about 950 for most questions. Totals are provided for yes and no answers for each question, and the percentages are calculated from those answers rather than the total for i l
| |
| the question including DNR. l l
| |
| Knowledge of General License Requirements General licensees for analytic devices indicated a relatively high degree of uncertainty about whether they possessed a copy of the General l License. This question was included to determine if general licensees were !
| |
| aware that the regulations themselves set the terms and conditions of the l license. A copy of 10 CFR Part 31 was included in each questionnaire package.
| |
| i
| |
| | |
| ,~
| |
| j l
| |
| 26 A significant number of hotline calls were received asking about this question, and the hotline response explained the situation to callers (see Appendix C). Even so, about 30 p rcent of the respondents-indicated they did not have a copy of the general license.
| |
| Question II.B Does your firm have a copy of the General Licensef I Yes 471 70%
| |
| No 222 30%
| |
| 673 DNR 20 question II.c Is there a company representative who knows the General Licanse Conditions?
| |
| Yes 601 88.1%
| |
| No _81 11.9%
| |
| 682 DNR 11 Question II.D Is that person responsible for ensuring compliance with the General License conditions? 1 Yes 593 90.0%
| |
| No _11 10.0%
| |
| 659 DNR 34 Generallicenseesforahticd not report difficulty in finding authorized recipients for storage or disposal of devices. Only 6 respondents (or less than one percent of respondents) reported difficulty in finding an authorized recipient to purchase devices; 11 (about 1.6 percent) reported difficulty in finding a recipient with whom to dispose the device; and only three reported diffievity in finding someone with whom to store unwanted devices.
| |
| Only a few respondents reporting holding either devices or sealed sources in secured storage (Questions F and G) as shown in Exhibit 4-6:
| |
| | |
| ., __ --. .. . - . _ . . . . . . . .~ ..- - _ _- -. . - . _ . - . - - - - - - .
| |
| i l .-
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| 1 i
| |
| i 27 I
| |
| l 3
| |
| EERIBIT 4-6
| |
| { DEVICES PfD SOURCES REPORTED AS IN STORAGE Number Held in Respondents With Respondents With
| |
| ] Secured Storage Devices in Storage
| |
| , Sources in Storage 4
| |
| : Number Percent Number Percent
| |
| - 0 644 92.9 648 93.5 1 26 3.8 24 3.5 2 14 2.0 l 3 12 1.7 3 0.4 1 0.1 l
| |
| { 5 3 0.4
| |
| ! 8 --
| |
| 1 0.1 l 9 2 0.3 1 0.1 -- --
| |
| l 11 --
| |
| i 12 1 0.1
| |
| # 1 0.1 1 0.1
| |
| ' 13 1 0.1 -- --
| |
| 19 -- --
| |
| 2 0.3 40 -- --
| |
| 1 a.1
| |
| | |
| ~
| |
| 28 i
| |
| Installation, Current condition, Inspection, and Maintenance 1
| |
| 'A number of survey que-tions secured information about the installation !
| |
| of the device, the current condition of che device, inspection and maintenance l practi- s, and recordkeeping. l About a third of the devices were reported as having been installed by the general licensees themselves (the number of responses is greater than the ;
| |
| number of respondents because some respondents reported on more than one device), l i
| |
| Question V.D Did your firm install this devicef
| |
| : i
| |
| . Yes 337 35.3% i No 111 64.7% '
| |
| 956 DNR 23 '
| |
| 1 A large number of respondents also reported that the Specific Licensee from whom they had obtained the device had installed the device at their site:
| |
| Question V.D1 Was the Specific Licensee identified in Part III the l same organization that installed this device at your site?
| |
| Yes 546 94.5% !
| |
| No _22 5.5%
| |
| 578 DNR 401 Finally, a small number of devices were apparently installed by neither the general licensee nor the specific licensee. Respondents indicated about ;
| |
| seven other firms that had installed devices (Questions V.D.2 to D.10).
| |
| Under Part 31, devices containing krypton-85 and tritium (gas) sources are not required to be leak tested; devices containing gamma / beta sources are required to be tested only when the source is greater than 100 uci and devices containing alpha sources are required to be tested only when the source is greater than 10 nei. The most numerous types of sources are listed in Exhibit 4-7. .
| |
| 1 1
| |
| | |
| i
| |
| * i I
| |
| i .
| |
| I i
| |
| 1
| |
| ! 29 1
| |
| i j n IBIT 4-7 l SOURCE TYPES MOST FREQUENTLY IDENTIFIED i
| |
| Source type Number Percent i
| |
| l Ni-63 387 39.5 l Cs-137 154 15.7
| |
| ! Po-210 93 9.5
| |
| ! Fe-55 52 5.3 i
| |
| I i
| |
| l l
| |
| i 9
| |
| , . 1 i !
| |
| }
| |
| l l.
| |
| | |
| M 30 l
| |
| ' Relatively few devices were reported containing krypton-85 (17 devices, 1.7%)
| |
| or tritium (30 devices, 3.1%). The questions pertaining to initial radiation testing and leak testing outlincl the estegories of devices exempt from testing, and instructed respondents to skip the questions if they possessed such devices. About 180 responses were received in which these questions were not answered.
| |
| ; Exactly half of the respondents, however, reported having a copy of the initial radiation survey.
| |
| ! Question V.E
| |
| ; Does your firm have a copy of the initial radiation survey performed at the time of installationt Yes 395 50%
| |
| ; No 111 50%
| |
| j 790
| |
| ; DNR 189 Slightly less than half reported that leak tests had been performed on the device following its receipt:
| |
| 4 question V.F Since your firm received this device, have leak tests been performed?
| |
| k Yes 396 49.7% !
| |
| No 492 50.3%
| |
| 796 4
| |
| DNR 183 Out of the 381 dates reported by respondents as the date of the latest leak test (Question V.F.1), about 16% (61 tests) fell before the beginning of 1990. Forty percent of the tests were performed after June 1,1990, and 25*4 were performed after July 1.
| |
| 2 A slightly smaller number of firms reported having a copy of the most recent leak test than reported having conducted such tests:
| |
| Question V.F.2 Does your firm have copies of the record of the latest test?
| |
| Yes 343 87.9%
| |
| No - 47 12.1%
| |
| 390 DNR 589 Finally, 395 of the 396 firms reporting tests provided data on who had conducted the leak test. About two-thirds of such tests were not conducted by the general licensee.
| |
| | |
| 1 31 Question V.F.3 Did your firm conduct the latest test?
| |
| Yes 142 35.9%
| |
| No 111 64.1% ,
| |
| 395 DNR 584 i
| |
| i 1
| |
| Section 31.5(c)(2) requires servicing of the on off mechanism and indicator of a device with such mechanism every six months, or as otherwise specified.
| |
| Records of the test results must be maintained for one year.
| |
| About one-third of the analytic devices covered by the survey had an on- I off mechanism. l j question V.C Does the device have an on-off mechanism?
| |
| Yes 286 34.9%
| |
| No ill 65.1%
| |
| 819 DNR 160 ,
| |
| 1 Less than half of those analytic devices with on-off mechanisms, however, had a test of the mechanism performed at any time since the device i 4
| |
| was received. I' j question V.G.1 Since your firm received this device, have on-off j , mechanism tests been performed?
| |
| Yes 109 39.8%
| |
| No 111 60.2%
| |
| 274 DNR 705 The dates of 97 out of the 109 tests were reported (Question V.G2). Of these dates,18 fell between March 1988 and December 1989, outside the six month period preceding the mailing of the survey. In contrast, about 70 percent of the tests had occurred after February 1990, indicating that those general licensees who were conducting on-off mechanisms testing were doing so 1
| |
| in a timely manner.
| |
| Relocation, Transfer, and Disposal i
| |
| i Respondents reported relatively few sicurtions in which analytic devices were relocated after installation, although a number of hotline calls were received asking for clarification on the meaning of relocation.
| |
| d 1
| |
| | |
| - l 32 Question V.H.
| |
| Has this and reinstalled) since device it was been installed?
| |
| initially relocated (noved within the site Yes 143 15.5%
| |
| No 124 84.5%
| |
| 952 DNR 27 Almost all of those analytic devices that were relocated were moved and reinstalled by the general licensee in possession of the device. :
| |
| Question V.H1 Did your firm relocate this device?
| |
| Yes 129 85.4%
| |
| No _22 14,6%
| |
| 151 !
| |
| DNR 828 )
| |
| )
| |
| Very few respondents (16) reported the date that the device was last relocated (Question V.H2). All but four reported relocating the device during 1990. i i
| |
| Sixteen of the 22 firms that reported that someone else had relocated a device for them provided information on the name of the company that had moved the device. Almost all of these firms appeared to be specific licensees; in 4
| |
| two cases,analytic relocated however, commercial moving and storage firms were listed as having devices.
| |
| Even fewer general licensees reported having transferred devices to another company than reported relocations of devices.
| |
| Question V.J1 Has your company transferred this device to another firm?
| |
| Yes 78 8.1%
| |
| No Eli 91.9%
| |
| 962 DNR 17 -.
| |
| Out of the 78 devices that were reported transferred, general licensees provided information on the firm receiving the device in approximately 60 cases. Of these, slightly more than half did nas appear on the list of specific licensees. ' Additional analysis is necessary before it can be determined if these transfers met the requirements of 10 CFR 631.5(c)(9), but in several cases the device does not appear to have remained at the same location. One respondent explained that until the survey had been received, he had been unaware of the transfer provisions in Part 31, and had immediately notified NRC of a transfer.
| |
| | |
| 33 In at least one case, a device initially shipped to a location in the United States was transshipped overseas. The respondent, who replied by )
| |
| letter, did not supply enough information to determine if NRC precedures for such shipment were followed and appropriate reports submitted.
| |
| Finally, ten situations were reported in which devices could have been or were damaged.
| |
| i i
| |
| Question V.I Has your firm experiencad any events that could have caused damage or did cause damage to this device?
| |
| Yes 10 1.0%
| |
| No 111 99.0%
| |
| 967 4
| |
| DNR 12 A review of the supplementary information on these cases is underway to
| |
| ; determine the circumstances of the damage.
| |
| 4.2.2 General Licensees for Gauges a
| |
| i The results reported below are based on two categories of responses --
| |
| ' fira-specific responses and device-specific responses. Overall, 1068
| |
| , respondents had returned forms at the time the tabulations were calculated.
| |
| As with analytic device respondents, fira-specific responses may not total l3I 4
| |
| 1068 because in some cases the form may have been returned with one or more questions unanswered.
| |
| the responses.
| |
| In addition, 1068 separate firms are not represented by A smaller number of firms is represented, since in some cases i different divisions or locations of the same firm were sent separate survey questionnaires.
| |
| 1 i
| |
| The results pertain to more than 1068 devices, since in a number of cases the same respondent possessed more than one J.vice at the same site and was sent more than one device-specific questionnaire. The number of devices i being reported upon varies from question to question, but is about 1750 for most questions. Totals are provided for yes and no answers for each question,
| |
| { and the percentages are calculated from those answers rather than the total
| |
| { for the question including DNR.
| |
| Knowledge of General License Requirements General licensees for gauges indicated a slightly higher degree of knowledge about whether they possessed a copy of the General License than did analytic device respondents. A copy of 10 CFR Part 31 was included in each questionnaire package. Potential gauge respondents placed a significant number of hotline calls asking about this question, and were told that the regulations constituted the terms and conditions of the general license. Even so, about 22 percent of the respondents indicated they did not have a copy of the general license.
| |
| | |
| 4
| |
| . l J'.* j i
| |
| 34 l Question II.B Does your firm have a copy of the General Licensef I
| |
| Yes 652 78.2%
| |
| j No 112 21.81 834 DNR 22 I
| |
| Question II.C Is there a company representative who knows the !
| |
| General License Conditions? j I
| |
| Yes 756 89.7% 1 No _11 10.3% ;
| |
| 843 DNR 13 Question II.D Is that person responsible for ensuring compliance with the General License conditions?
| |
| Yes 757 91.4%
| |
| No _11 8.6% ,
| |
| 828 )
| |
| DNR 28 .
| |
| General licensees for gauges also did not report difficulty in finding authorized recipients for storage or disposal of devices. Only 5 respondents (or about 0.6 percent of respondents) reported difficulty in finding an i authorized recipient to purchase devices; 6 (about 0.7 percent) reported I difficulty in finding a recipient with whom to dispose the device; and only three reported difficulty in finding someone with whom to store unwanted ;
| |
| i devices.
| |
| * 1 Only a few respondents reporting holding either devices or sealed I sources in secured storage (Questions F and G) as shown in Exhibit 4-8:
| |
| I 1
| |
| | |
| 1
| |
| . e l
| |
| s -
| |
| 1 1
| |
| 35
| |
| ,' t EIHIBIT 4-8 I CAUGES AND SOURCES REPORTED AS IN STORACE I i
| |
| i Number Held in Respondents With Respondents With Secured Storage Gauges in Storage Sources in Storage !
| |
| )
| |
| Number Percent Number Percent 1
| |
| O 796 93.0 801 93.6
| |
| { 1 28 3.3 21
| |
| ; 2 2.5 13 1.5 15 1.8 i 3
| |
| ' 6 0.7 7 0.8 4 2 0.2 l 5 3 0.4 1 0.1 1 0.1
| |
| { 9 1 0.1 -- -- !
| |
| i 11 2 0.2
| |
| ! 14 1 0.1 2 0.2 -- --
| |
| ; 15 1 0.1 i 16 --
| |
| 1 0.1 ;
| |
| i 20 1 0.1
| |
| ' 2 0.2 2 0.2 i 37 --
| |
| j 50 --
| |
| 1 0.1 4
| |
| 1 0.1 I
| |
| i i 1 1 l l
| |
| | |
| 4 36 Installation, Current Condition, Inspection, and Maintenance A number of survey qu stions secured information about the installation of the device, the current condition of the device, inspection and maintenance practices, and recordkeeping.
| |
| About a third of the devices were reported 2s having been installed by the general licensees themselves (the number of responses is greater than the number of respondents because some respondents reported on more than one device).
| |
| Question V.D Did your firm install this device?
| |
| Yes 351 20.7%
| |
| No 1242 79.3%
| |
| 1693 DNR 57 A large number of respondents reported that the Specific Licensee from whom they had obtained the device had also installed the device at their s!.ce:
| |
| Question V.D1 Was the Specific Licensee identified in Part III the same organization that installed this device at your site?
| |
| Yes, '5 91.2%
| |
| No _
| |
| ', 8.8% i 1279 l DNR 471 '
| |
| Finally, a significant number of devices were apparently installed by neither the general licensee nor the specific licensee. Respondents indicated j about 57 instances in which other firms that had installed devices, and identified at least 35 different firms that had performed installations. !
| |
| (Questions V.D.2 to D.10).
| |
| l i
| |
| Under Part 31, devices containing krypton-85 and tritium (gas) sources are not required to be leak tested; devices containing gamma / beta sources are j required to be tested only when the source is~ greater than 100 uci and devices I containing alpha sources are required to be tested only when the source is l greater than 10 nci. The most numerous types of sources are listed in Exhibit '
| |
| 4-9:
| |
| | |
| - _. -._ . - . -- . . . . . - - . . . . . . . . _ - - _ . . . _ - - - - ~ _ _ - . ~ . - . . . . . _ . - - - - - - - - . .
| |
| 1 -
| |
| I j
| |
| . 37 EIHIBIT 4-9 i
| |
| SOURCE TYPES MOST FP'.QUENTLY IDENTIFIED i
| |
| Source type Number Percent Kr-85 446 25.5 Cs-137 479 27.3 An 241 451 25.7 Sr-90 146 8.6 !
| |
| Pm-147 89 5.1 H3 16 l
| |
| 0.9 '
| |
| l l
| |
| 1 l
| |
| l
| |
| | |
| i 1
| |
| l 38
| |
| \
| |
| The questions pertaining to initial radiation testing and leak testing outlined the categories of devices exempt from testing, and instructed respondents to skip the question, if they possessed such devices. Over 250 responses were received in which these questions were not answered.
| |
| In contrast to the respondents for the analytical devices, a high proportion of the gauge respondents reported having a copy of the initial radiation survey.
| |
| Question V.E Does your firm have a copy of the initial radiation survey performed at the time of installation?
| |
| Yes 1229 83.7%
| |
| No 112 16.3%
| |
| 1468 DNR 282 A high proportion of gauge respondents also reported that leak tests had l been performed on the device following its receipt:
| |
| Question V.F Since your firm received this device, have leak tests been performed? I Yes 1191 80.3%
| |
| - No _222 1483 19.7%
| |
| !l DNR 267 Out of the 1146 dates reported by respondents as the date of the latest leak test (Question V.F.1), over 300 (about 28%) fell before the beginning of 1990. About 33 percent of the tests were performed after June 1,1990, and about 20 percent were performed efter July 1.
| |
| A slightly smaller number of firms reported having a copy of the most recent leak test than reported having conducted such tests:
| |
| Question V.F.2 Does your firm have copies of the record of the latest test?
| |
| Yes 1113 95.5%
| |
| No __11 4.5%
| |
| 1165 DNR 585 Finally, 1750 names of testing companies were provided as data on who had conducted the leak test if the general licenses had not done so. About 80 percent of such tests were not conducted by the general licensee.
| |
| | |
| 39
| |
| ; question V.F.3 Did your firm conduct the latest test?
| |
| Yes 232 19.8%
| |
| No _242 80.2%
| |
| 1174 1
| |
| DNR 576 4
| |
| Section 31.5(c)(2) requires servicing of the on-off mechanism and indicator I
| |
| 4 of a device with such mechanism every six months, or as otherwise specified.
| |
| Records of the test results must be maintained for one year.
| |
| Almost 90 percent of the gauges covered by the survey had an on/off mechanism.
| |
| j Question V.G Does the device have an on-off mechaniset f Yes 1465 87.3%
| |
| No _211 12.7%
| |
| 1 1678
| |
| : DNR 72 i
| |
| A high proportion of the gauges with on-off mechanisms had a test of the mechanism performed since the device was received. In one case, however, the i' respondent added a note explaining that since the' device was never turned off ,
| |
| no tests were performed on the on-off mechanism (G320/0666). I I
| |
| Question V.G.1 Since your fire received this device, have on-off mechanism tests been performed?
| |
| ) . Yes 1252 86.1%
| |
| ) No _222 13.9%
| |
| i 1454 J DNR 296 j The dates of 1199 out of the 1252 tests were reported (Question V.G2).
| |
| Of these dates, about 240 fell prior to December 1989, outside the six month
| |
| { period preceding the mailing of the survey. In contrast, about 75 percent of the tests had occurred after February 1990, indicating that those general
| |
| , licensees who were conducting on-off mechanisms testing were doing so in a timely manner.
| |
| Relocation, Transfer, and Disposal Respondents reported relatively few situations in which gauges were relocated after installation, although a number of hotline calls were received asking for clarification on the meaning of relocation.
| |
| | |
| ._. . -- - . _ - - - . . . . _ . ~.
| |
| . l l
| |
| i 40 1 Question V.H.
| |
| Has this device been relocated (noved within the site and reinstalled) since it was initially installed?
| |
| l Yes 128 7.6%
| |
| No lili 92.4%
| |
| 1682 DNR 68 A high proportion of the gauges that were relocated were moved and reinstalled by the general licensee in possession of the gauge, according to the survey responses. Respondents reported, for example, that the production line where the gauge had been used had been closed and the gauge removed from the line and stored (G305/643), or that a truck-mounted gauge had been removed and transferred to another section of the same company located several miles away (hotline contact).
| |
| Question V.H1 Did your firm relocate this device?
| |
| Yes 95 75.4%
| |
| No .,_3.1 24.6%
| |
| 126 DNR 1624 Few respondento (29) reported the date that the device was last relocated (Question V.H2). Less than half reported relocating the device !
| |
| during 1990.
| |
| Information was presented for 29 situations in which a gauge was relocated by a firm other than the general licensee on the identity of the firm that performed the relocation. At least 20 of the firms were specific licensees whose quarterly reports had been used to develop the survey frame; at least two other firms were included on NRC device vendor lists.
| |
| A greater number of general licensees reported having transferred gauges to another company than reported relocations of gauges.
| |
| Question V.J1 Has your company transferred this device to another fira?
| |
| Yes 150 8.7%
| |
| No lill 91.3%
| |
| 1728 DNR 22 Out of the 150 devices that were reported transferred, general licensees provided information on the firm receiving the device in almost every case. A substantial number of the approximately 70 firms listed did Des appear on the list of specific licensees who had been the initial vendors of the gauges.
| |
| Additional analysis is necessary before it can be determined if these transfers met the requirements of 10 CFR 531.5(c)(9), but in several cases the
| |
| | |
| ~ j
| |
| . l 41 gauge does not appear to have remained at the same location. Respondents
| |
| - reported, for example, that after a particular plant had closed the device had been transferred to another company but was currently in storage at the specific licensee's facility (G417/864). In some cases, respondents enclosed copies of reports made to NRC concerning transfers (C461/970).
| |
| In at least one situation, the respondent reported that the gauge had been shipped overseas without having been used or even unpackaged in the United States (G049/107).
| |
| l l
| |
| Finally, 13 situations were reported in which devices could have been or '
| |
| were damaged.
| |
| Question V.I Bas your firm experienced any events that could have caused damage or did cause damage to this device?
| |
| Yes 13 0.8%
| |
| No 1Z91 99.2%
| |
| 1714 DNR 36 A review of the supplementary information of these cases revealed the l following:
| |
| The shutter on a void detection device in a silo was found to be j broken in the closed position. The unit was locked in the closed position and the area secured. A new housing was installed, following modification by the manufacturer. About 16 months later the shutter failed again. Following consultation with the manufacturer, the gauge was moved to a position where it experienced less vibration. (G698/1422) e Electronic components failed due to por:r surges in the 110 v circuit. Electrical repairs were conducted by the manufacturer.
| |
| (C613/1268) e Following a plant fire the manufacturer's service personnel checked the gauge immediately. No source leakage was detected.
| |
| (C601/1252) 4.2.3 General Licensees for Tritium-Powered Exit Signs The results reported below are based on two eptegories of responses --
| |
| fira-specific responses and device specific responses. Overall, 377 respondents had returned forms at the time the' tabulations were calculated.
| |
| Firm-specific responses may not total 377, however, because in some cases the form may have been returned with one or more questions unanswered. In addition, 377 separate firms are not represented by the responses. A smaller
| |
| | |
| 42 number of firms is represented, since in some cases different divisions or i
| |
| locations of the same firm were sent separate survey questionnaires.
| |
| ' 1he results pertain to more than 377 signs, since in a number of cases the same respondent possessed more than one cign at the same site. In
| |
| ' contrast to the questionnaires for analytic devices and gauges, however, a l
| |
| " duplicate device-specific questionnaire was not sent for each tritium-powered exit sign held by a respondent. Instead, the responder.t was provided information about the number of signs covered by the survey and requested to respond to each question as it related to the sign or signs hnid. In some 4
| |
| cassa, when respondents had received shipments of signs in addition to those covered by the survey, respondents indicated that they held additional signs and a number of hotline telephone calls were received from general licensees inquiring whether they should respond 'or all signs held by them. Other potential respondents did not understand that the questionnaire pertained to more than one sign. The number of signs being reported upon varies from question to question. Totals are provided for yes and no answers for each question, and the percentages are calculated from those answers rather than the total for the question including DNR.
| |
| Knowledge of Caneral License Requirements General licensees for tritium-powered exit signs indicated a relatively high degree of uncertainty about whether they possessed a copy of the General License. This question was included to determine if general licensees were i aware that the regulations themselves set the terms and conditions of the license. A copy of 10 CFR Part 31 was included in each questionnaire package.
| |
| A significant number of hotline calls were received asking about this question, and the hotline response explained the situation to callers (see Appendix C). Even so, almost half of the respondents indicated they did not have a copy of the general license.
| |
| Question II.B Does your firm havs a copy of the Generai License?
| |
| Yes 163 $1.1%
| |
| No lli 48.9%
| |
| 319 DNR 9 -
| |
| Question II.C Is there a company representative who knows the General License Conditions?
| |
| Yes 189 59.4%
| |
| No 111 40.6.
| |
| 318 DNR 10 Some survey respondents reported that a company representative was appointed aft,r receipt of the survey (223), or that the receipt of the survey was their first knowledge that the signs were "anything special." (237)(485)
| |
| | |
| _. - . _ _ _ _ . .~ - - __
| |
| 1 43 question II.D
| |
| +
| |
| Is that person responsible for ensuring compliance with the General Lic.ise conditions?
| |
| j Yes 188 68.9%
| |
| No _11 31.1%
| |
| 273
| |
| : DNR 55 One important result of the survey was the development of information that certain respondents and/or specific licensees had been reporting emergency signs.for aircraft under 531.5 rather than 531.7. Steps were taken to ensure that results contained in this repore pertain only to 531.5 general licensees.
| |
| General licensees for self-povered lights reported that they had no i
| |
| difficulty in finding authorized recipients for storage or disposal of the lights.
| |
| TAy 2 respondents (or less than one percent of respondents) reported difficulty in finding an authorized recipient to purchase lights; similarly, only 2 reported difficulty in finding a recipient with whom to dispose of the '
| |
| 4 lights; and only 1 reported difficulty in finding someone with whom to suore l unwanted lights. One respondent suggested, however, that it had tried and failed to dispose of a sign in the local landfill, which would not accept it, and had not yet contacted the manufacturer about disposal. (218) only a few respondents reporting holding either exit lights in secured storage (Questions F and C) as shown in Exhibit 4-9:
| |
| | |
| 1
| |
| * i i
| |
| r J =
| |
| * i t
| |
| 1, i
| |
| 4 44 i,
| |
| 3 EIHIBIT 4-9 EXIT 'IGNS REPORTED AS IN STORAGE t
| |
| 3 Number Held in Respondents With 1 i
| |
| Secured Storage Signs in Storage Number Percent i
| |
| 4 j 0 309 94.2 '
| |
| ; 1 11 3.4 l
| |
| ; 4 2 0.6 !
| |
| j 7 1 0.3 i 12 2 0.6 j 14 1 0.3 4
| |
| 30 1 0.3 l i
| |
| 46 1 0.3 1' l
| |
| ] l l l 2
| |
| t 1
| |
| 1 i
| |
| 4 i
| |
| 1 l
| |
| f f
| |
| i.
| |
| ! J l
| |
| i 1
| |
| l a
| |
| l.
| |
| 4 i
| |
| i i
| |
| i I.
| |
| i .
| |
| 1 3
| |
| 4
| |
| * e
| |
| . . j l
| |
| l' 45 Installation, Current Condition," Inspection, and Maintenance A number of survey questions secured information about the installation of un exic signs, the current condition of the signs, inspection and maintenance practices, and recordkeeping.
| |
| Over half of the signs were reported as havin5 been installed by the general licensees themselver (the number of responses is greater than the number of respondents because several respondents reported on more than one sign).
| |
| Question V.B Did your firm install the exit sign (s)?
| |
| Yes 188 60.1%
| |
| No 116 37.1%
| |
| Some 9 2.9%
| |
| 313 1 DNR 15 Very few respondents reported that the Specific Licensee from whom they had obtained the signs had also installed the signs at their site:
| |
| Question V.BL Vas the Specific Licensee identified in Part III the ;
| |
| same organization that installed the exit sign (s) at ;7our site?
| |
| Yes 14 14.7%
| |
| }ll No 11 85.3%
| |
| 95 DNR 233 A significant number of respondents identified themselves as electrical contractors, and explained that they had purchased the signs either for a specific client or had purchased them for resale. In most cases, such respondents could not identify the ultimate recipient of the signs. #
| |
| Respondents also identified non-standard uses of signs that complicated their !
| |
| tracking (e.g., installation on movable partitions (119), use for aircraft I escape hatch marker (020), ese for commercial display sign, after modification (site visit information).
| |
| Under Part 31, devices containing tritium (gas) sources are not required to be leak tested; therefore, the questionnaire did not contain any questions l about leak testing. Similarly, testing of on;off mechanisms and indicators is not required for self-powered exit signs, and no questions were asked on that subject. Such signs are required, however, to be clearly labeled and to be installed in such a manner that the label is visible. Almost all respondents reported that the labels were durable, legible, and clearly visible.
| |
| r i
| |
| | |
| 5 I
| |
| 46 Question V.E Do the exit sign (s) have:
| |
| El. A durable labelf
| |
| } Yes 285 99.0%
| |
| No .,_1 1.0%
| |
| : 288 DNR 40 E2. A legible (easily readable) label?
| |
| Yes 281 98.3%
| |
| No __1 1.7%
| |
| 286 DNR 42 E3. A clearly visible (easily seen) label?
| |
| Yes 262 92.3%
| |
| No _12 7.7%
| |
| 284 DNR 44 These results should be considered in light of a number of calls that I were received on the telephone hotline from respondents who noted that the labels were quite small and could be located over doors that placed them ten or more feet over the head of someone trying to read the label. One respondent included a copy of a label indicating that the recommended maximum distance for legibility is 100 feet.(283) Another stated that the signs were not legible upon receipt, but that they had been relabeled by the general licensee (299)
| |
| Respondents also reported that a substantial proportion of the signs were checked for defects af ter receipt by the general licensee.
| |
| Question V.A Vere the exit sign (s) over checked for defects after receipt?
| |
| Yes 223 70.8%
| |
| No 90 28.6%
| |
| Some __2 0.6%
| |
| 315 DNR 13 Relocation, Transfer, and Disposal Respondents reported several situations in which exit signs were relocated after installation, although a number of hotline calls were received J
| |
| | |
| . I'
| |
| . i i
| |
| J 47 asking for clarification on the meaning of relocation.
| |
| g Question V.C Have the exit sign (s) been relocated (moved within the site and reinstalled) since initial installation?
| |
| 1 Yes 8 2.6%
| |
| No 300 96.8%
| |
| Some __1 0.6%
| |
| 310 DNR 18 4
| |
| One respondent, however, reported having relocated 34 such signs.
| |
| Almost all of the general licensees who reported signs that had been relocated also reported relocating the signs themselves.
| |
| ) Question V.H1 Did your firm relocate this device?
| |
| , Yes 6 54.5%
| |
| No _1 45,5%
| |
| ; 11 DNR 317 j
| |
| Few general licensees for tritium exit signs reported having transferred
| |
| ; signs to another company. I t
| |
| Question V.F Has your company transferred any exit sign (s) to I
| |
| another firm?
| |
| Yes 26 8.2%
| |
| No 121 91.8%
| |
| 318 DNR 10 As noted above, a number of survey responses were received from electrical contractors, vendors, and building contractors stating that they had received lights from specific licensees but had passed them on to ultimate users or installed them in structures. In such cases, the number of lights that were transferred was usually not specified; however, two respondents did state that they had transferred 100 and 150 lights respectively. In at least one case, a transfer of signs back to the manufacturer was initiated by the survey. This respondent reported that the survey had requested information on 11 signs; inquiries revealed that 24 had been purchased by plant personnel.
| |
| Two could not be located, 2 were removed from use and returned, and 20 that had not been used were also returned. (185)
| |
| Finally, two situations were reported in which lights could have been or were damaged.
| |
| I l
| |
| | |
| ; e .
| |
| 48 Question V.D Has your firs experienced events that could have caused damage or did cause damage to any exit sign (s)?
| |
| Yes 2 0.6%
| |
| No 114 99.4%
| |
| 316 DNR 12 j
| |
| A review of other information revealed the following additional situations:
| |
| 1 Signs installed by a previous owner were removed prior to receipt of the survey by a licensed demolition contractor and disposed.
| |
| , (123) i A forklift struck a sign and damaged the mounting device. The i
| |
| mounting device was replaced and the sign rehung. Geiger counter j monitoring was performed on the sign. (076) i
| |
| * A motel suffered the theft of 18 signs and damage to the casing of 4 others. The incident was reported to the NRC and the damaged i
| |
| signs were returned to the supplier. (390) e A motel suffered the theft of 2 signs. The incident was reported l
| |
| ; to the police.(391)
| |
| Finally, one respondent reported that all equipment delivered to the 4
| |
| location in question had been shipped overseas. (33) s k,
| |
| i 4
| |
| 4 i
| |
| 1' f
| |
| | |
| i
| |
| \
| |
| i I
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| a i
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| l
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| * I l
| |
| i APPENDIX A '
| |
| SURVEY QUESTIONNAIRES l
| |
| l 1
| |
| I l
| |
| i O
| |
| l l
| |
| M. & .- w w on a e..
| |
| | |
| i i
| |
| . '. Form Approved j ,
| |
| . OMB No. 3150-0152
| |
| , Expires 10/31/92
| |
| ! U.S. Nuclear Regulatory Commission o*
| |
| l Survey of General Licensees: Facility information
| |
| $1 l
| |
| i Part I contains proprinted intbrmation about your firm. Please review the proprinted idwirist%n
| |
| } in Part I and correct any incorrect informat.un by crossireg t*. rough the incorrect lnfeiristion and
| |
| ' writing in the correction above or to the right of the original inimiristion. Please complete any missing information by writing the information in the space provided for that information.
| |
| } Part ll of this survey asks additionalinformation about your firm and your experience dealing with i the devices identified in Parts ill and IV of this survey. Please answer all questions in Part ll as l j they relate to your firm as a whole. Please fillin or circle the answer as appropriate.
| |
| i I
| |
| 4 !
| |
| Past1: Respondent identification (Please correct any Incorrect infbrma#on)
| |
| A. Company Name l
| |
| ! B. Company Contact (First/Last Name) l C. Contact Title
| |
| )
| |
| l D. Address i
| |
| s i
| |
| l E. City l l 1
| |
| F. State G. Zip Code H. Phone Number Part II: Additional Respondent Information Please IlliIn or circle the answer as appropriate A. What is the principal business conducted at your facility?
| |
| 1 - Agriculture 6 - Laboratories 11 - State / Local Gov't 2 - CarVBottling 7 - Mining 12 - Steel 3 - Chemical 8 - Petroleum 13 - Waste Treatment 4 - Food 9 - Power-Coal Feed 14 - Other (ple specM O 5 - Foundry O 10 - Pulp and Paper B. Does your firm have a copy of the General Ucense? Yes No Page1
| |
| | |
| 9 Form Approved OMB No. 31504152 Expires 10/31/92 U.S. Nuclear Reculatory Commission o*
| |
| Survey of General Uconsees: Facility information (continued)
| |
| Part 11: Additional Respondent information (conunued),
| |
| C. Is there a cc4r.pmily representative who knows the O Yes O No General Ucense conditions?
| |
| D. Is that person responsbie for ensuring compliance with Yes No the General Ucense conditions?
| |
| E. Is your firm having any difficulty finding an authorized recipient to:
| |
| E1. Purchase any devices held under General Ucense you no longer want?
| |
| Yes O No E2. Dispose of any devices you no longer want? Yes No E3. Store any devices you no longer want? O Yes O No Wyee anenwed Ne_se; quessesif,4 and 4;:
| |
| 'at$pseguessesF4 E4. If your firm is having any difficulty finding an authorized recipient, please elaborate by describing who you contetod and describe the nature of the difficulty.
| |
| Ploese contmue your answer on additional blank pages as needed. Label each blank page with your company name and the 10 number at the top of this page.
| |
| F. How many devices, held under a General Uconse, are now in secured storage and are not intended for use?
| |
| G. How many sealed sources, held under a General Ucense, are now in secured storage and are not intended for use?
| |
| Page 2
| |
| | |
| l . . Form Approved i -
| |
| ' OM8 No. 31504152 l j
| |
| Expres 10/31/92 )
| |
| O# ;
| |
| U.S. Nuclear Regulatory Commission Daio* o * - 1
| |
| }
| |
| l Survey of General Licensees: Device information Parts lit and W contain proprinted informat'on about a device containing byproduct material tLit
| |
| } NRC records indicate was delivered to your firm under a General Uconse. Please review the
| |
| ! proprinted irdormation in Parts ill and W and correct any incorrect information by crossing j
| |
| through the incorrect information and writing in the correction above or to the right of the original -
| |
| information. Please complete any missing informadon by writing the information in the space j provided sor that information.
| |
| l Part V of this survey refers to the particular device identified in Part W. Please answer all l questions in Part V as they reiste to the device identified in Part N. Please 21 in or circle the l answer as appropriate. (In all cases, MM means month, DO means day, and YY means the last !
| |
| l two uhpNs of the yest.)
| |
| i Parte M, N, =uf V are reposted for each device the suppNor identNied as douvering to your
| |
| ! faculty. You have been supplied with a complete Part 111, N, and V for each device. Each i
| |
| answer in Parte M, N, and V should refer to the specNic device identifled in Part N of that stapied set of Parts lit, N, and V.
| |
| i P art W: SuppNer (Manufacturer or DistrRudor) of Device in Part IV (Please correct any Incorrect Information)
| |
| A. l Supplier's Specisc Ucense Number '
| |
| i B. Supplier's Company Name C. Supplier's Address D. City E. State F. Zip Code G. Phone Number Part IV: Device IslosellicMion (Please correct any incorrect inibtmeelon)
| |
| A. Device Type B. Supplier's Model Number C. Ref.onuclide Source D. Quantity of radioactivity (Curies)
| |
| E. Determination date of quantity (MM/DD/YY)
| |
| F. Device Serial Number G. Date device shipped to you (MM/DD/YY)
| |
| Page 3
| |
| | |
| 1 l
| |
| Form Approved j
| |
| * OMB No. 31504152 1 Expres 10/31/92 ID #
| |
| l U.S. Nuclear Regulatory Commission Device o
| |
| * Part V:
| |
| l Askssonal Informadon About Device Described in Part IV 1
| |
| (Please RH in or circle the answer as appmpriate)
| |
| ; P! ease answer all questions in Part V as they relate to the device identified in Part IV.
| |
| A.
| |
| l Was the radionuclide over rep;eced on this device? Yes O No i
| |
| ! mwyemassmenesme m3 j .. guesses 4 sem m (
| |
| m i ,j i
| |
| ; A1 If yes, what was the date of this replacement? (MM/DD/YY) i ,
| |
| i !
| |
| } B. Does this device have:
| |
| l B1. A durable label?
| |
| t Yes No !
| |
| j B2. A legible (easily readable) label? l Yes O No l l B3. A clearly visible (easily seen) label? Yes No
| |
| ! O 1 - On the Source Housing O 2 - On the Detector ij 3 - On the Supporting Frame i
| |
| C. Was this device ever checked for defects after receipt? Yes No l
| |
| apeuaneweredno to gueensa 4 a $ ef guesesnog ,
| |
| C1. If yes, when was the device last checked for defects? (MM/DD/YY)
| |
| D. Did your firm install this device?
| |
| O Yes No D1. Was the W Uconsee identood in Part lil the same ;g organization that installed this device at your site? A specMc ucenee le issued to namedpersone mon appecanone '
| |
| g ,,,,,,g lguesses y , w1 4 a $ mi g,,, nan ny ;;
| |
| medpursuant to the reguheone m to CFR 30.31 through 30.35.
| |
| SpecMc Ucenesee are aumorized to receke, store, demonseise. O Yes No and redneribune dowces conesining sealed sources honed on their Ucense (NRC Form 374).
| |
| . yy ,,,,,,,g ye, y ;
| |
| pmyn guaenen E: '
| |
| Pege 4
| |
| | |
| Form Approved OMB No. 31504152 Expires 10/31/92 ID #
| |
| U.S. Nuclear Regulatory Commission Device o
| |
| * Part V: Additional IrWormation About Device Deecribed in Part IV (continued) ,
| |
| if another organization installed this device, please complete the following questions:
| |
| D2. Installer's Specific License Number i D3. Company Name l D4. Contact (First Name/Last Name)
| |
| , D5. Contact Title
| |
| !, D6. Address 4
| |
| 4
| |
| ; D7. City
| |
| { D8. State j
| |
| { D9. Zip Code j D10. Phone Number l
| |
| ; If the device identitledin PartIV contains onlylaypton, sidp to question G. If the device Identitled $,
| |
| in Part IV contains only tritium not more than 100 mk: rocur
| |
| , ies of tho er beta and/or gamma i
| |
| emitting meterial, not more than to microcuries of alphe emitting material, or is held in storage i
| |
| in the original shipping container prior to initial Installation, skip to question H.
| |
| J 1
| |
| E. Does your firm have a copy of the initial radiation Yes No survey performed at the time of installation?
| |
| F. Since your firm received this device, have leak tests Yes No j been performed?
| |
| t j
| |
| 'M F, at$ es1 m e;;;
| |
| i l
| |
| Page5
| |
| | |
| l.
| |
| l Form Approved i
| |
| OMB No. 3150-0152 I ** Expires 10/31/92 ;
| |
| ID # . 1 i U.S. Nuclear Reculatory Commission Daica o
| |
| * i
| |
| ' Part V: Ad Rional kWormation About Devios Described in Part IV (continued)
| |
| F1. What was the date of the latest test? (MM/DD/YY)
| |
| F2. Does your firm have copies of the O Yes No record of the latest test?
| |
| F3. Did your firm conduct the latest test? O Yes O No
| |
| : #you answered Fes so : ,
| |
| quesses 4 @ 8e^
| |
| ^
| |
| j
| |
| ,===sk= a y ~ :
| |
| F4. Testing Company Name F5. Contact (First Name/Last Name) i F6. Contact Title !
| |
| F7. Address 4
| |
| F8. City l
| |
| F9. State F10. Zip Code F10. Phone Number G. Does the device have an onoff mechanism? O Yes O No
| |
| ; Eyouansweemd No ee;
| |
| ,gussess4,ahtsso.)
| |
| ~gumasana ne G1. Sinos your firm received this device, have on-off Yc.: No mechanism tests been peiiciined?
| |
| Myouananmend No ee::
| |
| quesman Of,'sh5 8e9 quosman N; G2. What was the date of the latest test? (MM/DD/YY)
| |
| Pagea
| |
| | |
| Form Approved l OMB No. 31504152 j o Expires 10fJ1/92
| |
| { 10 #
| |
| [ U.S. Nuclear Regulatory Commission De*= o
| |
| * i Part V: N Information Atund Devios DescrReti ki Part IV l (consnued)
| |
| L -
| |
| TW % h G4. Contact (First Name/last Name) i i
| |
| G5. Contact Title G6. Address G7. City i
| |
| 1 i
| |
| G8. m j G9 Zip Code l G10. Phone Number j H. Has this device been relocated (moved within the Yes No !
| |
| j site and reinstalled) since it was initially installed? ., .
| |
| 4 i
| |
| ! cn, .es edme=e j
| |
| e gusasse#4 ad$s304 y; gasseset:p e ,
| |
| I' 1
| |
| 1 i H1. Did your firm relocate this device? -
| |
| Yes No 1
| |
| 1#you asemesed Viss Spi:
| |
| ' gusseen Nf, adqsses , j Squessens t <, !
| |
| H2. What was the date that this device was last moved? (MM/DD/YY) i i
| |
| Page 7
| |
| )
| |
| | |
| i Form Approved OMB No. 3150 0152
| |
| > Expires 10/31/92 ID #
| |
| ! U.S. Nuclear Reculatory Commission De*= m
| |
| * 1 i
| |
| Port V: N IrWormation About Device DescrRed h Part IV
| |
| ! (conemed) 1 H3. Name of company that moved
| |
| !. the device?
| |
| l H4. Company Contact (First Name/Last j Name) l H5. Company Contact Etie
| |
| ! H6. Address i
| |
| 1 j H7. City H8. State l H9. Zip Code 1
| |
| l H10. Phone Number i
| |
| i
| |
| : 1. Has your firm expef.anced any events that could O Yes No I j have caused damage or did cause damage to this device?
| |
| ' ryesansummsfNoso:,
| |
| ; guesens4aApest l & queenanJl -
| |
| a l 11. If your firm experienced any event that could have damaged or did damage this
| |
| ! device, please explain what occurred and how you responded to the event. Include the dele of the event, a short d:::,44y, of the event, a C:::,ier, of any actual or
| |
| , suspected damage to the device, acdons taken to retum the devios to service, any l equipment or tests that were performed, and any organizations that assisted you.
| |
| L Please report all events that have occurred to this device. Please continue your answer on additionel blank pages as necessary.1.abel each blank page with your j company name and La ID number and device ID at the top of this page.
| |
| M..
| |
| A Page8
| |
| | |
| =
| |
| \
| |
| 4 Form Approved
| |
| ! OMB No. 3150-0152
| |
| ; EW 10/31/92 ID #
| |
| l U.S. Nuclear Regulatory Commission De@e D #
| |
| j Part V: Adduonal kWormadon h m m m m y '
| |
| ! (conumme) i
| |
| : J1. Has your company transfwred this device to another firm?
| |
| O Yes No 5
| |
| i j ~ l Fjou asemeRNI MD 00 h JI,)Wu AsWW completed the queedlens.se this deedsk )
| |
| 1 pgesse comenseis see asW desdeen '
| |
| it? -
| |
| l l
| |
| , J2. what was the date of the transfor? (MM/DD/YY) i
| |
| ' i J3. Name of Company to Whom I i
| |
| Deve Was Transferred J4.
| |
| Contact (First Name/Last Name)
| |
| J5. Contact Trtle l J6. Address i
| |
| J7, City J8. State J9. Zip Code J10. Phone Number Thank you for your assistance.
| |
| 1 i
| |
| Page 9
| |
| | |
| __ _ _ . _ m _ _. . _
| |
| t *
| |
| *
| |
| * Form Approved i
| |
| i
| |
| * OMB No. 3150-0152 Expres 10/31/92 i U.S. Nuclear Reculatory Commission o*
| |
| Survey of Generel Licensees: Focility Information Part I contains proprinted i' dorir.e" w i about your firm. Please review the proprinted information i
| |
| I in Part I and correct any incorrect informadon by crossing th, ,4 the incorrect information and writing in the correction above or to the right of the original information. Please complete any
| |
| { missing information by writing in the information in the space provided for that information.
| |
| Y.
| |
| I Part 11 of this survey asks for additionalinformation about your firm and your experience with the
| |
| }
| |
| tritiumpowered exit sign (s) identitled in Parts ill and IV of this survey. Please answer au
| |
| ! questions in Part il as they relate to your firm as a whole. Please fill in or circle the answer as
| |
| ! approprises.
| |
| i Part 1: Responslert identification (Please correct anyincorrect /rWbrmadon)
| |
| . A. Company Name 1
| |
| i B. Company Contact (First/Last j
| |
| Name) i C. Contact Title t
| |
| i 0
| |
| D. Address l
| |
| l 1
| |
| EW
| |
| : F. State l
| |
| G. Zip Code H. Phone Number
| |
| ~
| |
| i, j Part II: Ads 5tional Respondent Information i
| |
| Please tillin or circle tM answer as appropriate a
| |
| A. What is the principal business conducted at your facety?
| |
| l 1 - Aviation 5 - Laboratory O a - school i O 2- Entertainment O e - Marcafecturing o - state / Local Gov't O 3-Food and m ** O 10 - Other > wece O 4 - Hospital O 7 - office Building i
| |
| 1 i
| |
| 1 B. Done your firm have a copy of the General Ucense? Yes No 4
| |
| $ Page 1
| |
| | |
| t *
| |
| * ro,m Approv.o
| |
| ! OMB No. 31504152 l Expires 10/31/92 i U.S. Nuclear Regulatory Commission lo
| |
| * Survey of General Licensees: Facility information (continued)
| |
| Part 11: Additional R::;-:-Mat information (continued)
| |
| C. Is there a company representative who known the O Yes No General Ucones conditions?
| |
| D. is that person responsible for ensuring compliance with Yes No the General Ucense conditions?
| |
| E. Is your firm having any difficulty finding an authorized recipient to:
| |
| E1. Purchase any tritium-powered exit signs you no longer want? O Yes No E2. Dispose of any trftium-powered exit signs you no longer want? Yes No E3. Store any tritium-powered exit signs you no longer want? Yes No Wpeuamenered No se J guessesEfiat,and Eh
| |
| ?sesoguessen F/ '
| |
| E4. If your firm is having any difficulty finding an authorized recipient, please i!'
| |
| elaborate by describing who you contacted and describe the nature of the tilfliculty. !
| |
| Please continue your answer on additional blank pages as needed. Label each i blank page with your company name and the ID number at the top of this page.
| |
| i F. How many tritium-powered exit signs, held under a General Ucense, are now in secured storage that are not intended for use?
| |
| Page 2
| |
| | |
| Form Approved
| |
| { -
| |
| j OMB No. 3150-0152 j# Expires 10/31/92
| |
| ; U.S. Nuclear Regulatory Commission o*
| |
| J j Survey of General Licenseso: Exit Sign Information Parts ill and N contain proprinted inhinstui about tritium-powered exit sign (s) that NRC records indicatt were douvered to your s.rm under a General Ucons . Please review the proprinted infctmation in Parts ill and N and correct any incorrect information by crossing through the
| |
| ! incorrect information and writing in the correction above or to the right of the original ' Juisi.sf.ac.
| |
| Please complete any missing information by writing in the information in the space provided for that irWormation.
| |
| l
| |
| [ Part V of this survey refers to the exit sign (s) identitled in Part N. Please answer all questions
| |
| ! in Part V as they relate to the exit sign (s) identified in Part N. Please fill in or circle the answer i as =-; ;- upi':. (in allcases, Curien are a measure ofradioactMty, MMmeans month, DO means day, and YYmeans the hast two digits of the year.) '
| |
| ] Parte IN, IV, and V are repeated for each type of exit sign that the supplier reported it had l supp ged to your faculty. You have been provided with a complete Part III, W, and V for each j
| |
| type of exit sign. Each answer in Parte lit, N, and V should refer to the specific type of exit '
| |
| l sign identified in Part N of that stapled est of Parte lit, N, and V.
| |
| Port W: SuppNor (Manufacturw or Diebeutor) of Exit Sign (s) in Pert IV l (Please correct any incomsct Intbrmadon)
| |
| A. Supplier's -W License Number I B. Supplier's Company Name k l C. Supplier's Address i
| |
| l D. City 4
| |
| i E. State F. Zip Code
| |
| !. G. Phone Number i
| |
| i I
| |
| Part IV: ExR Slyi(s) latentification (Please comsct any locorrect frWormation) f l A. Exit Sign (s) Type i
| |
| i B. Supplier's Model Number i
| |
| C. Radionuc5de Source D. Quantity of radioactMty (Curies)
| |
| ]
| |
| j E. Number of exit sign (s) ht4d
| |
| ]. G. Date exit sign (s) shipped to you (MM/DD/YY) 4 l
| |
| Page 3
| |
| | |
| l .
| |
| Form Approved OMB No. 3150-0152 j
| |
| Expires 10/31/92 U.S. Nuclear Regulatory Commission 10 *
| |
| ! Past V: Ad Silonel Informellon About Exit Sign (s) Descrhed in Part IV l (Planee llN in or circle the answer as approprien) l PI aso snower all questions in PJ V as they rFate to the e.< .!;n(s) identlSed in Part IV.
| |
| ' 1
| |
| ; A. Were the exit sign (s) over checked Ibr defects after O Yee j receipt?
| |
| O No !
| |
| l O some but'not an exit i sign (s) checked l h hhY geesman Aatt so.: i
| |
| : quennen s [~ , -
| |
| A1. How many exit sigrw were checked for defects after receipt?
| |
| I j B. Did your firm install the exit sign (s)? Yes No 1 We installed some but not aN exit sign (s)
| |
| B1. Was the Specific Ucensee identitled in Part lit the same
| |
| .y p e,,,,,,,g ye,g,7 organization that installed this exit sign (s) at your site?
| |
| A Spec 6c License is issued to named persons soon
| |
| ,s.
| |
| Af; queenen S$8eV: !l
| |
| -:9 queenen^CF !'
| |
| apphcanone tHed pursuant to the regndeelone kt 10 CFR '
| |
| 30.31 through 30.36. Specmc Liceneeen are authorited to receke, store, demonstnme, and redistnibuso deMices containing sealed sources listed on their hcense (NRC Form 374). Yes No C. Have the exit sign (s) been relocated (moved within the O Yes No site and reinstaned) since initial installation?
| |
| some but not an exit signs relocated z.y.# year ansesesed No es,
| |
| '/ queegesA eA$p dei;c
| |
| , guseden'D '
| |
| C1. How many exit sign (s) were relocated?
| |
| C2. Did your firm relocate these exit sign (s)? O Yes No O we relocated some but not all exit signs Page 4
| |
| | |
| . . . . . - . . - . . . . - . . - - _ ..~ - _ --
| |
| i.
| |
| l i :
| |
| * Form Approved i
| |
| * OMB No. 3150-0152 3 Expires 10/31/92 j U.S. Nuclear Reculatory Commission o*
| |
| Port V: Adelional Information About Exit Sign (s) Descreed in Port IV
| |
| ! (conunued)
| |
| ! D. Has your firm experienced events that could have i Yes O No caused damage or did cause damage to any exit sign (s)r #ppassessed.Afeso
| |
| , gusanan SLatts.sek o
| |
| %gqueensa t? -
| |
| t
| |
| } D1. If your firm experienced any event that could have damm0ed or did damage any exit j sign (s), please explain what occurred and how you responded to the event. Include j the date of the event, a short description of the event, a -f::2,4^-6 of any actual or l
| |
| suspected damage to the exit sign (s), actions taken to retum the exit sign (s) to servloe, any equipment or tests that were performed, and any other organizations that assisted you. Please rorort all events that have occurred to the exit sign (s). Please l continue your answer on additional blank pages as needed. Label each blank with j your company name and the ID number at the top of this page.
| |
| 4 l
| |
| ! I t
| |
| l i s
| |
| i E. Do the exit sign (s) have:
| |
| l E1. A durable label? O Yes O No l E2. A legible (easily readable) label? O Yes O No l E3. A clearty visible (easily seen) label? Yes No l 1 - On the sign housing 2 - On the supporting frame l
| |
| F. Has your company transferred art; :%t sign (s) Yes O No to another firm?
| |
| f ,,
| |
| , , E' # '
| |
| .( ,.~+?^<
| |
| ~ ,
| |
| ' .-j ' ., ,
| |
| '4
| |
| _ . . . ...z ._,
| |
| j ' #ses asemeses see se[guessee 7, yev noe cornpassed me.gmessoas_se mese ese een(4 ;
| |
| J,
| |
| ; F1. How many exit sign (s) were transferred?
| |
| 1 l Ti;ank you for your assistance.
| |
| 1 Page5
| |
| | |
| l 1
| |
| r r
| |
| l i
| |
| APPENDIX B l I
| |
| SURVEY COVER LETTER 6 !
| |
| t
| |
| (
| |
| l I
| |
| \
| |
| w
| |
| _l i
| |
| 1 e 1
| |
| l l
| |
| 1 I,
| |
| i
| |
| ?
| |
| l l
| |
| | |
| m -m__ m . _m _ m m .
| |
| ._.__.__.._m.._ _ _._ _ ____ _ _ _
| |
| !e e
| |
| ; . e @ Glity
| |
| * + UNITED STATES j
| |
| l E
| |
| 7,
| |
| .i I
| |
| NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20666 i I
| |
| i July 1,1990 -
| |
| l ;
| |
| 1
| |
| ; Dear Sir / Madame-t j . This letter is to notdy you that your firm has been ctesm by the Nuclear Regulatory Commmaion 1
| |
| (NRC) to participate in a survey of general licensees for (pauging devices and laboratory equipment contaning radioactwo rnatenal. The survey is intended to confirm the placement, use, condition, and/or
| |
| ! disposstion of these devices for the purpose of evaluating the effectiveness of the current NRC regulations l (10 CFR Part 31, Section 31.5) portammg to general licensees.
| |
| 1 Each attached questionnaire describes a particular device (includog manufacturer or initial transferor, model number, type and quantity of radeonuclide, and date that you received the device) that Nucteer Regulatory Commission records indicate was shipped to you during the 1980's and is currently in your tw==aaman. (if you were shipped and currently poosess more than one such device, you should .,i recewe more than one questionnaire form.) Please confirm the proprinted information on each 1l T ^M-i- you receive or make any necessary changes if it is not correct. Then answer fgr gle ~'
| |
| particular device descrf;ed in each auestionnase ad of the other questions on that questionnaire. NRC records indicate that the pereco named in the address above le the contact person for this device or devices, if the listed contact person is not available, or if some other person might be better aisle to provide the requested ; A. c^': i, please review the detaled instructions included in the a - 1:- se.(s) and forward the =_ ^':-ine(s) to the appropriate respondent. Your failure to retum the complettd quesconnare(s) could result in your boeng selected for a followup onsite vielt to inspect the device (s).
| |
| - This survey has been approved by the Off co of Management and Budget it is being conducted for the Nuclear Regulatory Commission by ICF incorporated of Farfax, Virginsa. If you have any quaminns concommg the survey, caN the fotowing tot free number between the hours of 9:00 a.m. and 5:00 p.m.,
| |
| Eastem Standard Time: (800) 331-9212. The information that you provide in this survey wlN be treated
| |
| &ccv,G; 4 to conAdential procedures under the Privacy Act (5 U.S. Code $552a).
| |
| If you wish to dispose of the device (s) in this survey, or of any otive oevice containng radeonuclides that you hold under a general license and is currently under your control, you should contact the manufacturer of the device (s) directly, or the nearest Regional Office of the Nuclear Regulatory Commesion for assistance. For your information, a copy of those sections of the NRC's ruise and regulations portaining to you as a general licensee (Title 10 Code of Federal Regulations Parts 20,30, and
| |
| : 31) is also enclosed.
| |
| Please retum the survey wthin two wooks of receipt. Thank you for your assatance.
| |
| | |
| 1 ",
| |
| #f u:g%,
| |
| 1 UNITED STATES l
| |
| ! o
| |
| ! ; a NUCLEAR REGULATORY COMMISSION wAsa:Norom. o. c. 2 ossa 1
| |
| i
| |
| , Dear Sir / Madame.
| |
| Over three weeks ago your firm received one or more survey questionnaires from l Regulatory Commission containing radioactive requesting information about exit signs, gauging devices, byproduct material.
| |
| The survey is intended to confirm the placement, use, condition, and/or disposition of these devices for the purpose of evaluating the effectiven
| |
| , NRC regulations (10 CFR Part 31, Section 31.5) pertaining to generallicensees. Accor however, you have not returned a completed survey. ,
| |
| Each of the qucationnaires that you received describes signs, devices, or instruments t Regulatory Commission records indicate were shipped to you and are currently in your is extremely important that your firm return each questionnaire, so that the treatment an these could signs, result devices, in your or instruments being selected can for an onsite visit. be verified. Your failure to return the comp the survey, call the following toll-free number between the ho Standard Tirne: (800) 331-9212.
| |
| {
| |
| Please indicae your name, address, and telephone number, the "
| |
| identificmion number on the mailing label attached to the envelope for this letter, and leave description of your question. If you have recently returned your survey, it may have been in t this letter was mailed. If so, thank you for your assistance.
| |
| The information that you provide in this survey will be treated according to confidential pro under the Privacy Act (S U.S. Code $552a).
| |
| Management and Budget. This survey has been approved by the Office of Thank you for your assistance.
| |
| | |
| I i
| |
| 1
| |
| )
| |
| 1 l
| |
| APPENDIX C '
| |
| RESPONSES TO COMMONLY-ASKED HOTLINE QUESTIONS !
| |
| l l
| |
| i l
| |
| l i
| |
| 1 no
| |
| | |
| a u
| |
| RECOMMENDED RESPONSES TO COMMON QUESTIONS ON NRC SURVEY OF GENERAL LICENSEES FOR TRITIUM EXIT SIGNS
| |
| : 1. The proper person to answer the questionnaire is on vacation for two weeks?
| |
| Response: Please make certain that the que=tionnaire is filled out and returned as soon as the proper person returns. You will be contacted by phone, and you may receive a follow-up questionnaire if substantially more than two weeks elapses.
| |
| : 2. I don't have possession of the light; I only installed it for someone else.
| |
| Response: Please answer Question F and Question Fl. On a separate piece of paper, labeled with your fira name and the IDf at the top of the questionnaire, list the names and addresses of the persons for whom you installed the lights.
| |
| : 3. We don't have a Ceneral License.
| |
| Response: NRC doesn't issue written licenses for possession of exit lights (the devices that NRC is asking about in this survey] . Instead, NRC issues a license (called a specific l license) to the firm that manufactures the light [ device), !
| |
| and that firm is required to inforr the people to whom it )
| |
| ships lights [ devices] of their responsibilities. Those !
| |
| responsibilities are described in 10 CFR $31.5, and the terms of the regulation constitute the " General License."
| |
| You should have received a copy of the regulations in the same envelop as the rurvey questionnaire, and you may want j to review what they say, especially Section 31.5. i
| |
| : 4. What if we never received the d ivice?
| |
| Response: The exit signs that we are interested in should have a label. If you have checked for such'a label, and if you are certain that you don't have any exit lights that aren't connected to a source of electricity, then please write on the lines in Part IV on page 3 next to the sign information
| |
| ; "Did not receive this sign /these signs."
| |
| : 5. What if we received the device, but are no longer in possession of it; should we still complete the . survey in its entirety or simply fill out certain parts of it?
| |
| Response: You should fill out as much of the survey as you can. I l
| |
| : 6. If we are no longer in possession of the device, how much of the survey needs to be filled out?
| |
| Response: You should fill out as much of the survey as you can.
| |
| l
| |
| | |
| O 7.
| |
| We received the device, but never installed it; how much of the survey must we fill out?
| |
| Response: You should be able to fill out all of Parts I, II, III, IV, and V A., B., D., E., and F. -
| |
| 8 We don't know the answers to certain questions (i.e. the Installer's specific license number); should we just leave that space blank? ,
| |
| Response: Yes.
| |
| : 9. What is the Privacy Act and what are my rights in regard to confidentiality?
| |
| Response: The Privacy Act governs who can see records maintained by the federal government. Under the Privacy Act, the NRC will be able to see the results of this survey, including your responses, but the responses won' t be given to anyone else in a form that allows them to identify your answers individually. A final report on the results of the Survey may be prepared, but if it is it will not name individual firms.
| |
| : 10. Will any actions be taken against me if the answers to the survey reveal
| |
| , that my company is in violation of the regulations?
| |
| Response: NRC is conducting this survey in part to see if it needs to l
| |
| ! change the regulations. If your responses show that a !
| |
| potentially dangerous situation may exist, NRC will contact you in order to assist you in fixing the situation.
| |
| 11 What is tritium (or any of the other radionuclides)?
| |
| Response: Tritium is nuclear byproduc6 material. Its possession and
| |
| < use is licensed by NRC. You should contact the specific licensee from whom you obtained the exit light (device) for an explanation of its characteristius end detailed instructions on the care of the tritium exit light.
| |
| : 12. How can I determine if the label on the device is clearly visible?
| |
| Response: The label should be in a place where you can see it easily, 4
| |
| without removing the light from where it is installed. The purpose of the label is to inform people that the light contcins byproduct material, so it should be situated so that it can do that.
| |
| : 13. Does Hurricane Hugo count as an event which could have caused damage to
| |
| : the device (or other situations or events)?
| |
| Response: NRC is interested in making certain that proper precautions are taken, and that the specific licensee is notified in situations when the light might have been damaged. Please
| |
| | |
| l provide details of events that did raise a question in your mind about whether the light was damaged.
| |
| : 14. I lost the first two sections of the survey; can you send me a new copy? j
| |
| : j. Response: Yes. Pli.ase allow a few days F. d elivery.
| |
| l j 15. Is the survey optional? !
| |
| I L l Response: One of the conditione for having byproduct materir.1 is that l
| |
| ; you provide reasonable information about'how you are 2
| |
| treating the material in your possession. The Office of l j
| |
| Management and Budget, which has reviewed NRC's request to l conduct this survey, has given its approval for the survey. t
| |
| ' You are not subject to any penalty for not returning the )
| |
| survey questionnaire, but the NRC may send out an inspector !
| |
| to obtain the same information that is asked for on the l j survey form.
| |
| : 16. Where can I find the serial number of my device?
| |
| )
| |
| Response: Exit lights may not have an obvious serial number. You should contact the supplier if you want to know the number.
| |
| However, you are not required to provide the serial number-as an answer on the questionnaire.
| |
| ; For gauges and analytic devices,.the serial number should be !
| |
| j plainly marked on the housing or frame of the device. ,_
| |
| : 17. Does a given situation count as " difficulty" in finding an authorized l recipient; if so, will anything be done to offer us assistance in i l resolving the difficulties? !
| |
| ! Response: By difficulty, we mean a situation in which you have more
| |
| ; than usual trouble in finding someone to do what you want.
| |
| If you have such difficulty, your NRC Regional Office will I be able to assist you. The location and phone number of NRC j l Regional Offices are: i i
| |
| l [ Insert) , ,
| |
| 4 18. We were sent the wrong survey (i.e. name and address on envelope do not
| |
| ; match those on the cover letter); what should we do?
| |
| ; Response: If the survey questionnaire information is correct, please
| |
| , disregard the cover letter. If the questionnaire is not correct, please indicate that in the space provided in a
| |
| Part I by writing in the words "Sent to wrong firm." Please
| |
| } include your own name and address. ,
| |
| ! 19. How and/or why were we chosen by the NRC to participate in this survey?
| |
| 4 Response: Suppliers of exit signs are required by law to report to NRC l' jt
| |
| ! l t
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| 4 m y .- _ - . , _ _ - _ - . , .
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| | |
| b o
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| every three months the names and addresses of the firas/ persons to whom they have shipped exit signs containing byproduct materials. NRC chose survey respondents from those quarterly reports.
| |
| ! h t is your name and phone numb.r?
| |
| Response: Give your name. The phone number is 800-331-9212.
| |
| : 21. What Office are you with at NRC7
| |
| | |
| ===Response===
| |
| I an employed by ICF Incorporated, who has been hired by NRC to conduct the survey for NRC. -
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| )
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| 1 l
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| l i
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| o
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| J 4
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| I i ;I i
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| i 1,
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| i TECHNICAL LETTER REPORT:
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| ; TASK 7, FINAL REVIEW OF THE 1987 REPORT BY OAK RIDGE ASSOCIATED UNIVERSITIES, i " IMPROPER TRANSFER / DISPOSAL SCENARIOS FOR GENERALLY LICENSED DEVICES" 1
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| 4 -
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| i l
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| I 1
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| I
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| ~
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| NRC JOB CODE L2536 ,
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| PNL No. 20278 1
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| D. J. Strom!
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| R. L. Hill!
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| J. S Dukelow2 G. R. Cicottel 2
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| 1 Health Protection Ikpartment _ !I Nuclear Systems and Concepts Department Pacific Northwest Iaboratory , i Richinad, Washington 99352 June 3,1994 Prepared for the U.S. Nuclear Regulatory Comminnion Under Contract DE-AC06-76RLO 1830 0
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| Enclosure / 10 i e ^* 17 0 'l l 7 lyy p p. r
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| i i
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| 1.0 EXECUTIVE
| |
| | |
| ==SUMMARY==
| |
| | |
| l Task 7 of the project " Review of Improper Transfer / Disposal Scenarios for Generally i Licensed Devices Study" requires that, "after Tasks 5 and 6 are completed, PNL will prepare j a draft report based on the results of the individual reviews that documents its findings, l conclusions and recommendations. The report will present a critical evaluation of the 1
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| methods, data and assumptions used in the ORAU [ Oak Ridge AssociatcJ Universities; now Oak Ridge Institute for Science and Engineering] study (Stabin et al.1987). Any limitations i or inadequacies will be described, as well valid insight and conclusions reached by the
| |
| : ORAU team. The report will specifically address the suitability of the 1987 ORAU report as
| |
| ! a basis for revising regulatory requirements or guidance in 1993. It will provide the Staff i with reenmmendarians on how the contents of the ORAU report can be used in formulation
| |
| ; of policy or regulation, as well as additional facts and information to support their
| |
| ! decisions." "Within 2 months of receiving NRC staff comments on the draft report, PNL
| |
| ; will respond to the conunents and provide a final report to the Project Manager..."
| |
| ! As part of Task 5 of this project, Roger Cloutier, Kermit Paulson, Mike Stabin, and Evelyn
| |
| ; Watson (the "ORAU Team") attended a meeting with Robin Hill, George Cicotte, Jim
| |
| : Dukelow, and Dan Strom of the Pacific Northwest Laboratory (PNL) on October 5,1993, in l Oak Ridge, Tennessee. The meeting was a very productive discussion with the original l authors and team of the 1987 ORAU report. All PNL questions were satisfactorily resolved, i as is documented in the Appendices to this fm' al report.
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| ^
| |
| The PNL reviewers conclude that the 1987 ORAU Report pmvides a good start on assessing
| |
| : worst-case consequences of improper transfer and disposal scenarios for generally licensed l devices. For use as a basis for regulatory decision making, the principal shonenmir=s of the l ORAU Report identified by the PNL review are:
| |
| 1 l The 1987 ORAU Report does not include probabilities of the scenarios occurring on a j per source, per year basu. '
| |
| { 'Ihe 1987 ORAU Report does not include a complete enough enumeration of the numbers of sources, and the distributions of source activities, in each category.
| |
| * The 1987 ORAU Report does not include the probabilistic dutnbutens of outcomes i'
| |
| ! (as opposed to the worst case outcomes) needed to realistically assess the probable human health consequences of such scenanos.
| |
| 'Ibe original ORAU Team was constrained from considering many of the above issues by the i limited scope of work of their project. The PNL reviewers repeat that the ORAU Report is a good start on the collection of data to support regulatory decision making.
| |
| 4 i As in the preliminary PNL review, the PNL reviewers conclude that the 1987 ORAU report 1
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| is not an adequate basis for 1993 regulatory decision makmg, and that members of the NRC i s ,
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| l l 1 l
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| l l
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| l Staff may need to use additional facts and information.
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| 2.0
| |
| | |
| ==SUMMARY==
| |
| OF CHANGES SINCE 1987 Since 1987, there have been changes in several areas that impact the current relevance of the 1987 ORAU Report. These include changes in recommended limits on dose to the public, significant changes and developments in probabilistic dose menenament methodologies, and changes in dosimetric quantities and models that affect risk assessments.
| |
| 2.1 CHANGES IN, AND CREATION OF, PUBLIC DOSE LIMITS The original work under review in this project is referred to as the "ORAU Report" (Stabin et al.1987). Since that report was prepared, NRC has instituted a dose limit for the public that is in concert with recommendations of radiation protection advisory groups.
| |
| For doses to the public (when such doses arise from licensee activities), the NRC has implemented a limit of 0.1 rem (0.001 Sv) total effective dose equivalent (TEDE) per year (10 CFR Part 20.1301(a)(1)). In addition, the NRC now specifies that the provisions of the U.S. Environmental Protection Agency's 40 CFR Part 190 apply to licensee activities.
| |
| Parenthetically, it is noted that the U.S. Department of Energy has lowered its limits for exposure to the public (10 CFR 835) to 0.1 rem (0.001 Sv) TEDE.
| |
| Since 1987, both the NCRP (1993) and the ICRP (1991) have lowered their recommended l' i
| |
| limits for members of the general public, to the same value of 0.1 rem (0.001 Sv) for the quantity efective dose, which is similar to TEDE although not identical. 'Ibe NCRP anxi ICRP recommendations are based on new risk findings of the National Academy of Sciences !
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| (NAS 1988,1990) and the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR 1988). ;
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| With the establishment of a limit on TEDE to members of the public of 100 mrem (1 mSv, 10 CFR 20.1301(a)), or 2% of the occupational dose limit, there is less margin for error in i dose assessments for improper transfer and disposal of generally licensed devices. 'Ihe new, lower 10 CFR 20 limits for the public should be used in assessing impacts of improper transfer or disposal of generally licensed devices.
| |
| 2.2 CHANGES IN PROBABILISTIC RISK METHODOLOGY Since 1987, many changes have occurred in probabilistic risk methodologies. Recent summaries of these technic are provided by IAEA (1989), Finkel (1990), and Morgan and Henrion (1990). In addition, the advent of user-friendly Monte Carlo simulation software for probabilistic health risk analysis, such as Crystal Ball ("Decisioneering, Inc., Denver, CO),
| |
| makes it feasible to perform probabilistic risk assessments for this kind of work.
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| 2
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| i i
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| 1 i 4 I
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| Distributions of source activities, and distribution.s of consequence severities, and probabilities of incidents occurring should be used to predict likely outcomes, with worst
| |
| ; case outcomes being found in the ;xtreme value.; of the result- distributions.
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| ?
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| 2.3 CHANGES IN DOSIMETRIC QUANTITIES AND MODELS l
| |
| : l
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| , Since 1987, quar tities such as effective dose or total effective dose equivalent has replaced
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| ; " dose equivalent" in most radiarian protection recommendations and erandards. Many l improvements in internal dosimetry have occurred, as detailed in Appendix B.
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| l
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| ; 3.0 THE ADEQUACY OF THE 1987 ORAU REPORT AS A BASIS FOR 1993 !
| |
| l DECISION MAKING l i
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| : The 1987 ORAU Report, in the judgment of the PNL reviewers, is no longer adequate for
| |
| ! 1993 decision malcing. For use as a basis for regulatory decision mairing, the principal i j shortcomings of the ORAU Report identified by the PNL review are:
| |
| i d
| |
| The 1987 ORAU Repcrt does not include probabilities of the scenarios occurring on a j per source, per year basis.
| |
| 1 i The 1987 ORAU Report does not include a complete enough enumeration of the I i numbers of sources, and the distributions of source activities, in each category. 1 l The 1987 ORAU Report does not include the probabilistic distributions of outcomes i
| |
| (as opposed to the worst case outcomes) needed to realistically assess the probable i human health coamm** of such scenanos.
| |
| ! 'Ibe original ORAU Team was constrained from considering muy of the above issues by the i limited scope of work of their project. The PNL reviewers repeat that the ORAU Report is j a good start on the collection of data to support regulatory decision mairing.
| |
| i Additional shortcomings include I
| |
| * changes in dose quantities (e.g., the inhuductios of effective dose equivalent from external irradiation) and in regulatbus (e.g., public limits on total effective dose
| |
| { equivalent) have taken place; and i
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| j
| |
| * several pa**atially significant scenarios, such as an intact source out of a shield, and
| |
| : potentially significant consequences, such as doses to workers (rather than the public),
| |
| [ have been omitted.
| |
| i
| |
| ! 'Ihese topics are supported in individual reviewe cr omments in the Appendices to this report.
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| ? .
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| 4
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| ! 3 i
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| a i
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| m _ , __ _ - . . _ . . - -_ _
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| 3 1
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| I
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| , 4.0 ADDITIONAL FACTS AND INFORMATION THAT MIGHT BE USED BY THE NRC STAFF TO DECIDE ON THE NEED FOR REGULATORY ACTION As m the preliminary PNL review, the PNL reviewers conclude that the 1987 ORAU report is not an adequate basis for 1993 regulatory decision rnaking, ano that members of the NRC Staff may need to use additional facts and information.
| |
| 4.1 AREAS IN WHICH ADDITIONAL FACTS AND INFORMATION ARE NEEDED NRC Staff should consider
| |
| * the annual rate.r of incidents involuing improper transfer or disposal by source category. Rate suentments require improved data (from both NRC-regulated and Agreement States) on numbers of sources, numbers of incidents, and dosimetric consequences (both individual and collective) of incidents;
| |
| * probability distributions of severity and of occurrence for various accident scenarios.
| |
| There should be nueuments of the magnitudes of doses and the sizes of the exposed populations that are likely to result from each instance of improper transfer or disposal of these devices. Such assessments can be based on historical incidents such as the Juarez, Mexico; Goiinia, Brazil; Korea-to-USA; and Indiana, Pennsylvania incidents. These distributions for various accident scenarios are needed to develop both the individual and collective dose estimates;
| |
| * a state-of-the-art probabilistic risk annenament with predictions of the impact of !
| |
| improper transfer and disposal scenarios on individual and collective TEDE. Such a ;
| |
| risk nueument should include " worst case" scenarios only as limits of distributions -
| |
| on a probabilistic basis, not as simple point estimates;
| |
| * the impact of proposed changes in regulations on the benefits and economics of use of generally licensed devices, and on the reduction in risk to users and the public. Such namemamenen are optimization studies or regulatory impact analyses; and
| |
| * the need for much better data on mimhers of devices. For a complete risk analysis, no assimqptions should have to be made about the numbers of devices, their isotopes, their activities, Device Codes, design, and date placed into service: these numbers should be used directly or in categories with sufficient detail (e.g.,1600 137 Cs gamma gauges of design XYZ placed in service at a rate of 100 per year beginning in 1978...) to perform the risk analysis.
| |
| 4.2 NEW WORK BY PNL REVIEWERS FOR DISTRIBUTIONS OF ACCIDENT CONSEQUENCES Pacific Northwest laboratory reviewers have provided additional information in three areas.
| |
| 4
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| l l
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| 1 l i
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| i 4.2.1 Incident Rates on a Per Source. Per Year Basis
| |
| ; Appc..ciix E c mtains the Revised Task 3 Technic 1 Letter Repom Evaluation Of Historical j Scaled Source Device Experience. This report incorporates additional data provided by j NRC/NMSS on scaled source registrations. Incident rates, on a per source, per year basis,
| |
| ; can be calculated as detailed in Appendix G of this report, using data from Appendix E of i this report.
| |
| 4.2.2 Proposed Probabiliatic Framework for Risk Analysis l Appendix F contains the Revised Task 6 Technical letter Report on the Development of Additional Probability and Risk Information. A proposed framework for risk analysis in .
| |
| 1 included in that report. This famework could be used to support regulatory decision !
| |
| j makmg. 'Ihe decision makers would have to make judgements regarding acceptable levels, ,
| |
| i expressed in probabilistic terms, of collective effective dose equivalent, individual total II l effective dose equivalent, and individual local (or skin) dose equivalent for each category of I sources analyzed.
| |
| i 4
| |
| ! For example, one category of source may have a once-in-20-years probability that a member i of the public may receive a TEDE in excess of 0.1 rem from the practice of generally i licensing such sources. Another category may have a higher probability, or even a virtual certainty, of one or more members of the public exceeding this 0.1 rem TEDE every year, l but with a very large benefit to other members of the public. These are admittedly tough
| |
| ! decisions, but, in the view of the PNL review team, the pr.:babilistic information on which to :
| |
| i base them should be made available to the decision makers. l l !
| |
| 4.2.3 Exposure Probabilities in Accidents i
| |
| i The revised Task 6 Report in Appendix F contains analyses of 42 historical accidents l involving external doses to workers or the public, and gives distributions of Time-and-
| |
| ; Proximity Factors for individual whole body, individual local (or skin), and collective doses.
| |
| ! Distributions of Time-and-Proximity Factors from historic:.1 accidents can be used in l probabilistic risk analyses for both whole-body and local irradiation from external sour.es. '
| |
| l An analysis of 42 accidents for which source identity and strength are available show that the l average accident victim gets a whole body dose equal to that from being at 1 meter from the accident's unshielded source for an hour. The average accident is characterized by a value ;
| |
| of 46 hours at a meter. In other words, the population-weighted average is about I hour at a meter, while the accident-weighted average is 46 hours at a meter. Clearly, the accidents
| |
| ] with large numbers of victims (e.g., Goiinia and Juarez) dominate the former average. The maximum value seen for whole-body doses is about 700 hours at a meter (from the 1972 j Texas child-abuse case). The average, geometric mean, and maximum values for local
| |
| ; irradiation are 3100, 60 and 24,000 hours at a meter, respectively.
| |
| i I i 5 l l :
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| i i
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| . . _ - . , , ..m, . . _
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| b 1
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| i i
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| i Such distributions should be used in probabilistic risk analyses to determine likely
| |
| ] distributions of risks or doses from improper transfer and disposal scenarios for generally i licensed devkes.
| |
| i 1
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| j Accidents that were terminated due to the appearance of clinical symptoms of acute irradiation have less value for risk analyses than accidents that were terminated by other means, or never terminated.
| |
| ] 'Ibe current NRC Office of Analysis and Evaluation of Operational Data (AEOD) Nuclear j Regulatory Event Report (NRER) incident database does not contain the kinds of information j neMed to perform analysis of accidents for Time-and-Proximity Factors. It is recommended
| |
| ! that the darshnee either be modified to include this information, or a separate database be created. There is a great deal of work to be done to refine these preliminary analyses,
| |
| { extend them to additional accidents, and develop the logical framework for extrapolating to l other kinds of sources and scenarios.
| |
| i l For intakes of radioactive materials, 60 historical accidents have been characterized by i distributions of individual Fractions-Taken-In, that is, the fraction of the activity in the f original source that was taken in by each individual involved in the accident. For most i accidents involving radioactive sources, the fraction taken in is zero (0). For one accident, j that at Goiania, Brazil, in 1987, hundreds of persons had intakes, including 194 cleanup 4 workers and at least 77 members of the public. The former were characterized by Fractions-j Taken-In on the order of 10-12, and the latter by Fractions-Taken-In averaging 5 x 104 .
| |
| 'Ihese values are far below the maximum possi:>le value of 1, ar of the value of 0.3 adopted
| |
| ! as a worst case in the ORAU Report. In no accident has a value greater than 0.01 been j seen.
| |
| 5.0 PRELIMINARY RISK ANALYSIS i
| |
| i A full implementation of the risk analysis described in Section 2 is beyond the scope of work
| |
| ! of the cunent project. However, a sample risk analysis for 137Cs gamma gauges is given in l Appendix O.
| |
| j
| |
| | |
| ==6.0 CONCLUSION==
| |
| S i
| |
| : The PNL reviewers conclude that the 1987 ORAU Report provides a good start on assessing
| |
| ; worst-case consequences of improper transfer and disposal scenarios for generally licensed
| |
| ; devices. For use as a basis for regulatory decision making, the principal shortcomings of the
| |
| ! ORAU Report identified by the PNL review are:
| |
| 4
| |
| ; The 1987 ORAU Report does not include probabilities of the scenarios occurring on a j per source, per year basis.
| |
| i
| |
| ~~
| |
| I 6 l
| |
| i 1
| |
| | |
| I The 1987 ORAU Report does not include a complete enough enumeration of the
| |
| ; numbers of sources, and the distributior.s of source activities, in each category, j -
| |
| The 1987 ORAU Report does not include the probabilistic distributions of outcomes (as opposed to the worst case outcomes) needed to realistically assess the probable
| |
| ! human health consequences of such scenarios.
| |
| i
| |
| 'Ibe PNL reviewers conclude that the 1987 ORAU repert is not an ava basis for 1993 l regulatory decision making, and that the NRC Staff should consider additional facts and l
| |
| : information in the areas described above. In particular, the use of worst case scenarios with ;
| |
| , unrealistically high exposure factors tends to make the consequerres of improper transfer and !
| |
| : disposal seem worse than they would probably 've.
| |
| An additional conclusion is that existing databases are not a#quate for performing modern, l
| |
| ;_ probabilistic risk analyses. It would be desirable to collect and store in database format !
| |
| j dosimetric information (including quantitative measurements or estimates of intakes) for all j individuals involved in accidents. Such a data base would have to have an entry for each exposed individual, rather than simply one line of data per accident.
| |
| ! .7.0 ACKNOWLEDGEMENTS
| |
| ) The PNL reviewers are pleased to acknowledge the help and support of numerous colleagues
| |
| ; in the preparation of this report. At NRC headquarters, significant technical help and
| |
| { feedback has been given by Steven L. Baggett and Sterling Bell. The ORAU team of authors
| |
| : of the 1987 report, Roger Cloutier, Kermit Paulson, Mike Stabin, and Evelyn Watson, were very helpful in discussions held concerning that report. At PNL, original project manager ,
| |
| j James R. Jamison and technical contributor Peter C. Olsen have helped, as have techmcal reviewers Paul S. Stansbury and Eva Eckert Hickey. We are grateful to Betty Anderson and j' Rebecca Webster for clerical support. :
| |
| | |
| ==8.0 REFERENCES==
| |
| | |
| 10 CFR 20. 1991. U. S. Nuclear Regulatory Commission (USNRC), " Standards for Protection Against Radiation." U.S. Code of Federal Regulations. 56 FR 23360-23474.
| |
| Amended and corrected 56 FR 61352 (3 December 1991) and 57 FR 57877 (8 December 1992).
| |
| 10 CFR 30. 1993. U.S. Nuclear Regulatory Commission (USNRC), " Rules of General Applicability to Domestic Licensing of Byproduct Material." U.S. Code of Federal Regulations, February 26,1993.
| |
| 10 CFR 31. 1992. U.S. Nuclear Regulatory Commission (USNRC), " General Domestic Licenses for Byprodu-t Material." U.S. Code of Federal Regulations, November 30,1992.
| |
| 7
| |
| | |
| 1 10 CFR 32. 1992. U.S. Nuclear Regulatory Commission (USNRC), " Specific Domestic Licenses to Manufacture or Transfer Certain Items Containing Byproduct Material." U.S.
| |
| Code of Federal Regulations, November 30,1951.
| |
| 10 CFR 835. 1993. U.S. Department of Energy (DOE), " Occupational Radiation Protection." U.S. Code of Federal Regulations. Federal Register, in press.
| |
| 40 CFR 190. U.S. Environmental Protection Agency (EPA). U.S. Code of Federal Regulations.
| |
| Ayers, J. 1985. " General License Study - Analysis of Hazard." Internal memo to Division Director through channels, March 1985, attributed to J. Ayers. 8 pp. Washington, DC:
| |
| Scaled Source Safety Section, U.S. Nuclear Regulatory Commission.
| |
| Baggett, S. 1987. General License Study Repon. vii + 127 pp. Washington, DC: Sealed Source Safety Section, U.S. Nuclear Regulatory Commission. j Dean, C. M., M. S. Lawrence, and H. D. lester. 1991. Repon on Survey of Generc:
| |
| Licensees Under 10 CFR 31.5. NRC FIN D 2554-0. ICF Inc., Fairfax, Virginia.
| |
| Finkel, A. M. 1990. Confronting Uncenainty in Risk Management. A Guidefor Decision- ,
| |
| Makers. Center for Risk Management, Resources for the Future, Washington, DC. j International Atomic Energy Agency (IAEA). 1989. Evaluating the Reliability of Predictions Made Using Environmental Transport Models. IAEA Safety Series No.100, i IAEA, Vienna, Austria. I International Atomic Energy Agency (IAEA). 1992. Applicraion of Eremption Principles to ;
| |
| the Recycle and Reuse ofMaterialsfrom Nuclear Facilities. IAEA Safety Series No.111-P- i 1.1, IAEA, Vienna, Austria.
| |
| International Commission on Radiological Protection (ICRP). 1977. Recommendations of j 1
| |
| the International Commission on Radiological Prctection. ICRP Publication 26, Pergamon Press, Oxford. I International Commission on Radiological Protection (ICRP). 1991. 1990 Recommendations I of the International Commission on Radiological Protection. ICRP Publication 60, Pergamon Press, Oxford.
| |
| Morgan, M. G. and M. Hention. 1990. Uncenainty. A Guide to Dealing with Uncertainty in Quantitative and Policy Risk Analysis. Cambridge University Press, New York. .
| |
| 8
| |
| | |
| National Academy of Sciences (NAS), Naticnal Research Council. 1988. Health Risks of Radon and Other Internally Deposited Alpha-Emitters: BEIR IV. National Academy Press,
| |
| "~ashingtor., DC.
| |
| National Academy of Sciences (NAS), National Research Council. 1990. Health Efects of Exposure to Low Levels oflon zing Radiation: BEIR V. National Academy Press; Committee on the Biological Effects oflonizing Radiation, Washington, DC.
| |
| "ORAU Report:" see Stabin et al.1987.
| |
| . Stabin, M., K. Paulson, and S. Robinson. 1987. Improper Transfer / Disposal Scenariosfor Genemtly Licensed Devices. "The ORAU Report" produced under NRC FIN B0299. Oak Ridge Associated Universities, Oak Ridge, Tennessee.
| |
| Unger, L. M., and D. K. Trubey. 1981. Specipe Gamma-Ray Dosc Constantsfor Nuclides In;portant to Dosimetry and Radiological Assessment. ORNL RSIC-45. United States Government Printing Office, Washington, DC.
| |
| United Nations Scientific Committee on the Effects of Atomic Radiatica (UNSCEAR).1988.
| |
| . Sources, Efects, and Risks ofIonizing Radiation. United Nations Pchlications, New York.
| |
| l l
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| l l
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| r l
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| i l
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| l i
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| i l
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| 9
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| _ l
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| _. - -. -- , ..- = _ - - . - - - . - - . _ _ - _ - - _ - _ - - , ~ -
| |
| i i APPENDIX A:
| |
| [ FINAL REVIEW OF 1987 ORAU REPORT BY D.J. STROM, INCORPORATING COMMENTS FROM VISIT TO ORAU TEAM, OCTOBER 4,1993.
| |
| cor this appendix, the following conventions are used:
| |
| r The original reviewer's comments make in the Task 4 draft letter report are shown in italics.
| |
| i The comments and results received during the n~ ting with the authors of the 1987
| |
| ; ORAU report are given in standard text format following specific Task 4 comments.
| |
| 1 A.I.
| |
| | |
| ==GENERAL COMMENT==
| |
| S
| |
| [ l
| |
| \
| |
| : 1. :l Efective dose equintent (EDE), committed efectiw dose equivalent (CEDE: which is usedfor internal doses), and total efective dose equivalent (TEDE) should be used, j l i
| |
| not simply dose equivalent. The ORAU team was in agreement on this point.
| |
| ; 2.
| |
| Which sources are normalform, and which are specialfann (IAEA/ DOT i
| |
| ! classipcation)? This afects transpon andfate of radionuclides in many accident j scenarios. This should be determined for sources in future analyses. ORAU team l did not consider this. However, Dodd et al.1989 have considered this as important.
| |
| * l
| |
| : 3. i Probabilistic risk assessment methodologies have changed dramatically since 1987.
| |
| \ \
| |
| This reviewer would suggest using makmg better use of 3istorical accidents to evaluate
| |
| \
| |
| \
| |
| orobabilities of dose relationshipsfor accidents. If risk is depned as (probability) x (senrity), then probability distributionsfor senrity andforprobability of occurrence l
| |
| ; should be given for wrious accident scenarios. There should be assessments of how
| |
| \ likely a given scenario is to happen per source or device, and what kinds of doses to \
| |
| how many people are likely to result. Distributions of doses to individuals and the numbers ofindivutuals receiving dosesfor wrious accident scenarios are needed to 4
| |
| develop individual and collective dose estimates. The ORAU team made it explicitly I
| |
| clear that they had been requested not to consider the probabilities of accidents happening, but rather what the scenarios and con =;m would be if an incident of !
| |
| } improper transfer or disposal did occur. In our discussions with the ORAU team, !
| |
| j they made it clear that they understood a risk analysis to consist of three parts: 1) the '
| |
| i probability that an event will occur per source; 2) the number of soutres; and 3) what the consequences are if an event does om':r. They said that their Ltement of work )
| |
| i ;
| |
| limited them to the third part of this, even though they had addressed, to some extent,
| |
| ) the second part. l i
| |
| 1
| |
| : 4. 1 Specspc equationsfor the wrious models used are not present. This reviewer would i
| |
| ! prefer to see an equationfor each dose that is anived at. The ORAU Team thought *
| |
| \
| |
| this was a good idea. t 1
| |
| I A-1 I
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| l l 1
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| 1 l
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| ! \
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| l 5. Better assessments of the rates ofincidents per source or device, along with the 1 l numbers of sources and devices in use, .:re needed. The ORAU Team concurred I i wh)leheanedly, but made it clear that this vas outside t' 4 scope of work.
| |
| I i
| |
| l 6. 10 CFR 32.51(a)(2)(iii) provides thejust@cationfor the probabilistic approach in its i use of the w>rd "unlikely. " The intent of this regulation seems not to befocused on
| |
| , the worst case scenario, but rather on the likehhood of a given scenario. This
| |
| { sign @mntly afects the likelihood of source dupersal in some accident settings. There l l was considerable discussion with the ORAU Team concerning worst case versus j average case incidents. They insisted that the worst case analysis had been r=W. -
| |
| l
| |
| ; They felt that average cases and distributions should also be considered. They l
| |
| ! considered, for example, a " worst case" intake fraction of 0.3, with a " realistic" '
| |
| intake fraction in the range of 10 4 to 104 .
| |
| ' Specific Comments l l
| |
| : 1. p2 Generic scenarios: Why no " release"from incinerator, scrap dealer, and metal }
| |
| recovery on the generic drawing? Certain!y such releases have been seen, e.g.,
| |
| Juarez, Auburn Steel. This should be considered. !
| |
| i
| |
| : 2. p311 Relative radiotoxicity per unit activity given in Sqfe Handling of Radsonuchdes (IAEA 1973): obsolete and ordmal: mterni AUs should be used. Also, something along the lines of the derimtion of the wlues the A; 42systemfor ennsportation could be usedfor relative radiotoxicity. hsentially, the A1 (Special Form, not expected to duperse in an accident) mlue uns 9/r curies, where I' is the external exposure rate in R h d m2cjd. While this comment is true, the risk analyses do not really depend on the old IAEA classification; PNL reviewers concur.
| |
| : 3. p313 The "estimatedprobabilities are educated guesses ... " Was a Delphiprocess ,
| |
| used? How was consensus reached? The ORAU Team explained to us that for many needed data items, there was no way to scientifically get the numbers. They simply discueeM some valuer, and came to a group consensus on their values. Examples of Macard guesses are 12, p.3, 62.2; $2.3.2.2; $2.3.2.3. Many other unsupported data items were taken directly from Buckley et al. (NUREG/CR 1775; 1980).
| |
| : 4. p315 Table 4: the known number of cases of mishandled devices. This table doesn't list uggt of accidents, i.e., per source and per year. The incidents arefrom NRC and agreement state data. The numbers of sources arefrom NRC data. What is needed is the accident rate per source-year, by type of source. This would be the number of incidents reported in a category divided by likelihood of an accident being reported in that categoryperyear. Let denote the incident rate, per source and per year (sourced year d );
| |
| R(C,y)
| |
| C denote the source category, e.g., A-1; A-2
| |
| | |
| i ll i
| |
| k l y denote the year ofincident reponing and source use;
| |
| : n(C,y) denote the number ofincidents reponedfor that source categoryfor that year; d(C.a denote the number of sources of cangory C in use in yeary; and \
| |
| p(C) derwte thefraction ofincidents involving sources in a given category that are
| |
| , reponed (presumed to be independent ofyear).
| |
| ! ' Then i
| |
| =
| |
| n(C,yj) 1 R(C,yg)
| |
| N(C,yg)p(C)
| |
| ! I I l l and, letting R denote the average rate, I l E(C) =
| |
| { R(C,yj) . I
| |
| ; i-t
| |
| } Tofully address the rates at which accidents occur in each category, it would be
| |
| ; necessary to have values of the above variables by yearfor a period of time long enough to give con)fdence in the rates and theirfluctuations. In particular, the variable p(C) is problematic, since estimation of under-reponing is always dificult to ascertain.
| |
| In the 1987 ORAU report, it seems that the incidents reponed represent an unknown fraction ofincidents. Funkermore, it is not clear whether those incidents arefor all sources, agreement state sources, whatfraction of agreement states, since there may be some reporting bias. For exanple, data may not be availablefrom all Agreement States or may not be available in the same level of detail. The degree of extrapolation needed to get rates should be specijfed. Repons probably underestimate incidents, and the underestimation is probably worstfor smaller sources. Again, the ORAU team agreed that this was an important point, but that it was outside the scope of their work.
| |
| : 5. 9 41212.3 line 2: Using the " largest source:" will result in a high-biased estunate of risk. Worst-case analysis was eqwM by NRC in 1987, according to the ORAU team.
| |
| : 6. 9 412 92.3: Between " intact, shielded" (Case 1) and " wide dispersal" (Case 2) there should also be " source intact but unshielded. " The latter is a likely outcome, e.g., in the Indiana, PA 19 Ir incident and the bulk of the radiography incidents.
| |
| A-3
| |
| | |
| i 1
| |
| i 1
| |
| i Discussions with the ORAU Team led us to create Table A-1 of exposure potentials j for various scenarios. The question of whether the ingestion of an intact source is j ' internal" exposure is moot; ingestion may or may not result in uptake of rad.oactive
| |
| ; material from the gastrointestinal tract.
| |
| Table A-1. Exposure potentis.1 for external, contact, and internal irradiation for seven j general device and source scenarios.
| |
| ! WP e Potential j Scenario External Contact Internal s 1. Source inside device x
| |
| : 2. Source out of device or sh'elding compromised (e.g., xx shutter open) i l
| |
| : 3. Source in device and source seakmg x x l !
| |
| l 4. Source removed from device and source not leaking xx xx i
| |
| i 5. Source removed from device aral source leaking xx xx x l - 6. Source removed from device rmd source dispersed xx xx xx i
| |
| : 7. Source removed from device and intact source ingested x xx x 1
| |
| ? Tk. accidents involving the most people, such as the Juare, , Mexico accident in 1983-4 and the Goiania, Brazil accident in 1987, have been in scenario 6. Many fatalities and acute radiation injuries have resulted from scenario 2 and 4 ace'.lents :
| |
| : 7. 9 412 52.3: Anrage dose equal to 1/2 maximum (NUREG 1980) does not agree with history. Doses are likely to be lognormally distributed (e.g., uranium in urine, occupational doses (UNSCEAR 1977), etc.) 'lhe 1/2 value was taken from Buckley et al.(1980). l
| |
| : 8. 9 415: "... population dose equinients were not calculated. " Collectwe dose equinient can and should be cein I-*ed, using a probabiliatic basis. While there is concemfor the maximally-erposed individual, the ontall detriment under a linear, ,
| |
| no-threshold dose-response ^ 3pothesis is proponional to the collectin dose equimlent. \
| |
| The size of the population to be used in collectin dose cein>In' ions should be addressed. Collective dose was cons'utered for inc~meration scenarios.
| |
| : 9. 9 416: " Gamma ray dose constants were takenfrom Unger and Trubey..." All work should be in EDE rate coratants; see attached. For doses derindfrom ulues in roentgens, l' mlues in roentgens should be usedfor ce!n>Inring. If new risk factors (e.g., ICRP 1991, UNSCEAR 1988, NAS 1990) are used, then appropriaa: quantities and units should be used. This may make a significant difference for external A-4 f
| |
| | |
| i j
| |
| 4 l exposures to some low energy photon emitters.
| |
| v.
| |
| . 10. pp4-5: Mexican "Co Accident (Andrews 1163): The most highly exposed boy's exposure corresponded to 687 hours at I meter, a high ve.*y time-and-proximityfactor, l .
| |
| . with aurage being 446, geometric mean 408, GSD 1.7. Howewr, the ORAU Report l assunption (Exec. Summary) of 20 weeks at 100 cm (1 m) corresponds to 3360 h at a j meter. (Note: Generalization of high-dose accidents is limited because of the
| |
| \ canonth& that occurs due toJktahty. Presumably, with nofatalities, the source
| |
| ! would be therefor a wry long time.) See TASV 6, development of additional i probability and risk information. Both whole body exposures and localized (often extremity) exposures should be considered in a risk analysis, since the latter may i result in deterministic (formerly non-stochastic) effects even though the former may l not.
| |
| , 11. 9612: 'These wlues are given with the quahpcation that the listed values must be
| |
| } increased by 25-45 percent to accountfor electron production in the stainless steel
| |
| < walls assumed to encapsulate the source. " This " buildup" is notfounded in either
| |
| { experiment or theory. To venfy that there is no needfor this correction, one could perform Monte Carlo calculations using3the MCNP codegrphoton emissionsfrom
| |
| ? stainless steel encqpsulated sources of Cs, "Co, and Ir. Indeed, the correction that is needed is one to efectin dose equimlent rather than air kerma (" exposure is i l the ionization equimlery sf collision kerma in air" - Attix,1980) or " free-air dose. ' '
| |
| j Table A-2 shows the peicent (efectin dose equiwlent)/(ambient dosc equinient) ,
| |
| i l
| |
| 1
| |
| [i.e., EDE/H*(10)]for wrious nuclides, expressed as u %, by irradiation geometry. !
| |
| i l
| |
| i i
| |
| , i
| |
| : i
| |
| ?
| |
| i A-5 ,
| |
| i
| |
| | |
| . - . . . . . - - . - . ~ . . - . . - . . . - . . - . . - - . - - - -
| |
| I b
| |
| l
| |
| .l Table A-2. EDE/H*(10) by nuclide and irradiatson geometry using ICRP Si conversions of photonfluence vs. energy to EDE and H*(10) (Strom,1993).
| |
| Nuclide (AP) (PA) (ROT) (ISO) (LAT) l Cd-109 33.2 % 14.6 % 16.4 % 13.6 % 8.4%
| |
| 1 I-125 42.6 % 22.0 % 21.9 % 18.0 % 12.4 %
| |
| ; i-133 61.2 % 40.2 % 36.0 % 29.3 % 23.9 %
| |
| 'd-201 72.6 % 52.5 % 46.1 % 37.7 % 33.2 %
| |
| j l Co-57 78.9 % 61.6 % 53.2 % 42.9 % 38.8 %
| |
| t j Tc-99m 81.4 % 64.1 % 55.3 % 44.9 % 40.5 %
| |
| f I-131 86.8 % 73.7 % 64.9 % 64.4 % 51.7 %
| |
| 7 Ir-192 87.3 % 74.1 % 65.3 % 54.6 % 51.9 %
| |
| f Ra-226 87.5 % 78.6 % 70.7 % 61.7 % 60."%
| |
| Cs-137 87.8 % 77.3 % 69.0 % 58.7 % c/.2%
| |
| ! . - Al-26 89.7 % 81.9 % 73.7 % 64.8 % 64.5 %
| |
| .- Co-60 90.0 % 82.8 % 74.7 % 65.8 % ' 65.7 % ;
| |
| i Na-24 91.6 % 86.3 % 78.5 % 71.0 % 72.1 %
| |
| l
| |
| : 12. p513: Maximum contact time of 3 hours: where did this " hypothesis
| |
| * comefrom?
| |
| ORAU Team consensus arrived at this value.
| |
| : 13. p514: De use ofinternal dose assessments based on committed efective dose equinient methods ofICRP 30 is good. No comment.
| |
| : 14. pp5-6: Brodsky uns not ested (Broddy, A. 1980, ion Factors and Probabilities ofIntake of Material in 1% cess (or "Is 1 a Magic Nkunber in Health Physict?'). Haakh Phys. 39(6):992-1000.). Intake d>ould depend on mass inmived.
| |
| Intakes of signffcant masses are not plausible (e.g., M ***U and *'*D). This is addressed above under worst case scenarios.
| |
| : 15. 9 692.3.2: An important scenario is missing, namely, the scenario of an intact, but unshielded, source. This has happened repeatedly in industrial radiography settings.
| |
| See Table A-2, above.
| |
| : 16. p6 last 1: ne wlue of 'S devices per year" needs clanfication. Is 5 devices per year of each kind or 5 devices per year of all kinds? An ORAU Team consensus.
| |
| I 1?. p711: There is nojustificationfor the.wfue of a nearby population of 73,000.
| |
| 4 I I A-6 1
| |
| l l
| |
| | |
| I i
| |
| t Buckley et al., p. D-31.
| |
| : 18. 9 912: average dose half as large as maxii.uun dose? anjustified based on accident histories. Review NUREG 1980. Buckley et al.
| |
| : 19. pi4 incineration studies: this seems incongplete, but may be adequate. Some incineration references are missing, e.g., Hamrick'a & Watson's nork. Dere are doubtless other Ni1S reports or DOE reports, in addstion to the refereed literature.
| |
| Jim Tripodes, who has hosted the recent incmeration conferences, should be contacted. De question to be addressed is releasesfrom low-tech incinerators, not incinerators with high-tech, scrubbed efluents. The ORAU team did a fairly thomugh search.
| |
| In conclusion, many of the major shortcomings that this reviewer found in the ORAU Report stemmed from the assumption that it was to be a complete risk analysis, when in fact it was a worst-case ant. lysis starting from the assumption that a device had already gotten out of control. Other issues, as discussed above, remain unresolved.
| |
| o i
| |
| l 1
| |
| I l
| |
| l A-7
| |
| | |
| APPENDIX B: R.L. HILL'S REVIEW OF THE 1987 ORAU REPORT, !
| |
| " IMPROPER TRANSFER / DISPOSAL SCENARIOS !
| |
| FOR GENERALL) LICENSED DEVICES" l
| |
| For this appendix, the followin; conventions are used:
| |
| 1he original reviewer's commenta make in the Task 4 draft letter report are shown in italics.
| |
| The comments and results received durmg the meeting with the authors of the 1987 ORAU report are given in standard text format following specific Task 4 comments.
| |
| Major conclusions from this review are given in the subsection following the specific Comments.
| |
| New information pertaining to aspects related to internal dosimetry are given in a separate section at the end of this appendix.
| |
| B.1
| |
| | |
| ==GENERAL COMMENT==
| |
| S l
| |
| Gl. This M ent should be a complete, stand-alone docenent where the methods are Jkily descnbed in the docenent. As it stands now, NUREG/CR-1775 is referr\ed to in order to obtain descrrptions of many of the methods used in the study. The assumpnons usedfrom NUREG/CR . 773for determination of external dose should be reviendfor pppropriateness and consistency. For instsnce, on Pnge 4 of the NUREGICR report, it is stated that a pus *nt source is assumed, while in Appendix A, a line source is assumed. Also, in Appendix A of the NUREGICR report, the internal doses are calculated using dose conversionfactors based on ICRP 2 methodology, which is not the casefor the GED study.
| |
| 'Ibe authors of the 1987 ORAU report indicated that, while NUREG/C5 m5 was heavily relied upon, the internal dose calentations were done using ICRP 30 methodology.
| |
| G2. It qppears that a lot of "short-cuts" were taken in this study, i.e., referring to another document for methods and not including worker dose calculations, food pathways, or intruder-type scenarios. A more in depth analysis is needed in order to encorrgpass all probable scenarios that may lead to a public dose.
| |
| The authors of the 1987 ORAU report indicated that the include the worker doses as part of the doses to members of the public.
| |
| B-1
| |
| | |
| l t
| |
| i G3. Betterpresentation and sununary of the resulting dose estimates are needed. It wmtid ;
| |
| be much easier to reach conclusions on the study if all results were located m one location, such as a collectson ofsewal summary tabla 7 t
| |
| The ORAU authors agreed that a better 9-mary of the data would greatly enhance j the readability of the repon.
| |
| G4. The dsference between MEl and awmge individual, and maximum indivsdual and realsstic individual needs to be better defined.
| |
| The ORAU authors agreed that more realistic values for the MEI and average !
| |
| individual doses need to be used.
| |
| G5. 1%o inportant pathways / scenarios missingfrom this study that may han potentially large inpacts on the reported dose estimates arefood pathways and worker scenarios. ,
| |
| l The authors of the 1987 ORAU repon indicated that the worker doses were lumped in i as pan of the public doses; and that, since they were not expens in using the dispersion and groundwater models, they relied heavily with methods used in l NUREG/CR-1775, which did not include food pathways.
| |
| G6. Better estimatesforparameters used to model incineration are needed.
| |
| 'Ibe authors of the 1987 ORAU repon indicated that at the time the report was prepared, most, if not all, of the available literature on the topic was based on incineration of medical wastes. They used what information could be obtained from the available literature.
| |
| G7. More sp-to-date conputer codes are anilablefor modelirog the dosesfrom plumes, incineration and langitis (i.e., GENII, GENII-5, CAP 88-PC, etc.).
| |
| 'Ihe authors of the 1987 ORAU report they realize that they were not experts in using ,
| |
| the dispersion and groundwater models, and that they probably did not used the most ,!
| |
| up-to-date models for the 1987 time frame. They stated th:_t now they realize the '1 I
| |
| project would have benefitted from havir4 specialists in these areas involved on the project.
| |
| G8. An uncertamty and sensitivity analysis is needed since the mluesfor most parameters ured in this modeling exercise how wide or largely unknown ranges.
| |
| 'Ihe authors of the 1987 ORAU report stated that they were not asked to perform an uncertainty and sensitivity analysis for that report, and they probably could not have done such an analysis with the data and analytical tools available when the repon was being prepased.
| |
| B-2
| |
| | |
| l 9
| |
| Y a B.2. SPECIFIC COMMENTS
| |
| . Sl. Page i, last 1: It appearsfrom the information presauw l,cre, that thefour cases i considered in this reponfor internal doses are 30% intakefor inhalation, 30% intakefor ingestion, inhalation after incineration, and ingestion aber leachedfrom a landfil. Thefractional intake values l
| |
| i need to be substantiated.
| |
| I The authors of the 1987 ORAU report stated that the 30% values was taken as an i educated guess or a benchmark value. However, they could not provide data to support the selected value.
| |
| l S2. Page ii, 'last 1: A metal recycling scenario was only consideredfor '0Co. For a more o realistic assessment, any of the GLD materials considered in this stuay could possibly pnd their way to a metter. Thus, all nuclides involved for GLDs should be considered in the dose assessments.
| |
| The authors of the 1967 ORAU report agreed that the portion of the report dealing with recycling should be updated.
| |
| S3. Page ii, last 1: A justifcation (i.e., calculated dose estimates) is neededfor the l.
| |
| statement thatfor metal recycle, "... dose equiwntents received by l members of the generalpublic who purchase contaminatedproducts 1 would most likely not exceed 500 mremfyr (0.005 Sv/yr) in most cases". l This might not be the case since in IAEA Safety Series 111-P-1.1 (IAEA il 1992), where generic exposure andpathmsy analyses were usedfor recycle of steel, a lim l ting dose of 8.8E-5 Sv per Bq/g in the scrap (0.33 mrem per pCi/g) mis estimatedfor '0Co.
| |
| No response was given for this comment.
| |
| S4.Page1,Ch1: it is stated that " potential" scenarios are developed and assessments .
| |
| I prowdedfor "reakstic" and " maximum dose equiwalent to individuals" (MEI). However, in the current regulatory environment (for the NRC),
| |
| rulemaking is not necessarisy based on " worst-case" but rather on
| |
| " prudently conservative" assumptionsfor the dose assessments. To this end, both determuustic assessments (based on the most realistic data awxilable from literature, etc.) and stochastic uncenainty and sensitivity analyses are needed. The Latin Hypercube Monte Carlo methodfor uncertainty and sensitivity (U&S) analyses was published in 1964 by staf at the Sandia National Esboratory and is a possible approach to performing U&S analyses. .
| |
| I B-3
| |
| | |
| 1 i
| |
| i
| |
| ! The authors of the 1987 ORAU repon stated that they were not asked to perform an
| |
| ; uncertainty and sensitivity analysis for that report, and they probably could not have j done such an analysis with the data and analytical tools available when the repon was 4
| |
| being prepared.
| |
| } S5. Page 3,1st 1: The raaiotoxicity class @ cation from the 1973 (AEA :focument cited is l perhqps out of date. More recently, the IAFA has used thefollowing
| |
| ! class $antion which may be substituted: ayha emitters, photon emitters with large extemal dose conwrsionfactors (DCE), no-photon emission with moderate internal DCFs, and other low dose mdionuclides (IAEA Safety Series 111-P-1.1,1992).
| |
| 'Ibe authors of the 1987 ORAU repon stated that they used this radiotoxi::ity classifk:ation to narrow oown from 600 categoric., to 16. It was used because the authors were familiar with it at the time the report was being prepared.
| |
| S6. Page 3, 3rd 4: The probabilities assigned were " educated guesses". This brings up another reason to perform an uncertainty and sensitivity analysis (U&S) for the assessment. De upper and lower boundsforparameter mlues can be researched and used in the U&S analysis. This mry, a range of possible dose estimates can be reported instead ofpoint esti: nates.
| |
| See response to comment S4.
| |
| S7. Page 4, Sect.2.3: For the dose assessments, only ' intact device' and 'diapersed device our wide areas' were considered. Shouldn't damaged device with limited dispersion be considered since the concentrations amilablefor intake or exposure would be higher for limited dsspersion (i.e., less dilution)?
| |
| 'Ibe authors of the 1987 ORAU report agreed with this annessment S8. Puge 4, 4th 1: Since only seuralpeople are assumed to be ===~I to intact devices, no population doses are &I"sd. Is this ulid, or shouldpopulation dose be considered?
| |
| The authors of the 1987 ORAU repon contend that contact with an ' intact' source would limited to a very small number of people. However, as shown in the appendix to the report that deals with additions! risk information, data from actual accidents of mishandled sources indicate that a great many people can be exposed from an ' intact' source. (See Tables 3.3.1 and 3.3.2. of that appendix.)
| |
| 1 B-4 i
| |
| l
| |
| | |
| 1 l
| |
| ; i
| |
| ) !
| |
| 4 59. Page 5, last 1: It is wry hard tojus:sfy the intakefractions assumed in this report.
| |
| l Using DOT-based scenarios as a source of these mlues is not entirely -
| |
| \ i i
| |
| appropriate. More recent acidents or accidents inniving inproperly
| |
| \
| |
| handling or disposal ofspecs/cally licensed sealed sources can be used
| |
| { i to apdate the ' realistic' parameter mlues used. '
| |
| See response to S1, and the internal dosimetry section of the additional risk j information in App =h g. -
| |
| 1 ;
| |
| l 510. Page 6, 2.3.2:
| |
| 1 For extensinly damaged devices with wide dispenion and many people inntwd, thefollowing scenarios wre desciibed: Incinerator with l l release to the environment, incinrator with subsequent burial in landfil, metal recycling into consumerproducts, and metal recycling into construction materials. One wry sigmpcant set ofdose scenarios i that were lep out of this study were the dose to wrkers at the i
| |
| incinerators or metters. Generic exposure andpathmy analyses perfonned by PNLfor US DOEfor incineration of hazardous materials
| |
| )
| |
| andfor smelting associated with recycling and reuse of contaminated \
| |
| materials how shown that, in many cases, the limiting scenarios and
| |
| . doses werefound to be those specipedfor workers.
| |
| The ORAU authors stated that worker doses were considered as and lumped in with the doses to members of the public. ('
| |
| Sil. Page 6, 2.3.2, Information on incinerationfrom NUREG/CR-1775 ms used as last '
| |
| 1 basisfor this study. Assunprions that are needed are stated to be (1) \
| |
| the number of devices incineratedper year, (2) thefraction of activity l released by the incineration process, and (3) thefraction of activity released that escapes into stack emissions. Fin devices incinerated per year us chosen as the wlue used in this study. All other dose eni~Indons in the report are based on one device. De assumed mlues takenfcr the three mqforpammeters need to bejustyled, m*rin''y the "arbitrardy chosen" mlue of 5 devices meineratedper year. Aho, severalparameter mines that may haw a mqfor inpact on the resulting dose estimates that were not discussed are thi' thrsughput of the incinerator, thefeed conposition, the incineration temperature, and the pnalpopulation size considered.
| |
| The arbitrarily chosen value of 5 devices / year was chosen only as a benclunark value for this study. No data to support this selection was given.
| |
| B-5
| |
| | |
| l 1
| |
| S12. Page 6, 2.3.2, Dose calculation methods similar to thosefrom NUREGICR-1775 la:t 1 were used in this study. The methods stated in NUREGICR-1775 do not state thatfoodpathwys were considered in the analysis oflandfils, l thus it is assumed that they were not considered in this study either.
| |
| Other studies have calculated estimates of doses to the publicfrom radionuclides present in landfils, in some cases, the highest potential doses werefound to resultfrom intader scenarios and weU drilling scenarios some years aper a langfu has closed. Both external and l
| |
| internal doses resultfrom scenarios describing a residence being built i on a langfu, where excavation is usedfor a basement and the soil is 1 spread out over the yard and usedfor residential gardens. Doses can also result when a core is brought ypfrom weU-driUing on the site and j dispersed on the surface. For large scale landfil sites, animal !
| |
| pathways may be needed (i.e., radionuclide uptake by plants, plants l eaten by animals, animals used asfoodfor humans living nearby). l l
| |
| More recent computer codes can be used to perform the exposure and pathwy analyses. For instance, the GENII code and its many modules l have been used extensivelyfor dose calculations using many dgerent l types of scenarios associated with landfils and incinerators. External and intemal doses can be obtained; air, wter andplant and animal l foodpathmys are included. The GENil-S code incorporates a Latin l Hypercube uncertainty and sensitivity analysis capability into the GENIl code. The EPA's code, CAP 88-PC, can be usedfor modeling the ii transport andfate of radioactive air emissions and doses to the public j surrounding incinerators. This code also includes thefoodpathways in the dose calculations. Use of such codes shouldprovide more ;
| |
| " realistic" dose estimates.
| |
| l The authon of the 1987 ORAU report they realize that they were not experts in using the dispersion and groundwater models, and that they probably did not used the most up-to<iate models for the 1987 tirae frame. 'ney stated that now they realize the project would have benefitted from having specialists in these areas involved on the project.
| |
| S13. Page 7, 2nd 1: The incineration of medical wastes was used to select valuesfor the last two parameters described in the comment above. The incineration process usedfor medical wstes may be quite dgerent than that used for municipal wastes (i.e., which is a more likely occurrencefor GIDs). Dijferentfracticns of the radionuclides can be release through eact emissions under the dderent incineration conditions. Thus, more meirch into appropriate panitioning values is needed.
| |
| The IAEA (IAE4 Safety Series 111-P-1.1 [iAEA 1992]) has taken one B-6
| |
| | |
| , ~ _ . .
| |
| i 1-i.
| |
| h l approach to this problem oflack of knowledge about spectpc 1
| |
| radionuclide partitioning by conserustively assuming that 100% of the
| |
| ; initial radioactivity is retained in all three resultant phases. For l incinerator scenarios, these phases are slag, fly ash, andflue gases in i stack emissions. This triple accounting qproach will maximize the ,
| |
| f potential imponance of the scenarios associated with each possibility.
| |
| 1 A similar approach should be considered as a possible way to estimate the efect of radionuclide panitioning when no or limited panitioning i data is available.
| |
| * 1 1 A dsferent approach that may be used is a slight vanation of the IAEA method. An assunption is made that 100% of the radionuclides will end up in each of the three resulting incinerator phases. .But, instead of summing the dose resultsfrom all three resultant phases, only the \
| |
| maximum dosefor a given resultant phase is chosen and reported.
| |
| The ORAU authors stated that most of the literature available at the time the report ,
| |
| was prepared dealt with medical incinerators. Thus, the only parameter values they l felt comfortable in using were taken from those pertaining to the available literature.
| |
| 1 S14. Page 7, last 1: A metal recycling scenario was only consideredfor "Co. However, for a more realistic assessment, any of the materials couldpossiblypnd i their my to a metter. Thus, all nuclides involvedfor GLDs should be l considered in the dose assessments.
| |
| 1
| |
| 'Ibe ORAU authors agreed that the recycling section of the 1987 ORAU repon needs to be updated.
| |
| SIS. Page 8, 2nd 4: Inhalation of susperded panicles is not considered as an exposure pathwyfor the landfil burial scenario. As mentioned above in other comments, this scenarios should be includedfor both wrker scenarios, }
| |
| intruder scenarios, and well-drilling scenarios. For these scenarios, l inhalation may be the major exposure pathwy leading to internal doses l for some of the radionuclides considered.
| |
| The ORAU authors contend that they relied heavily on NUREG/CR-1775 as a general 3 basis for their repon. Since this panicular exposure scenario was not include in that l report, they did not feel in necessary to "melude it in the 1987 ORAU repon.
| |
| l i
| |
| S16. Page 8, last 1: Valuesfor teach ratesfrom landfills are give s. More up-to-date information on nuclide-specipe leach rates needs to be reviewed and B-7
| |
| | |
| l 2
| |
| i J
| |
| considerea.
| |
| ! See response to S12.
| |
| i S17. Page 9, 2nd 1: It is arbitrarily assumed that the dose to the average individual is one-1 half that of the MEl. The difTerence in dose to the average individal wrses the dose to the MEI should be dependent on the radionuclide, its '
| |
| half-hfe, the ytake pathmys considered, hold-up time through the l scenarios, time in a ginn location, amount of a ginnfood eaten per
| |
| ! year, etc., and generally is not a constant number as assumed. A more j in-depth dewtopment of scenarios and pathmys is needed in this study.
| |
| ) The ORAU authors agreed that the conversion to average individual dow from MEI dose should be a factor of 1/10 or less. 1 1-l l S18. Page 30, last 1: For internal dose calculations, afraction of 0.3 (30%) was l chosen as the maximum amount of the initial activity that can be
| |
| ; taken in. Since this value can vary with radionuclide, the value(s) need to be validated and changed as needed. \
| |
| i .
| |
| See response to St.
| |
| ' t i S19. Page 30, last 1: It is stated that hydrogen gas is usually converted to tritiated l mter in the atmosphere, and thus is not considered in the dose i calculations. Results of recent research indicate that this is not i the case; elemental tritium gas (HT) is not readily converted to
| |
| : tritiated uter in the atmosphere, but rather entymes present in
| |
| ! soil microbes are necessaryfor tius conwrsion to readily occur.
| |
| HT gas release experiments performed in Canada and France l
| |
| indicate that a sigmpcant ponion of the dose resultingfrom a l release of HT gas to the atmosphere is the secondary plume of j i tritiated uter releasedfrom the soit qtter the primary plume of \
| |
| l HT has contacted the soil. Also, a smallportion of the HT gas that is inhaled is absorbed into the blood, and is converted to tritiated water by gutflora. The signiffcant portion of the dose from inhalation of HT gas is associated with the tritiated water that isformedfrom the HT by this mechamsm and should be considered. (NOTE: The ICRP has not yet accepted these items in its dose commitment scheme.)
| |
| .$lternatively, a review of the paper " Maximum Permissible Amounts of Accidentally Released Tritiumfrom an Environmental Experiment to Meet Dose Limitsfor Public Exposure", by Taeschner, Bunnenberg, and Gulden, in Fusion Technology, August 1991, reporting on the 1986 B-8
| |
| | |
| ; 4 1
| |
| 1 French experimentsfound that the authors contend that the kinetics of 1 the reaction, l HT + H2O < - ----- > !I2 + TTO l favored theformation of tritiated wter (HTO) in all environments with suficient mass of water available, including thatfound in moist air.
| |
| Following an HT release under dry air cor.ditions, the absorption of HT in the soil, conwision to HTO, and re-release to the air will dominate the overall exposure to tritium (the dose due to exposure to the ir:itial HTplume will be sewral orders of magnitude lower). Under moist conditions, the owrall exposure will consist of roughly equalparts ;
| |
| exposure to the intial HTO plume (following rqpid conwrsion of HT to HTO in the atmosphere) and exposure to HTO absorbed in the soil and reemitted to the atmosphere over thefollowingfew days.
| |
| No response was made by the ORAU authors on this comment.
| |
| S20. Page 32, The " realistic individual" dosesfrom tritium are calculated 2nd 1: as,uming uptake of10%. This value needs to be researched and )
| |
| updated. This comments applies generally to all radionuclides l considered in this study. l No response was made by the ORAU authors on this comment. f.
| |
| S21. Page 114: No scenarios are consideredfor the class J devices with natural uranium or thorium. Howewr, the doses associated with these two radionuclides could be sigm)fcant and should be determined.
| |
| No response was made by the ORAU authors on this comment. ;
| |
| B.3. MAJOR CONCLUSIONS PERTAINING TO THE REVIEW OF THE 1987 ORAU REPORT The meeting with the authors of the 1987 ORAU report proved to be a significant help in identifying the constraints under with the study was done. 'Ibe report was designed to ,
| |
| address only the development of scenarios in which the public could receive doses from the )
| |
| improper transfer or disposal of generally licensed devices. It was their understanding that they were not charged with the task of establishing the total number of reported incidents ;
| |
| with these devices or with calculating doses specific to actual incidents involving these devices. As a result, we now understand that many of the parameter values in question ,
| |
| during this review were only used as benchmark values by the ORAU authors with the knowledge that more definitive values would be needed for the next phase of the analysis.
| |
| One item not addressed above is that while the committed dose equivalents to organs and B-9
| |
| | |
| l l
| |
| l 1
| |
| l l tissues (the units used in the ORAU report) is important, doses should be reported in unit: of '
| |
| , committed effective dose equivalent (CEDE) and total effective dose equivalent (TEDE = l aum of external and internal doses). This wcald have made cunparison of the resultant i doses between the different devices much easier.
| |
| Even with these caveats, it is still very difficult tojustify the use of 0.30 as the maximum fraction-taken-in value used in the 1987 ORAU report for calculation of internal doses.
| |
| First,'if only worst case scenarios were to be considered, the maximum fraction-taken-in should be 1.0, that is assume that all of the source is taken into the body. Second, as discussed in Appendix g, in the review of reports of over 60 actual accidents, in no case did an intake fraction ever reach the maximum 30% level used in the ORAU study. Actual observed values ranged from 2E-4 to 2E-8 for individuals involved in the accidents, and ranged from 7E-11 to 2E-15 for the Goiania cleanup workers (IAEA 1998b). If the doses for the various scenarios were recalculated using a defensible range of values for the fraction-taken-in, with all other parameter values the same, the CEDE and TEDE will be significantly lower than those that would result from using 0.30 value.
| |
| B. 4. GENERAL LISTING OF NEW DEVELOPMENTS IN INTERNAL DOSIMETRY SliiCE THE 1987 GLD REPORT There stillis no regulatory limit establishedfor collective dose to either workers or the public. k De revision to NRC 10CFR20 published in Federal Register and goes into efect January 1,1994. Dis updates the methodological basisfor the report dose limits from ICRP 2 (1959) to ICRP 26/30 (1977;1979). De dose limits in units of CEDE and TEDE.
| |
| Age dependence of doses is being looked at more closely Rejfnements to biokinetic models (eg.1992 Leggett Am raodel, etc.) have been made New tung models are being developed by ICRP and NCRP subcommittees ICRP60/61 methodologies (1991) changed recommended dose limits to 20 mSv per yearfor occupational exposures and 1 mSv in a yearfor exposures to the public. The ICRP 26/30 limits were 50 mSv per year for occupational exposures.
| |
| i i
| |
| i B-10 1
| |
| l
| |
| | |
| APPENDIX C: COMMENTS ON THE ORA U REPORT BY J. S. DUKELOW OrigNal Review Comments appean in the AppenA in italics. Nformation gained during the October 1993 visit to Oak Ridge to confer with the authors of the ORAU report appears interspersed in standard non-italic font. Additional relevant inforination gained since the October trip appears in a separate section at the end of this Appendix, also in non-italic Dnt.
| |
| C.J.
| |
| | |
| ==GENERAL COMMENT==
| |
| S Page i - The qppropriate "wrst case" assurrgptionfor exposure to the external radiation feldprobably ought to be direct contact of the encapsulated source with the bodyfor some number of hours. This case is plausible and is likely to produce more severe consequences that the report's assumption of 20 weks at 100 cm (panicularlyfor alpha and beta' sources).
| |
| The report assumes a "wrst case"for ingestion as ingestion or inhalation of 30% of the radioactive. Nojustipcation is given for using 30%; an obvious " worse" than worst case is inhalation or ingestion of100% of the radioactive material. Assuming 100% is not unreasonable of one remembert that some of the previous sealed source incidents involved sourcesfalling into the hands of small children.
| |
| We still consider these comments to be reasonable. For many of the sealed source accidents described in the body of the present report, the victim " finding" the scaled source has put it in his pocket. For the ingestion cases, several of the accidents have involved small children j, '
| |
| 2 finding and playing with sealed sources or sealed source material. A soft beta sealed source would require a delicate " shield window", which would be unlikely to survive stomach acid.
| |
| On the other hand, the analysis provided in the text of the present report establishes that these worst case assumptions are not at all representative of exposures resulting in the known sealed source accidents.
| |
| Page 2 - Some additional explanation is needed of the interior structure of Figure 1. For instance, one could argue that the " incinerator" block belongs on the downstream side of "salmge deckr" in addition to or instead of the upstream side. One of the things a salmge dealer might do with materials not deemed ofinterestfor metal recovery is incineration. On the other hand, the pathwy " trash handler to incinerator to salmge dealer" presumes that the trash handler sends to the salmged dealer incinerator ashes containing intact or dispersed sources, all of which seems implausible.
| |
| Page 3 - The " probabilities" described in Section 2.2 are properly called " conditional probabilities". Also, Section 2.2 ought to say that the probability of reaching apnal state by my of a particular pathway is the product of all the pathway segment conditional probabilities and that the probability of reaching thepnal state is the sum of all those pathway probabilities.
| |
| Duefact that the pathwy segment conditionalprobabilities are all (or mostly) " educated C-1
| |
| | |
| 1 l
| |
| l guess" deserves more emphasis, as well as some description of the basis used to make those
| |
| . educated guesses.
| |
| Page 5 - There is a discussion ofincreasing published dose rates by 25-45 per cent to accountfor " electron production in the stainless steel walls" of the capsule (presumably Congpton scattering, pairproduction, andphotoelectric absorption). Is this the entire rationale or should it include: "brenrstrahlung radsation resultingfrom the deceleration of beta particles, electrons, andpositrons in the stamiess steel"?
| |
| + ,
| |
| $ I 1he secondparagrqph refers to dose rates at a " depth of 7mg/cm**. Cember'sphrase I i
| |
| " density thickness" is more understandable; at any rate, some explanation should be provided 1 l for the reader who is not a health physicist.
| |
| 4 i
| |
| i The last two paragraphs of Section 2.3.1 disc:a;s internc! dose resultingfrom " intact" sources. How does this diferfrom the internal dose resultingfrom dispersed sources?
| |
| ?
| |
| Wilmot's 1981 report on spentfuel transponation accidents may not be a good guidefor l releasefractionsfor improper handling ofsealed sources because of the sigmpcant l i diferences in the barriers to rclease and in the purposeful nature of behavior that can be l i assumed in sealed source mishandling incidents.
| |
| Page 7 - No basis is givenfor the assumption that the population near the incinerator is l 1 73000. '
| |
| I
| |
| ; \
| |
| l Page 9 - The population of the US is 250M, so :he average population served by a landfil is 250000000/18500 = 13500. l 1
| |
| 1 i
| |
| , Page 13, Table 4 - The table would be clearer sf the last column were " Fraction"instead of ;
| |
| "Pers.entage". Thus, for Class A-1, the last column would show 5.0 X 105. i 4
| |
| Table 4 safersfrom a signipcant censoring problem. It does not repect those sources that have been mishandled in somefashion, butfor which the mishandling has not been detected.
| |
| l This is sigmfeant, because in known incidents, the detection was either accidental or i announced by the severe radiological consequences. Thus, the portion of the san;ple space l
| |
| ) that leads to low consequences will be sigmpcantly under-represented in Table 4, which is l
| |
| } nonetheless used to assign probabilities to event initiators.
| |
| l l
| |
| j The text of the present report discusses the c.uaoring issue in more detail and proposes a j methodology for estunating the anoint of cemoring in the available data, and thus, for j
| |
| arriving at estimates of the initiating event probabilities for various types of diversion or mishandling of Generally Licensed Sources.
| |
| C-2
| |
| | |
| i i
| |
| j 4
| |
| I
| |
| $~
| |
| C. 2.
| |
| f WHATIS NEEDED FOR REGUE.ATORY APPI2 CATION. BUT MISSING FROM THE ORAU REPORT?
| |
| lhe main thing missingfrom the ORAU report, but requiredfor regulatory application is a i set of realistic estimates of tise risk associated with the production and use of generally i
| |
| licensed sources. To obtain these realistic estimates of risk, we will need to make some l realistic estimates of thefrequencies associated with initiating ennts, the conditional l probabilities associated with the transitionsfrom state to state in the ORAU block diagrams,'
| |
| l and best estimates of the mishandling incident source terms. Wilmot (1981) does not strike
| |
| ! me as a gan basisfor those best estimate source terms, but I d:n't have an alternative to
| |
| ; ofer at the' moment.
| |
| C.3 ADDITIONAL INFORMATION GAINED SUBSEOUENT TO OCTOBER MEETING
| |
| } WITH ORAU AUTHORS l The present report now provides a detailed proposal for the estimation of initiating event i probabilities, taking the data censoring into account. These probabilities, together with best-
| |
| ! estimate calculations of the public health consequences of a diverted / mishandled source 4
| |
| ending up in a particular " final status", provide the basis for calculation of the public health i risk associated with Generally Licensed Sources of a particular type. ,
| |
| (
| |
| Although we had criticized the ORAU report for not including the initiating event ;
| |
| : i. probabilities and realistic estimates of the public health consequences of a
| |
| : diverted /midiandled source reaching a specific " final status", it har= clear durmg the October n-ing that the ORAU report authors were aware of this difficulty. They had been specifically constrained by the Scope of Work on their project not to estimate those quantities. Such constraints are not unreasonahk fn a first-cut analysis of whether any changes are needed in the regulation of generey bcensed devices; had all of the worst case consequences calculated by the ORAU authors been acceptably low, no further analysis would be needed to support a decision to maintain current regulation of generally licensed devices. With some of the worst case consegnences being naw=pahly high, it may be nereanary to collect the additional data needed c calculate the risk, i.e., the probability-weighted consequences, of generally licensed device diversion / mishandling.
| |
| l C-3
| |
| | |
| i 9 .
| |
| L
| |
| : APPENDEX D: COMMENTS ON 1987 ORAU REPORT BY G.R. CICOTTE i
| |
| D. J. REGblATORY CHANGES 1
| |
| Ihe USNRC made no sigmpcant changes in its regulations afecting the quantities or
| |
| { categories of generaRy licensed radioactive sources since 1987. There have also been other changes in the reponing, record keeping, and enforcement portions of the ppplicable j regulations. These are summarszed asfouows:
| |
| ; e The new 10 CFR 20 (56 FR 23360-23472, May 21,1991) resulted in changes to the i incident reporting requirements of10 CFR 30,10 CFR 31, and 10 CFR 32, invoking
| |
| : l. a choice between the old part 20 (f20.1-20.601) and the new part 20(f20.1001-0 20.2401),
| |
| e As part of the new part 20, general licensees were specupcally limited to doses of 10%
| |
| of the limits in either 10 CFR 20.101(a), i.e.,125 mrem (1.25 mSv) per calendar quaner, or 10 CFR 20.1201(a), i.e., 500 mrem (5 mSv) per calendar year. .
| |
| 1 e Record retention requirements were clanped (53 FR 19240-19246, May 27,1988) e' The new incident reporting requirements summarized in 56 FR 40757-40767 were ingposed on general licensees, by invokmg 10 CFR 30.50 in 10 CFR 31.2, and .
| |
| specsycany on general licenseesfor Am-241 calubration and reference sources, as specsped in 10 CFR 31.8.
| |
| 4 e Enforcement authority was placed on general licensees, specupcany authorizing iryunctions and orders deemed necessary by NRC, andproviding criminalpenaltiesfor wiR)ki violations. (57 FR 55063-55075)
| |
| D. 2. EFFECTS ON CONCLUSIONS IN ORAU REPORT OF CHANGED AND EXISTING REGULATIONS ;
| |
| The changes in the regt:lations should result in reporting of a greaterfraction of the incidents which occur. The ratios ofincident causes pppear to have changed as described below.
| |
| The regulations require that in order to use generally licensed devices in accordance with 10 CFR 31.5, the source must not be likely to result in a dose in etcess of 125 mrem (1.25 mSv) per quarter, which correlates to 500 mrem (5 mSv) per year. The Executive Summary of the ORAU Report states in part that: ". . . dose equivalents received by members of the general public who purchase contaminated products would most likely not exceed 500 mremlyr (0.005 Svlyr) in most cases. " Nofurther action would be necessary to protect the public sf the goal is to meet current regulatory limits as called out in 10 CFR 20.101(a) or 10 CFR 20.1201(a).
| |
| The change to the goal dose of 1 mSv per year to a member of the public may result in the need to regulate the manufacture oflicensed sources to assure that exceeding 1 mSv per year D-1
| |
| | |
| l t
| |
| i I
| |
| is unlikely.
| |
| D. 3. OTHER CONSIDERATIONS RELATED TJ THE ORAU REPORT \
| |
| The ORAU Report addressed thefollowing generally licensed uses:
| |
| : 1. Certain measuring, gauging, or controlling devices as authonzed by 10 CFR 31.5 -
| |
| Po-210, Am-241, Ra-226, H-3, Kr-85, Co-60, Cs-137, Sr-90, TI-204, Ru-106, Pm-147, C-14, Pb-210, Ni-63, Cm-244, Cd-109, and Fe-55.
| |
| : 2. Luminous safety devicesfor use in aircraft as authorized by 10 CFR 31.7 - H-3 and Pm-147.
| |
| l
| |
| : 3. Calibration or reference sources as ::utho:izcd by 10 CFR 31.8 - Am-241.
| |
| The ORAU Report did not appear to address thefollowing generally licensed uses:
| |
| : 1. Isotopes [ byproduct materialfor certain in vitro clinical or laboratory testing] as authorized by 10 CFR 31.11(a)(1) through (7).
| |
| ' 2. Ice detection devices containing Sr-90 as authorized by 10 CFR 31.10(a).
| |
| : 3. Beta- and/or gamma-enutting materials in measuring, gauging or controlling devices containing radioisotopes other than those specspcally listed in Table 1 of the ORAU ;
| |
| Report, authorized in 10 CFR 31.5.
| |
| : 4. Alpha-emitting materials in measuring, gauging or controlling devices ccantaining radioisotopes other than U-238, Am-241, Ra-226, or Pu-239, as authorized in 10 CFR 31.5.
| |
| It may be appropriate to consider whether scenariosfor iteru 1 and 2 above should be addressed in ca.y report developedfor the same purpose as the ORAU Report.
| |
| Generally licensed sources authorized in 10 CFR 31.5 are constrained to meet certain dose limits, even if damaged or misused. The general license in 10 WR 31.5 requires that the sources must have been manufactured or produced through adherence to a speapc license i usued pursuant to the condstions speaped by 10 CFR 32.11. 10 CFR 32.51 requires that individual sources must not cause doses to any person [not necessarily just a member of the !
| |
| public] in acess of the organ doses listed in 10 CFR 32.24. The organ dose limits have not changed since June 13,1%9, and are sursmarized bes'ow:
| |
| l l
| |
| D-2
| |
| | |
| - = - _ . - - - - .- . .. - - _ . _ _ - - - - - . - - - - _ . - - -
| |
| i 4
| |
| 1 i
| |
| e normal use of failure -
| |
| al.' units in failure - low negligible ,
| |
| : normal use of one location probability of probability of l
| |
| } Part ofBody one unit (rem) (rem) dose (rem) dose (rem)
| |
| { Whole body; head and
| |
| ; trunk; active blood-forming organs; gonads; or lens of eye. . . . . . . . . . . . . . . . . . 0.001 0.01 0.5 15 Hands andforearms; feet and ankles; localized are.s of skin awraged over areas no larger than i square centimeter. . . . . . . . . . . \
| |
| 0.015 0.15 7.5 200 Other organs.. . . . . . . . 0.003 0.03 1.5 50 D. 4. REFINEMENT OF SELECTED SCENARIOS IN THE ORAU REPORT i The ORAU Report hypothesizes a maximum skin contact. time of three hours (2.3.1). The limited time was based on the assunption that most of the devices cannot be corried in an individualperson's pocket due to size considerations. Ihe great majority of reponed incidents involving generauy licensed devices were related to portable static eliminators (task 3 repon), which are often small enough to be carried in a pocket, e.g., recording equipment static etuninator brushes.
| |
| Ihe dupersed source scenario assumes that d.'spersed material in a lan@ill would, if previously incinerated, leach into surface or ground unter apphes at e rate of 1.0 per year (2.3.2). Since most langflis now have teaching requirements whid rendt in much greater timefor breakthrough and there is some naturalJfitration by the lan&fil material itself, inclusion of this consideration could reasonably be expected to reduce the available dose l sigmffcantly.
| |
| The metals recycling section of the ORAU Repon assumes continuous contact our a yearfor a postulated maximum dose of 360 millirem (3.6 mSv) based on a dtstancefrom contact which reduces the dose to 0.1 of that calculatedfor continuous contact. In addition, Table 6 of the ORAU Repon states that whole body exposures to all categories are based on 20 weeks at 100 cm distance. A better approach might be to assume continuous contact over a reasonablefraction of the time, with the appropriate distancefactor. Two possibilities are householdproducts:
| |
| D-3
| |
| | |
| i b 1. Furniture - assume an occupancyfactor combined with a distancefac:or, and account
| |
| } Jbr some shielding by protective coas.ngs .iuch as pai::t (for beta emitters).
| |
| i i 2. Hand-held qppliances - assume conta tfor a postulated usage time, e.g., 30 minutes
| |
| ; per day in contact wi:h the OMUprojected dose rate of 410 pR/hfor Co-60.
| |
| i When asked about the 20 weeks at 100 cm criterion, the OMU teams indicated that it was j based on an analysis of the Juarez, Mexico accident, which indicated that apfth of the time
| |
| ! was . spent in that conpguration.
| |
| 1 The scenario development sectionfor static eliminators (3.1.2) asserts that static eliminators
| |
| , are unkkely to be recycled. In general, recycling has increased since 1987, and is expected l to continue to increase (although remaining at a relatively lowfraction of total recyclables),
| |
| { resulting in more of these devices entering the recycle scenario.
| |
| I' D.5. ADEQUACY OF THE ORAU STUDY Given the information on which the ORAU study was based, the conclusions regarding doses appeared adequate, with some exceptions. However, there are tv/c additional caveats to this statement. First, the study clearly states that a large number of assumptions.were conservatively made, due to lack of data regarding specific disposition. pathways. Second, l, the study clearly states that the " worst case" scenarios are excessively conservative, e.g., in section 3.7.3.1, the report states an assumption of ingestion / inhalation of 30% of the activity after damage to the source, even though the repo:t considered the possibility of access to the source to be " extremely unlikely."
| |
| D. 6. NEED FOR REGUIATORY ACTION In the opinion of this reviewer, there are two major considerations regardmg the needfor regulatory action. Deprst is related to the reliability of OMU study estimates. The second is related to the change in public dose. De change to occupational doses do not appear to reflect a needfor regulatory action.
| |
| : 1. Based on the number of assumptions which were required to be made, the estimates given are signipcantly in doubt, in particular as they relate to the probabilities of misuses resulting in skin contact or internal exposure. That is, the estimates cannot necessarily be given supicient credence tojustify regulatory action based only on the statement of a high estimate. In order to use the conservative values in the OMU study as the basisfor regulatory action, it would be necessary to conprm as realistic information the estimates given, or determine representative values with greater assurance as to accuracy.
| |
| : 2. The public dose limit is introduced in 10 CFR 20.1301, " Dose limitsfor individual D-4
| |
| ~ .,. __ . . - -
| |
| | |
| _ ___ - _~ . . - _ . . --
| |
| i i
| |
| l 4
| |
| 1 members of the public. " The new limit is 100 mrem (1 mSv) total efective dose
| |
| ; equinxient (TEDE). Thus, many of the conclusions in the ORAU study related to the .
| |
| : old limit would need to be reassessed relative to the new limit. Estimates meeting all f thefollowing (or similar) criteria muld need to be addressed through additional restrictions on generally licensed quantities or categories:
| |
| : a. The estimated TEDE exceeds the limit of1 mSv.
| |
| 4
| |
| } b. The probability of occurrence of the scenario is accurate within a margin l which indicates the limit would be exceeded with a apec@ed coq 6dence level, 5 e.g., when the estimate exceeds the limit by afactor of 2, the accuracy is within afactor of 2 to a specuped conpdence of 95%.
| |
| ; c. The dose estimatefor the pathay itselfis accurate within a margin which indicates the limit would be exceeded with a specip" conpdence level, e.g.,
| |
| !l when the estimate exceeds the limit by afactor of 2, the accuracy is within a j factor of 2 to a speciped conpdence of 95%.
| |
| i l d. The combined efect of b. and c. is accurate within a margin which indicates
| |
| { the limit would be exceeded with a speciped conpdence level, e.g., when the estimate exceeds the limit by afactor of 2, the accura:y of b. and c. combined is within afactor of 2 to a specspcd conpdence of 90%.
| |
| i
| |
| : e. Any arbitrary assumptions made in the estimate have been supplanted by l
| |
| i estimates based on actual information.
| |
| l D.7 ADDITIONAL INFORMATION OBTAINED IN MEETING WITH ORAU TEAM l AND SUBSEQUENTLY i
| |
| The ORAU team considered the pathways represented in the probability networks used in
| |
| ; their report to represent a condensation of hundreds or tlxmmt of possible pathways. Their i intent was to display only the pathways responsible for the "first" 95% or so of the risk.
| |
| I 2
| |
| When asked why Po-210 was considered a significant hamd, given its short half-life (138 days), the ORAU team responded that it was considered an m " gestion hazard due to the large number of Po-210 source and the loose conuol exercised over them.
| |
| The ORAU team indicated that sensitivity analysis was not part of the scope of their project.
| |
| When asked what they intended by the term " intact" sources, the ORAU team indicated that it varied somewhat, on a case-by-case basis, but that it was generally synonymous with
| |
| " localized". External dose was not calculated for dispersed sources, with the exception of recycled material, such as scrap. j D-5 I
| |
| | |
| , The ORAU team indicated that the " average" dose for intact sources was arbitrarily defined to be half of the worst-case dose.
| |
| The ORAU team was asked how population statistics were applied to external sources. They indicated that maximum and average population figures were applied only to dispersed sources (i.e., no external dose was assumed for populations, except for the case of recycled material).
| |
| l The ORAU team was asked to describe any special considerations applicable to their dose calcidarians: 1) absorption was not used,2) Maximum external dose was nasunal to be 0.3 rines the worst case for an intact source, 3) the most likely external dose was assumed to lx 104times the worst case for an intact source, 4) encapsulated source doses were calculated j based on NCRP 40, and 5) non-encapsulated source dose were calculated based on beta dose, using the methodology of Kocher and Eckerman The assumptions in Section 2.3.2 of the ORAU report are based on NUPJiG/CR-1775.
| |
| 'Ihe :,fmbol nC in Section 2.3.2.2 of the ORAU report refers to nanoCoulombs; the distance l l
| |
| 0.1 is in meters.
| |
| Re Section 2.3.2.3: Intmsion at landfills was not considered. The value e* refers to the 4 mean leach rate. ;
| |
| Re Section 2.3.1: The 50 year dose equivalent from ICRP 30 did not refer to the source remaining in place.
| |
| D-6
| |
| | |
| l l
| |
| l 1
| |
| I APPENDIX E TO FINAL REVIEW OF THE 1987 REPORT BY OAK RIDGE ASSOCIATED UNIVERSITIES,
| |
| " IMPROPER TRANSFER / DISPOSAL SCENARIOS FOR GENERAI,LY LIrWSED DF. VICES:"
| |
| REVISED TASK 3 TECHNICAL LETTER REPORT:
| |
| EVALUATION OF HISTORICAL SEALED SOURCE DEVICE EXPERIENCE NRC JOB CODE L2536 I PNL No. 20278 D. J. Strom G. R. Cicotte Health Protection Department Pacific Northwest Iaboratory Richland, Washington 99352 first draft August 27,1993 revised February 4,1994 revised May 31,1994 Prepared for the U.S. Nuclear Regulatory Commission Under Contract DE AC06-76RLO 1830 i
| |
| l
| |
| | |
| 1.0 EXECUTIVE
| |
| | |
| ==SUMMARY==
| |
| | |
| Data and reccrds provided to the Pacific Northwer Laboratory (PNL) by Steven L. Baggett and Sterling Bell of the U.S. Nuclear Regulatory Commission (NRC) have Seen reviewed to establish the historical experience of tested source device use and reported events of improper transfer or disposal.
| |
| : 2.0
| |
| | |
| ==SUMMARY==
| |
| OF FILES AND REPORTS USED IN THIS WORK i
| |
| Darnments supplied to PNL are listed in Section 6.0, References. The original work under
| |
| ; review in this project is referred to as the "ORAU Report" (Stabin et al.1987). '
| |
| Computer database files (in dBase III format for DOS) received and reviewed by PNL are listed in Attachment 1.
| |
| 3.0 NUMBER OF SEALED SOURCE LICENSES AND DEVICES OF EACH TYPE EXISTING IN EACH YEAR FOR WHICH RECORDS ARE READILY AVAILABLE l 4
| |
| 3.1 REVIEW OF SEALED SOURCE DEVICE : JSTRY The Scaled Source Device Registry (SSDR) provided basic information about the design, construction, uses and authorized maximum activity for each type of device. Devices are categonzed in the fashen of Table 1 of the ORAU Report, included here as Attachment 2.
| |
| Cor detailad, source-specific risk analyses, it would be zwe<=ary to determine the numbers of sources in each category by isotope, date placed in service, and activity. Use of design infonnaten regarding shielding would help to determine external doses in cases of improper transfer or disposal in which the source was not removed from the shield. Review in this level of detail was beyond the scope of the present project.
| |
| 3.2 REVIEW OF NUMBER OF GENERALLY-LICENSED DEVICES Four hardcopy reports from " General License Database System" were provided to the PNL reviewers attachad to a letter from Steven L. Baggett to Daniel J. Strom dated November 2,1993 (Baggett 1993). Two of the repons were entitled " General f renne Darsha=e System Report for Peer Review.' 'Ibe first of these, dated 10/20/93, was 6 pages long and cerniaad scaled source registrations in the years 1987-1992. 'Ibe second, dated 10/20/93, was 2 pages long and contained additional scaled source registrations in the ye a 1991-1992. Column handings on the two reports were year, isotope, device code, number of devices, number of general licensees, total activity, and average activity / device. (Device codes are shown in Table Task-3-2. Activities are in millicuries (mci).)
| |
| These two reports were keyed and analyzed as a review of historical data for a revision of the Task 3 report. 'Ibe ORAU report was limited to devices of less than 20 mci, but the " General License Database System Report for Peer Review" printouts contained many sources of activities significantly greater than 20 mci, so these data are difficult to use directly in a risk analysis.
| |
| E-1
| |
| | |
| These two printouts are summarized over the 6-year period -s Tables Task-3-1 A, Task-3-1B and Task-3-IC. In each table, the Device Codes are those identif ed in the printouts as being licensed !
| |
| under 10 CFP 31.5 and listed in Appendix C of Ragulatory Juide 10.10 (NRC 1987), with "W7" denoting self-luminous sources licensed under 10 CFR 31.7. He tables are identical except for ;
| |
| the sort order. Rows are labeled by nuclide and device code. Rows in Table Task-3-1 A are sorted by nuclide (alphabetically, the way the data were received) and within nuclide by Device Code; rows in Table Task-3-1B are sorted by Device Code and within Device Code by nuclide in order of increasing atomic number; and rows in Table Task-3-IC are sorted by fraction of total ingestion Aus contributed by a nuclide-Device Code combination.
| |
| In order to get an idea of the steady-state activity for sources, the activities of sources for years 1987-91 were decay-corrected to 1992. For some sources, e.g.,138-day 2''Po, this means that the 1987 sources had essentially disappeared. Each table shows the total number of devices registerA l between 1987 and 1992, the decay-corrected sum of the activities by nuclide and device code. I To rank the relative hazards of the nuclide-device code combinatuns, the ingestion Aus for each row were divided by the total number of ingestion Aus contained in sources registered during the 6-year span. Using the decay correction, there were 3.37 x 10' ingestion Aus in thest 325,681 I sources in 1992. The fractions are expressed in parts per million (ppm) to make the nu.nbers easier to compare.
| |
| 2 Table Task-3-IC shows that the ''Am D (gamma gauge) sources account for nearly half of the ingestion Aus (489,184 ppm or 49%), with 3H W self-lummous sources accounting for 26% of the Aus (using the admittedly incorrect, that is, too high by one or more orders of magnitude, AU for 'H 0,2 not the unlisted AU for 'H-labeled luminous materials). Promethium-147 E gauges 2
| |
| account for 9%, followed by *'Cm U devices (4.1%),232Cf E and U devices (2.9% and 2.6%), ;
| |
| and 28'Po static eliminators (2.5%). Tritium self-luminous light sources licensed under 10 CFR 31.7 account for about 1%, as do H 2 gas sources. l Also shown in the tables are crude risk assessmer,t r, umbers in the last 5 columns: the average ;
| |
| number of ingestion Aus per device; the average activity per device; the average source strength, PA, in rems per hour at I meter from an unshielded source with the average activity; the committed effective dose equivalent from ingesting 1/10,000 (104) of the Jource; and the dose equivalent one would receive by spending 1000 hours at I meter from an unshielded source. A justification for use of the factors,104 fraction-taken-in and 1000 hours used in the last two columns, is given in the main body of this report.
| |
| A number of difficulties were found in the SSDR printouts. These included invalid isotopes
| |
| ("K85," "KR-95," "KR84 4," "CE137," "CS137 SR"), invalid device codes (GAUGE, blank, V, and 31.7U), and an invalid number (0) of general licensees that occurs on 6 occasions. The obvious corrections were made for these cases (K85 changed to Kr-85; zeros changed to ones for general licensees). There were also some puzzling uses of isotopes which may be explainable as keying errors: 'H, "Fe, and "Sr gamma gauges; "h, "Co, '"Cs, 2''Po, and 2''Am beta gauges;
| |
| ''Ni and '"Cs neutron sources; Device Code I sources with activities less than 30 mci; and "Sr gas sources. There was no hasis for correcting these latter errors, so they remain as reported when included in the Tables Task-3-1 A and IB.
| |
| E-2
| |
| | |
| Table Task-3-1A. Summary of " General License Database System Report for Peer Review." Sorted by Atomic Number, Device Code.
| |
| '87 thru '92 sum of Average Dose Dos" Total Fractions of Source from Spc.iaing Number Sum of Totaling. Average Strength, Ingestion 1000 hours Dev- of Activity Alls Activity / Gamma *A of 10^-4 at 1 m from ice Devices, decayed to decayed to Average Ing. Device (rem /h of Source Source Nuclide Code Device Definition '87'92 1992 (mCl) 1992 (ppm) - Alls / Device (MCI) @im) (rems) (rems)
| |
| 'ii-3 D Gamma Gauges 1176 1.38E+07 1,364 3,909 11,727 -
| |
| 2.0 -
| |
| H-3 E Beta Gauges 2390 2.47E+0I 2,438 3,439 10,317 -
| |
| 1.7 -
| |
| H-3 N lon Gen, Chromatog 1378 5.62E+06 556 1,360 4,079 -
| |
| 0.68 -
| |
| H-3 0 ton Gen, Static Elim. 1392 1.51E+07 1,494 3,617 10,852 -
| |
| 1.8 -
| |
| H-3 R Gas Sources 10771 1.02E+08 10,066 3,150 9.451 -
| |
| 1.6 -
| |
| H-3 S Foll Sources 25 1.25E+03 0.12 17 50 -
| |
| 0.0083 -
| |
| H-3 T Other 20 2.93E+ 02 0.029 - 5 15 -
| |
| 0.0024 -
| |
| H-3 W Self-Lum Light Src 242505 2.66E+09 262,841 3,654 10,961 -
| |
| 1.8 -
| |
| H-3 W7 31.7 Self-Lum t,lght Src 18519 1.07E+08 10,596 1,929 5,786 -
| |
| 1.0 -
| |
| C-14 E Beta Gau0es 6 6.00E-01 0.0020 1 0.10 -
| |
| 5.55E-04 -
| |
| C-14 N lon Gen, Chromatog 10 1.00E-02 0.000033 0 0.0010 -
| |
| 5.55E-06 -
| |
| C-14 R Gas Sources 42 2.10E+0n 0.0069 1 0.050 -
| |
| 2.78E-04 -
| |
| C-14 T Other 33 7.25E+00 0.024 2 0.22 -
| |
| 0.0012 -
| |
| C-14 W Self Lum Ll0ht Src 267 3.67E+01 0.12 2 0.14 -
| |
| 7.64 E-04 -
| |
| C-14 Y Calibrators 24 1.20E-01 0.00040 0 0.0050 -
| |
| 2.78E-05 -
| |
| Sc-46 N lon Gen Chromatog 167 4.91E+02 0.16 3 2.9 0.0034 0.0016 3.4 Tl-44 N lon Gen, Chromatog 13 6.43E+02 _ 0.64 165 49 0.0072 0.082 7.2 Fe-55 D Gamma Gauges 57 1.10E+03 0,036 2 19 -
| |
| 0.0011 -
| |
| Fe-55 E Beta Gauges 13 4.64E+02 0.015 4 36 -
| |
| 0.0020 -
| |
| Fe-55 N lon Gen. Chromatog 18 7.67E+02 0.025 5 43 -
| |
| 0.0024 -
| |
| Fe-55 T Other 12- 5.60E+02 0.018 5 47 -
| |
| 0.0026 -
| |
| Fe-55 U X-Ray Fluorescence 811 6.51E+04 2 9 80 -
| |
| 0.0045 -
| |
| Co-60 D Gamma Gauges - 41 4.13E+03 2 201 101 0.14 - 0.10 - 135 Co-60 E Beta Gauges 3 3.52E+01 0.021 23- 12 0.016 0.012 16 i'
| |
| Ni-63 E Beta Gauges 22 2.17E-01 7.15E-06 0 0.010 -
| |
| 5.48E-07 -
| |
| N!-63 H Gen Neut Src Apps 22 3.30E+02 0.011 2 15 -
| |
| 8.33E-04 -
| |
| NI-63 N lon Gen, Chromatog 3451 2.59E+05 9 8 75 -
| |
| 0.0042 -
| |
| Ni-83 0 ton Gen Static EHm. 56 8.28E+02 0.027 2 15 -
| |
| 8 21E-04 -
| |
| E-3
| |
| | |
| Table Task-3-1 A continued. Sorted by Atomic Number and Device Code. i
| |
| '87 thru '92 sum of Average Total Dose Dose from Fractions of Source Number Sum of from Spending Totaling. Average Strength, Ingestion 1000 l'ours Dev- of Activity Alls Activity / Gamma *A of 10^-4 at 1 m from ice l Devices, decayed to decayed to Avera0e Ing. Device (rem /h of Source Nuclide Code Device Defintion Source .i
| |
| '87 *92 1992 (mCO 1992 (ppm) ALis/ Device Ni-63 S Foil Sources 6 (mCO Gim) (rems) (rems) 6.99E+0,1 0.0023 1 12 Ni-63 T Other
| |
| -- 6.47E-04 -
| |
| 12 1.19E-01. 3.93E-06 0 i
| |
| 0.010 -
| |
| 5.52E-07 - -'
| |
| Ni-63 U X-Ray Fluorescence 8 3.91E+ 01 0.0013 1 4.9 -
| |
| 2.725-04 N163 W Self-Lum Ll9ht Src 19 1.20E+05 4 702 6,320 Kr-85 D Gamma Gauges 0.35 - !
| |
| 291 1.20E+05 0 0 Beta Gauges 413 0.00E+00 -
| |
| 0.00E+00 Kr-85 E 623 2.43E+06 0 0 3,905 0.0061 -
| |
| 6.1 Kr-85 N lon Gen, Chromatog 135 6.94E+04 0 0 514 8.05E-04 -
| |
| 0.80 Kr-85 0 lon Gen Static Elim. 128 3.44E+02 i 0 0 '2.7 4.21E-06 '
| |
| Kr-85 R C as Sources 2 2.11E+03 0.0042 0 0 1,054 0.0017 Kr-85 T Other 1.7 :
| |
| 11 2.20E+03 0 0 200 3.13E-04 Kr-85 U X-Ray Fluorescence 13 0.31 I 5.44E+01 0 ,
| |
| 4.2 6.55E-06 Sr-90 D Gamma Gauges 104 0.0065 i' 2.28E+04 226 7.2.4 220 -
| |
| 3.7 St-90 F Bela Gauges 1153 3.94E+05 3,895 11,387 ,
| |
| 342 -
| |
| 5.7 Sr-90 1 Calib Src A>30 MCI 1 5.00E-01 0.0049 17 0.50 -
| |
| 0.0083
| |
| 'l Sr-90 R Gas Sources 135 2.44E+02 2 60 1.8 -
| |
| 0.030 -
| |
| St-90 T Other 182 8.63E+01 0.85 16 0.47 1
| |
| 0.0079 -
| |
| Sr-90 Y Calibrators 2 2.41E-03 0.000024 0 0.0012 2.01E-05 ;
| |
| Ru-106 D Gamma Gauges 1 4.00E+01 0.059 100 20 -
| |
| 0.050 -
| |
| i Ru-106 E Beta Gauges 1 9.65E-04 1.43E-06 0 0.0010 -
| |
| 2.41E-06 -
| |
| Cd.109 D C amma Gauges 44 5.27E+01 0.052 4 1.2 2.21E-04 0.0020 0.22 ;
| |
| Cd-109 E Beta Gauges .85 4.60E+32 0.45 18 5.4 0.0010 0.0090 1.0 !
| |
| Cd-109 N lon Gen, Chromatog 8 2.65E+01 0.026 11 3.3 6.11E 04 0.0055 0.61 !
| |
| Cd-109 T Other 10 4.90E+01 0.046 16 4.9 9.03E-04 0.0082 0.90 [
| |
| Cd-109 U X-Ray Fluorescence 567 6.25E+02 0.62 4 1.1 2.03E-04 0.0018 0.20 i 1-129 T Other 24 2.96E+01 0.0029 0 1.2 1.55E-04 2.05E-04 0.16 [
| |
| Cs-137 D Gamma Gauges 2493 2.20E+06 6,533 8,833 883 0.34 4.4 337 Cs 137 E Beta Gauges !
| |
| 31 6.05E+05 1,795 195,228 19,523 7.5 98 7.455 {
| |
| Cs-137 H Gen Neut Src Apps 45 3.56E+03 11 791 79 0.030 0.4 ") 30 ;
| |
| E-4 i
| |
| i t
| |
| | |
| Table Task-3-1 A continued. Sorted by Atomic Number and Device Code. '
| |
| '87 thru '92 sum of Average Total Dose Dose from Fractions of Source Number from Spor' ding Sum of Totaling. Average Strength, InDestion 1000 hours Dev- of Activity Alls Activity / Gamme*A of 10^-4 at 1 m from Ice Devices, decayed to decayed to Average Ing. Device (remIh of Source Source Nuclide Code Device Definition '87*92 1992 (mCQ 1992 (ppm) ALis/ Device Cs-137 i Callb Src A>30 mCl 17 (mCO Gim) (rems) (rems) 6.49E+02 1.9 382 38 0.015 0.19 15 Cs-137 N lon Gen, Chromatog 11 3.13E+02 0.93 284 28 0.011 0.14 11 Cs 137 T Other 1060 1.40E+06 4,139 13,163 1,316 0.50 6.6 503 Cs 137 U X Ray Fluorescence 18 2.60E+03 8 1,445 144 0.055 0.72 55 Cs-137 W Self-Lum L10ht Src 5 2.25E+03 7 4,503 450 0.17 2.3 172 Cs-137 Y Calibrators 109 4.64E+01 0.14 4 0.43 1.62E-04 0.0021 0.16 Ba-133 i Calib Src A>30 MCI 93 1.53E+00 0.00023 0 0.016 7.51E-06 4.12E-06 0.0075-Ba-133 N lon Gen, Chromatog 5 8.79E-02 0.000013 0 0.018 8.01E-06 4.39E-06 0.0080 Ba-133 T Other 644 5.88E+02 0.087 0 0.91 4.16E-04 2.28E-04 0.42 Pm-147 D Gamma Gauges 11 6.20E+03 0.46 141 564 1.51E-06 0.070 0.or!15 Pm-147 E Beta Gauges 2685 1.09E+C9 80,974 101,660 406,639 0.0011 51 1.1 Pm-147 T Other 1 5.90E+02 0.044 147 590 1.58E-06 0.074 0.0016 Pm-147 W Self-Lum Light Src 1 4.42E+04 3 11.055 44,220 1.18E-04 5.5 0.12 Eu-152 T Other 56 1.08E+00 0.00040 0 0.019 1.44E-05 1.21E-05 0.014 TI.204 E Beta Gauges 538 4.05E+0e 601 3,764 7,528 0.0084 1.9 Beta Gauges 8.4 Po-210 E 98 3.85E+01 4 131 0.39 2.07E-09 0.085 lon Gen, Static Ellm. 2.07 E-06 Po-210 O 27235 2.55E+05 25,251 3,125 9.4 4.94E-08 1.6 - 4.9 tE-05 Po-210 U X-Ray Fluorescence 535 9.23E+03 912 5,749 17 9.09E-08 2.9 9 Ot E-05 Bi-210 E Beta Gauges 1 2.79E-89 1.04E-92 0 2.79E-89 -
| |
| 1.75E-92 -
| |
| Ra-226 i Callb Src A>30 MCI 1 2.00E-02 0.0030 10 0.020 2.85E-05 0.0050 0.029 Ra-226 S FoilSources 53 1.24E+00 . 0.18 12 0.023 3.34E-05 0.0058 0.033 Ra-226 T Other 12 1.20E-01 0.018 5 0.010 1.42E-05 0.0025 0.014 Pu-238 D Gamma Gauges 2 5.84E+01 19 32,425 29 0.0023 16 2.3 Pu 238 U X Ray Fluorescence 12 3.52E+02 116 32,620 29 0.0023 16 2.3 Am-241 D Gamma Gauges 2261 1.32E+06 489,184 729,319 ~ 583 0.18 365 183 Am-241 E Beta Gauges 82 7.89E+04 29,240 1,202,034 962 0.30 601 302 Am-241 H Gen Neut Src Apps 3 1.36E+03 504 565,999 453 0.14 283 142 Am-241 i Calib Src A>30 mCl 9 2.17E-01 - 0.080 30 0.024 7.55E-06 0.015 0.0076 E-5 m___ -
| |
| _ _ _ _ _ _ . _ - . _ . ___---____m. _ _ _ - - _ . _.___-_..-_____a _ _ _ _ . _ _ - - _ . - _ . ---_A__________.o___.i - - - - --_m>_---a__
| |
| | |
| ~_ ._.__...m . . _ . _ _ . _ . _____._____._..-_....._.._.m_ __._ ._ _ _ . . _ _ _
| |
| I J
| |
| . Table Task-3-1A continued. Sor1ed by Atonde Num%r and Device Code.
| |
| '87 thru '92 -i sum of Aversee Dose Dose from Total Fractions of Source from Spending Number , Sum of TotalIn0 Average Strength, Ingestion 1000 hours 4
| |
| Dev. of Activity Alls Activity / Gamme*A of10^-4 et 1 m from i ice Devices, decayed to decayed to Average Ing. Device (remm of Source Source '
| |
| Nuclide -Code Device DeAnition 87-12 1992(mCO 1992(ppm) Alls / Device (mCl) Sim) (rems) lon Gen, Chromatog (rems) t Am-241 N 1 1.30E-01 0.048 162 0.13 4.07E-05 0.081 0.041 Am-241 O lon Gen, Static Eum. 107 1.47E+03. 545 17,184 14 0.0043 8.6 4.3 ;
| |
| Am-241 S Foll Sources 2 2.60E-01 0.10 162 0.13 4.07E-05 0.081 0.041 e Am-241 T Other 7 2.00E+01 7 3,572 2.9 8.96E-04 1.8 0.90 Am-241 U- X-Ray Fluorescerne 350 6.97E+04 25,830. 248,772 199 0.062 124 6;.
| |
| Am-241 Y Cahbretors 8 6.79E-05 0.000025 0 1.13E-05 ~ 3.55E-09 7.06E-06 3.55E-06 Cm 244 D Gamme Gauges 16 2.16E+03 .642 135,307 '135 0.0087- 68- I 8.7 Cm-244 U X-Ray Fluorescence 239 1.35E+05 40,171 566.579 567 0.036 283 36 Cf-252 H Gen Neut Src Apps 16 1.60E+00 0.24 50 0.10 4.18E-06 0.025 0.004,.
| |
| 325681 I
| |
| r I
| |
| I i
| |
| t
| |
| [
| |
| I E-6 )
| |
| i a,
| |
| j r
| |
| _- . _ _ _ - . _ - . _ - _ - - _ - _ - _ _ - _ _ - _ _ . . _ _ . _ - _ - - - _ _ _ _ _ _ _ _ _ . _ _ - _ _ _ _ . - _ _ _ _ _ _ _ _ _ - - _ _ _ . . - _ - _ - _ - _ _ _ _ - _ _ - _ - _ _ _ _ - _ _ _ -__._____--____-_Y
| |
| | |
| Table Task-3-1B. Summary of " General License Database System Report for Peer Review." Sorted by Device Code and Nuclide.
| |
| '87 thru '92
| |
| :;um of Average Dose Dose from Total Fractions of Source from Epending Number Sum of Total ing. Average Strength, ingestion 1000 hours Dev- of Activity Alls Activity / Gamma *A of 10^-4 at 1 m from ice Devices, decayed to decayed to Average ing. Device (rem /h of Source Source Nuclide Code Device Definition '87'92 1992 (mci) 1992 (ppm) Alls / Device (mCl) @1m) (rems) (rems)
| |
| H-3 D Gamma Gauges 1176 1.38E+07 1,364 3,909 11,727 -
| |
| 2.0 -
| |
| Fe-55 D Gamma Gauges 57 1.10E+03 0.036 2 19 -
| |
| 0.0011 -
| |
| Co-60 D Gamma Gauges 41 4.13E+03 2 201 101 0.14 0.10 138 Kr-85 D Gamma Gauges 291 1.20E+05 0 0 413 1.53E-04 -
| |
| 1.53E-01 Sr-90 D Gamma Gauges 104 2.28E+04 226 7,324 220 -
| |
| 3.7 -
| |
| Ru-100 D Gamma Gauges 2 4.00E+01 0.059 100 20 -
| |
| 0.050 -
| |
| Cd-109 D Gamma Gauges 44 5.27E+01 0.052 4 1.2 2.21 E-04 0.0020 0.22 Cs-137 D Gamma Gauges 2493 2.20E+06 6,533 8,833 883 0.34 4.4 337 Pm-147 O Gamma Gauges 11 6.20E+03 0.46 141 564 1.51E-06 0.070 0.0015 Pu-238 O Gamma Gauges 2 5.84E+01 19 32,425 29 0.0023 16 2.3 Am-241 D Gamma Gauges 2261 1.32E+06 489,184 729,319 583 0.18 365 183 Cm-244 D Gamma Gauges 16 2.16E+03 642 135,307 135 0.0087 68 8.7 H-3 E Beta Gauges 2390 2.47E+07 2,438 3,439 10,317 -
| |
| 1.7 -
| |
| C-14 E Beta Gauges 6 6.00E-01 0.0020 1 0.10 -
| |
| 5.55E-04 -
| |
| Fe-55 f:: Beta Gauges 13 4.64E+02 0.015 4 38 -
| |
| 0.0020 -
| |
| Co-60 E Beta Gauges 3 3.52E+01 0.021 23 12 0.016 0.012 16 Ni-63 E Beta Gauges 22 2.17E-01 7.15E-06 0 0.010 -
| |
| 5.48E-07 -
| |
| Kr-85 E Beta Gauges 623 2.43E+08 0 0 3,905 0.0061 -
| |
| 6.1 Sr-90 E Beta Gauges 1153 3.94E+05 3,895 11,387 342 -
| |
| 5.7 -
| |
| Ru-106 E Beta Gauges 1 9.65E-04 1.43E-C6 0 0.0010 -
| |
| 2.41E-06 -
| |
| Cd-109 E Beta Gauges 85 4.60E+02 0.45 18 5.4 0.0010 0.0090 1.0 Cs-137 E Beta Gauges 31 6.05E+05 1,795 195,228 19,523 7.5 98 7,455 Pm-147 E Beta Gauges 2685 1.09E+09 80,974 101,660 406,639 0.0011 51 1.1 TI-204 E Beta Gauges 538 4.05E+06 601 3,764 7,528 0.0084 1.9 8.4 Po-210 E Beta Gauges 98 3.85E+01 4 131 0.39 2.07E-09 0.065 2.07E-06 Bi-210 E Beta Gauges 1 2.79E-89 1.04E-92 0 2.79E-89 -
| |
| 1.75E-92 -
| |
| Am-241 E Beta Gauges 82 7.89E+04 29,240 1,202,034 962 0.30 601 302 Ni-63 H Gen Neut Src Apps 22 3.30E+02 0.011 2 15 3
| |
| 8.33E-04 -
| |
| E-7
| |
| | |
| Table Task-3-1B continued. Sorted by Device Code and Nuclide.
| |
| '87 thru '92 sum of Average Dose Oc e from Total Fractions of Source from Spending Number Sum of Total Ing. Average Strength, Ingestion 1000 hours Dey- of Activity ALis Activity / Gamma *A of 10^-4 at 1 m from ice Devices, decayed to decayed to Average Ing. Device (rem /h of Source Source Nuclide Code Device Definition '87'92 1992 (mCl) 1992 (ppm) Alls / Device (MCI) @1m) (rems) (rems)
| |
| Cs-137 H Gen Neut Src Apps 45 3.56E+03 11 791 79 0.030 0.40 30 Am-241 H Gen Neut Src Apps 3 1.36E+03 504 565,999 453 0.14 283 142 Cf-252 H Gen Neut Src Apps 16 1.60E+00 0.24 50 0.10 4.18E-06 0.025 0.0042 Sr-90 1 Calib Src A>30 mCl 1 5.00E 01 0.0049 17 0.50 -
| |
| 0.0083 -
| |
| Cs 137 / Calib Src A>30 mCl 17 6.49E+02 1.9 382 38 0.015 0.19 15 Ba-133 i Calib Src A>30 mci 93 1.53E+00 0.00023 0 0.016 7.51E-06 4.12E-06 0.0075 Rt-226 i Calib Src A>30 mC; 1 2.00E-02 0.0030 10 0.020 2.85E-05 0.0050 0.029 Am-241 1 Calib Src A>30 mCl 9 2.17E-01 0.080 30 0.024 7.55E-06 0.015 0 0076 H-3 N lon Gen, Chromatog 1378 5.62E+06 556 1,360 4,079 -
| |
| 0.68 -
| |
| C-14 N lon Ger, Chromatog 10 1.00E-02 0.000033 0 0.0010 -
| |
| 5.55E-06 -
| |
| Sc-46 N lon Gen Chromatog 167 4.91 E+02 0.16 3 2.9 0.0034 0.0016 3.4 Ti-44 N lon Gen, Chromatog 13 6.43E+02 0.64 165 49 0.0072 0.082 7.2 Fe 55 N lon Gen, Chromatog 18 7.67E+02 0.025 5 43 -
| |
| 0.0024 -
| |
| Ni-63 N lon Gen, Chromatog 3451 2.59E+C5 9 8 75 -
| |
| 0.0042 -
| |
| Kr-85 N lon Gen, Chromatog 135 6.94E+04 0 0 514 8.05E-04 -
| |
| 0.80 Cd-109 N lon Gen, Chromatog 8 2.65E+01 0.026 11 3.3 6.11E-04 0.0055 0.61 Cs-137 N lon Gen, Chromatog 11 3.13E 602 0.93 284 28 0.011 0.14 11 Ba-133 N lon Gen, Chromatog 5 8.79E-02 0.000013 0 0.018 8.01E 06 4.39E-06 0.0080 Am-241 N. lon Gen, Chromatog 1 1.30E-01 0.048 162 0.13 4.07E-05 0.081 0.041 H-3 O lon Gen, Static Elim. 1392 1.51 E+07 1,494 3,617 10,852 -
| |
| 1.8 -
| |
| Ni-63 O lon Gen, Static Elim. 56 8.28E+02 0.027 2 15 -
| |
| 8.21E-04 -
| |
| Kr 85 O lon Gen, Static Elim. 128 3.44E+02 0 0 2.7 4.21E-06 -
| |
| 0.0042 Po-210 0 lon Gen, Static Elim.. 27235 2.55E+05 25,251 3,125 9.4 4.94E-08 1.6 4.94E-05 Am-241 O lon Gen, Static Elim. 107 1.47E+03 545 17,184 14 0.0043 8.6 4.3 H-3 R Gas Sources 10771 1.02E+08 10,066 3,150 9,451 -
| |
| 1.6 -
| |
| C-14 R Gas Sources 42 2.10E+00 0.0069 1 0.050 -
| |
| 2.78E-04 -
| |
| Kr-85 R Gas Sources 2 2.11E+03 0 0 1,054 0.0017 -
| |
| 1.7 Sr-90 R Gas Sources 135 2.44E+02 2 60 1.8 -
| |
| 0.030 -
| |
| E-8
| |
| . - _ _ _ _ _ - _ . . _ _ _ _ _---_-_-____7_ -. _ _ _ _-__ __-_ ______ _-__ ---- .-_____
| |
| | |
| Table Task-3-1B continued. Sorted by Device Code and Nuclide.
| |
| '87 thru '92 sum of Average Dose Do.e from Total Fractions of Source from Spending Number Sum of Total Ing. Average Strength, Ingestion 1000 hours Dev- of Activity ALli Activity / Gamma *A of 10^-4 at 1 m from ice Devices, decayed to decayed to Average Ing. Device (rem /h of Source Source Nuclide Code Device Definition '87'92 1992 (MCI) 1992 (ppm) Alls / Device (MCI) @1m) (rems) (rems)
| |
| H-3 S Foil Sources 25 1.25E+03 0.12 17 50 -
| |
| 0.0083 -
| |
| NI-63 S Foil Sources 6 6.99E+ 01. 0.0023 12 1 -
| |
| 6.47E-04 -
| |
| Ra-226 S Foll Sources 53 1.24E+00 0.18 12 0.023 3.34E-05 0.0058 0.033 Am-241 S Foil Sources 2 2.60E-01 0.10 162 0.13 4.07E-05 0.081 0.041 H-3 7 Other 20 2.93E+02 0.029 5 15 -
| |
| 0.0024 -
| |
| C-14 7 Other 33 7.25E+00 0.024 2 0.22 -
| |
| 0.0012 -
| |
| Fe-55 T Other 12 5.60E+02 0.018 5 47 -
| |
| 0.0026 -
| |
| Ni-63 7 Other 12 1.19E-01 3.93E-06 0 0.010 -
| |
| 5.52E-07 -
| |
| Kr-85 T Other 11 2.20E+03 0 0 200 3.13E-04 -
| |
| 0.31 Sr-90 7 Other 182 8.63E+01 0.85 16 0.4 / -
| |
| 0.0079 -
| |
| Cd-109 7 Other 10 4.90E+01 0.048 16 4.9 9.03E-04 0.0082 0.90 1-129 7 Other 24 2.96E+01 0.0029 0 1.2 1.55E-04 2.05E-04 0.16 Cs-137 7 Other 1060 1.40E+06 4.139 13.163 1,316 0.50 6.6 503 Ba-133 7 Othe. 644 5.88E+02 0 '187 0 0.91 4.16E-04 2.28E-04 0.42 Pm-147 7 Other 1 5.90E+02 0. 14 4 147 590 1.58E-06 0.074 0.0016 Eu-152 7 Other 56 1.08E+00 0.00040 0 0.019 1.44E-05 1.21 E-05 0.014 Ra-226 T Other 12 1.20E-01 0.018 5 0.010 1.42E-05 0.0025 0.014 Am-241 7 Other 7 2.00E+01 7 3,572 2.9 8.98E-04 1.8 0.90 Fe-55 U X-Ray Fluorescence 811 6.51 E+04 2 9 80 -
| |
| 0.0045 -
| |
| Ni-63 U X-Ray Fluorescence 8 3.91E+01 0.0013 4.9 1 -
| |
| 2.72E-04 -
| |
| Kr-85 U X-Ray Fluorescence 13 5.44E+01 0 0 4.2 6.55E-06 -
| |
| 0.0065 Cd-109 U X-Ray Fluorescence 567 6.2SE+02 0.62 4 1.1 2.03E-04 0.0018 0.20 Cs-137 U X-Ray Fluorescence 18 2.60E+03 8 1,445 144 0.055 0.72 55 Po-210 U X-Ray Fluorescence 535 9.23E+03 912 5,749 17 9.09E-08 2.9 9.09E-05 Pu-238 U X-Ray Fluorescence 12 3.52E+02 116 32,620 29 0.0023 16 2.3 Am 241 U X-Ray Fluorescence 350 6.97E+04 25,830 248,772 199 0.062 124 62 Cm-244 U X-Ray Fluorescence 239 1.35E+05 40,171 566,579 567 0.036 283 36 H-3 w self-Lum Light Src 242505 2.66E+09 262,841 3,654 10,961 -
| |
| 1.8 -
| |
| E-9
| |
| | |
| Table Task-3-1B continued. Sorted by Device Code and Nuclide.
| |
| '87 thru '92 sum of Average Dose Oose fro n Total Fractions of Source from Spending ,
| |
| Number Sum of Total Ing. Average Strength, Ingestion 1000 hours '
| |
| Dey- of Activity ALis Act'vityl Gamma *A of10^-4 at 1 m from ice Devices, decayed to decayed to Average Ing. Device (rem /h 3 Source Source Nuclide Code Device Definition '87 '92 1992 (mCl) 1992 (ppm) Alls / Device (mCl) Gim) (rems) (rems)
| |
| C-14 W Self-Lum Light Src 267 3.67E+01 0.12 2 0.14 - 7.64E-04 -
| |
| Ni-63 w Self-Lum Light Src 19 1.20E+05 4 702 6,320 -
| |
| 0.35 -
| |
| Cs-137 W Self-Lum LIGht Src 5 2.25E+03 7 4,503 450 0.17 23 172 Se!f-Lum Light Src 1 4.42E+04 3 11,055 44,220 1.18E-04 5.5 0.12 Pm-147 w W7 31.7 Self-Lum Light Src 18519 1.07E+08 10,596 1,929 5,786 -
| |
| 1.0 -
| |
| H-3 Y Calibrators 24 1.20E-01 0.00040 0 0.0050 -
| |
| 2.78E-05 -
| |
| C-14 Sr-90 Y Calibrators 2 2.41E-03 0.000024 0 0.0012 -
| |
| 2.01E-05 -
| |
| Os-137 Y Calibrators 109 4.64E+01 0.14 4 0.43 1.62E-04 0.0021 0.16 Am-241 Y Calibrators 6 6.79E-05 0.000025 0 1.13E-05 3.55E-09 7.08E-06 3.55E-06 325681 I
| |
| E-10
| |
| _ _ _ . . _ _ . _ _ . _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _- s-- _ _ _ _
| |
| | |
| _ _ _ . - - ____.._,..__._ _ _ .~_.._._ _ _ - _ . ._._.._. ____.__.--_-_ __ _ _.._____. _._ _._
| |
| Table Task-3-1C. Summary of " General Licose Database System Report for Peer Review." Sorted by " Fractions of Total Ing. ALIs."
| |
| . '87 thru 92 sum of Average Dose Dose from Total Fractions of Source , from Spending 3
| |
| Number Sum of Totafing. Average Strength, Ingestion 1000 hours Dev- of Activity Atis Activity / Gamma *A of 10^-4 at 1 m from Ice Devices, decayed to deceyedto Average Ing. Device (remm of Source Source Nuclide Code Device Definition '87'92 1992(mCl) f992(ppm) Alls / Device (mCl) Gim) (rems) (rems)
| |
| Am-241 D Gamma Gauges 2261 1.32E+06 489,184 729,319 583 0.18 365 183 H-3 W Self-Lum LIOht Src 242505 2.66E+09 262,841 3,654 10,961 -
| |
| 1.8 -
| |
| Pm-147 E Beta Gauges 2685 1.09E+09 80,974 101,660 406,639 0.0011 51 - 1.1 ,
| |
| Cm-244 U X-Ray Fluorescence 239 1.35E+05 40,171 566,579 567 0.036 283 36 Am-241 E Beta Gauges 82 7.89E+04 29,240 1,202,034 962 0.30 601 302 Am-241 U X-Ray Fluorescence 350 6.97E+04 25,830 248,772 199 0.062 124 62 !
| |
| Po-210 O lon Gen, Static Ellm. 27235 2.55E+05 25,251 3,125 9.4 4.94E-08 1.6 4.94E-05 H-3 W7 31.7 Self-Lum Light Src 18519 1.07E+08 10,596 1,929 5,786 -
| |
| 1.0 - -
| |
| H-3 R Gas Sources 10771 1.02E+08 10,066 3,150 9,451 -
| |
| 1.6 . ;
| |
| Cs-137 D Gamma Gauges 2493 2.20E+06 6,533 8,833 383 0.34 4.4 337 !
| |
| Cs-137 T Other 1060 1.40E+06 4,139 13.163 1,316 0.50 6.6 503 Sr-90 E Beta Gauges 1153 3.94E+05 3,895 11,387 342 -
| |
| 5.7 -
| |
| f H-3 E Beta Gauges 2390 2.47E+07 2,438 3,439 10,317 -
| |
| 1.7 - ,
| |
| Cs-137 E Beta Gauges 31 6.05E+05 1,795 195,228- 19,523 7.5 98 7,455 H.3 O lon Gen. Static E8m. 1392 1.51E+07 1,494 3,617 10,852 - 1.8 -
| |
| ! H-3 D Gamma Gauges 1176 1.38E+07 1,364 3,909 11,727 - 2.0 -
| |
| i po-210 U X-Ray Fluorescence 535 9.23E+03 972 5,749 17 9.09E-08 2.9 9.09E-05 Cm-244 D Gamma Gauges 16 2.16E+03 642 135,307 135 0.0087 68 6.7 i TI-204 E Beta Gauges $38 4.05E+06 607 3,764 7,528 0.0084 1.9 8.4 N lon Gen, Chromatog 1378 5.62E+06. 556 1,360 4,079 -
| |
| 0.68 -
| |
| H-3 Am-241 O lon Gen. Static EIm. 107 1.47E+03 545 17,184 14 0.0043 8.6 4.3 Gen Neut Src Apps 3 1.36E+03 504 565,999 453 0.14 283 142 !
| |
| Am-241 H Sr-90 D Gamma Gau0es 104 2.28E+04 228 7,324 220 -
| |
| 3.7 - t Pu-238 U X-Ray Fluorescence 12 3.52E+02 1f6 32,620 29 0.0023 16 2.3 -
| |
| 2 5.84E+01 19 32,425 29 0.0023 16 2.3 Pu-238 D Gamma Gau0eis Cs-137 H Gen Neut Src Apps 45 3.56E+03 ff 791 79 0.030 0.40 30 NI-63 N lon Gen, Chromatog 3451 2.59E+05 9 8 75 - 0.0042 - t X Ray Fluorescence 18 2.60E+03 8 1,445 144 0.055 0.72 55 i Cs-137 U i E-11 ;
| |
| l-ww,
| |
| | |
| Table Task-3-1C continued. Sorted by " Fractions of Tc,talIng. ALIs.'
| |
| '87 mru '92 som of AveraSe Dose Dose from Total Fractions of . Source from Spending Number Sum of Totaling. Average Strength, Ingestion 1000 hours Dev- of Activity Alls Activityl Gamma *A of 10^-4 at 1 m from Ice Devices, decayed to decayedto Average Ing. Device (remlh of Source Source Nuclide Code Device Definition '87'92 1992 (mCl) 1992(ppm) Allst Device (mCl) Gim) (rems) (rems)
| |
| Other 7 2.00E+01 7 3,572 2.9 8.96E-04 1.8 0.90 Am-241 T 5 2.25E+03 7 4,503 450 0.17 2.3 172 Cs-137 W Self-Lum L10ht Src '
| |
| Ni-63 W Self-Lum L10ht Src 19 1.20E+05 4 702 6,320 - 0.35' -
| |
| Po-210 E Beta Gauges 98 3.85E+01 4 131 0.39- 2.07E-09 0.065 2.07E-06 Self-Lum Light Src 1 4.42E+04 3 11.055 44,220 1.18E-04 5.5 0.12 Pm-147 W Gamma Gauges 41 4.13E+ 03 2 201 101' O.14 0.10 138 Co-60 D Sr-90 R Gas Sources 135 2.44E+02 2 60 1.8 - 0.030 .
| |
| U X-Ray Fluorescence 811 6.51 E+04 2 9 80 -
| |
| 0.0045 -
| |
| Fe-55
| |
| ;alib Src A>30 MCI 17 6.49E+02 1.9 382 38 0.015 0.19 15 Cs-137 i lon Gen Chromatog 11 3.13E+02 0.93 284 28 0.011 0.14 11 Cs-137 N Sr-90 T Other 182 8.63E+01 0.85 16 0.47 -
| |
| 0.007tv -
| |
| N lon Gen. Chromatog '13 6.43E+02 0.64 165 49 0.0072 .0.082 7.1 Ti-44 X-Ray Fluorescence 567 6.25E+02 0.62 4 1.1 2.03E-04 0.0018 0.20 Cd-109 0 Gamma Gauges 11 6.20E+03 0.46 141 564 1.51E-06 0.070 0.0015 Pm-147 D Cd-109 E Beta Gauges 85 4.60E+02 0.45 18 5.4 0.0010 0.0090 1.0 Cf-252 H Gen Neut Src Apps 16 1.60E+00 .0.24 50 0.10 - 4.18E-06 0.025 0.0042 Ra-226 S Foll Sources 53 1.24E+00 0.18 12 0.023 3.34E-05 0.0058 0.033 lon Gen, Chromatog 167 4.91E+02 - 0.16 3 2.9 0.0034 0.0016 ' 3.4 Sc-46 N Calibrators 109 4.64E+01 0. f 4 4 0.43 1.62E-04 0.0021 0.16 Cs-137 Y S ' Foil Sources 25 1.25E+03 0.12 17 50 - 0.0083 -
| |
| H-3 W Self-Lum Light Src 267 3.67E+01 0.12 2 0.14 -
| |
| 7.64E-04 -
| |
| C-14 Foll Sources 2 2A0E-01 0.10 162 0.13 4.07E-05 0.081 0.041 Am-241 S Other 644 5.88E+02 0.087 0 0.91 4.16E-04 2.28E-04 0.42 Ba-133 T Calib Src A>30 mCl 9 2.17E-01 0.080 30 C.024 7.55E-06 0.015 0.0076 Am-241 l Gamma Gauges 2 4.00E+01 0.059 100 20 - 0.050 -
| |
| Ru-106 D Gamma Gauges 44 5.27E+01 0.052 4 1.2 2.21E-04 0.0020 0.22 Cd-109. D Other 10 4.50E+01 0.048 16 4.9 9.03E-04 0.0082 0.90 Cd-109 T lon Gen Chromatog 1 1.30E-01 0.048 162 0.13 4.07E-05 0.081 0.041 Am-241 N E-12
| |
| ^ ~~~ J ? _~- -_ ~~__ _ - -_ _____ _ _ _ _ _-_ - - - __ - __ - __ _ _ _ _ -
| |
| | |
| .3 Table Task-3~1C continued. Sorted by ." Fractions of Total Ing. AL.Is."
| |
| '87 thru '92 sum of ' Average Dose Dose from Total Fractions of Source from Spending Number Sum of TotafIng. Average Strength, Ingestion 1000 hours Dev- of Activity Alls Activity / Gamma *A of 10^-4 at 1 m from ice Devices, decayed to decayedto Average Ing. Device (remth of Source . Source Nuclide Code Device Definition '87*92 1992 (MCI) 1992 (ppm) Alls / Device (mCl) 81m) (rems) (rems)
| |
| Pm-147 T Other 1 5.90E+02 0.044 147 590 1.58E-06 0.074 0.0016 Fe-55 D Gamma Gauges 57 1.10E+03 0.036 2 19 -
| |
| 0.0011 -
| |
| H-3 T Other 20 2.93E+02 ' O.029 5 15 -
| |
| 0.0024 -
| |
| Ni-63 0 lon Gen. Static Elim. 56 8.28E+02 0.027 2 15 -
| |
| 8.21E-04 -
| |
| Cd-109 N lon Gen, Chromatog 8 2.65E+01 0.026 11 3.3 6.11E-04 0.0055 0.61 Fe-55 N lon Gen, Chromatog 18 7.67E+02 0.025 5 43 -
| |
| 0.0024 -
| |
| C-14 T Other 33 7.25E+00 0.024 2 0.22 -
| |
| 0.0012 -
| |
| Co-60 E Beta Gauges 3 3.52E+01 0.02f 23 12 0.016 0.012 15 Fe-55 T Other 12 5 30E+02 0.0f8 5 47 -
| |
| 0.0026 - .
| |
| Ra-226 T Other 12 1.20E-01 0.018 5 0.010 1.42E-05 0.0025 0.014 '!
| |
| Fe-55 E Beta Gauges 13 4.64E+02 0.015 4 36 -
| |
| 0.0020 -
| |
| N!.63 H Gen Neut Src Apps 22 3.3 E+02 0.0f f 2 15 -
| |
| 8.33E-04 -
| |
| C-14 R Gas Sources 42 2.10E+00 0.0069 '1 0.050 -
| |
| 2.78E -
| |
| Sr-90 I Calib Src A>30 mCl 1 5.00E-01 0.0049 17 0.50 -
| |
| 0.0083 - ;
| |
| Ra-226 I Calib Src A>30 MCI 1 2.00E-02 0.0030 10 0.020 2.85E-05 0.0050 0.323 l-129 T ~Other 24 2.96E+01 0.0029 0 1.2 1.55E-04 2.05E-04 0.16 Ni-63 S Foil Sources 6 6.99E+01 0.0023 1 12 -
| |
| 6.47E-04 -
| |
| C-14 E Beta Gauges 6 6.00E-01 0.0020 1 0.10 -
| |
| 5.55E-04 - ,
| |
| Ni-63 U X-Ray Fluorescence 8 3.91E+01 0.0013 1 4.9 . 2.72E-04 .
| |
| Eu-152 T Other 56 1.08E+00 0.00040 0 0.019 1.44E-05 1.21E-05 0.014 C-14 Y Calibrators 24 1.20E-01 0.00040 0 0.0050 -
| |
| 2.78E-05 -
| |
| Ba-133 1 Callb Src A>30 mCl 93 1.53E+00 - 0.00023 0 0.016 7.51E-06 .4.12E-06 0.0075 i C-14 N lon Gen, Chromatog 10 1.00E-02 0.000033 0 0.0010 -
| |
| 5.55E-06 - !
| |
| Am-241 Y Calibrators 6 6.79E-05 0.000025 0 1.13E-05 3.55E-09 7.08E-06 3.55E-06 Sr-90 Y Calibrators 2 2.41E-03 0.000024 0 0.0012 -
| |
| . 2.01E-05 -
| |
| Ba-133 N lon Gen, Chromatog 5 8.79E-02 0.000013 0 0.018 8.01E-06 4.39E-06 0.0080 E Beta Gauges , 22 2.17E-01 7.15E-06 0 0.010 -
| |
| 5.48E-07 -
| |
| Ni-63 T Other 12 1.19E-01 3.93E-06 0 0.010 -
| |
| 5.52E-07 Ni-63 -
| |
| E-13 r
| |
| _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ __ . _ . _ _ _ _ . _ _ . _ __.__.m ..______.mm___.____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ - m .m___________._.- _m._ _m_____ _ _ . _
| |
| | |
| Table Task-3-1C continued. Sorted by " Fractions of Totaling. ALIs."
| |
| '87 ffrv 12 sum of Avera0e Dose Dose from Total Freedons of Source from Spending Number Sum of Totefing. Average Strength, ingestion 1000 hours Dev_- of Activity Alls Activity / Gamma A of 10^-4 at 1 m from Ice Devices, decayed to decayedto Average Ing. Device (romm of Source Source Nuclide Code Device De6nition '87*92 1992(mCO 1992(ppm) AList Device Ru-106 E Beta Gouges (mc0 Sim) (rems) (rems) 1 9.65E-04 1.43E-06 0 0.0010 -
| |
| 2.41E-06 -
| |
| Bi-210 E Beta Gau0es 1 2.79E-89 1.04E-92 0 2.79E-89 -
| |
| 1.75E-92 -
| |
| Kr-85 D Gamma Gauges 291 1.20E+05 0 0 413 0.00E+00 -
| |
| 0.00E+00 Kr-85 E Beta Gau0es 623 2.43E+06 0 0 3,905 0.0081 lon Gen, Chromatog 6.1 Kr-85 N 135 6.94E+04 0 0 514 0.05E-04 '
| |
| lon Gen, Static Elim.
| |
| 0.80 Kr-85 O 128 3.44Ev02 o 0 2.7 4.21E-06 -
| |
| 0.0042 Kr-85 R Gas Sources 2 2.11E+03 0 0 1,054 0.0017 -
| |
| 1.7 Kr-85 T Other 11 2.20E+03 0 0 200 3.13E-04 -
| |
| 0.31 Kr-85 U X Ray Fluorescence 13 5.44E+01 0 0 4.2 6.55E-06 .- 0.0065 325681 i
| |
| E-14
| |
| * i
| |
| _ _ _ _ _ _ . - r-- - - - - - - - - - - - - - - - - - - - -- - -- -
| |
| | |
| l, l
| |
| Table Task-3-2 shows the total number of sources in each Device Code category. Since many of these contain activities greater than 20 mci, the numbers of sources used as a basis for the ORAU - )
| |
| Report would be less than these numbers reported here.
| |
| Table l'ask-3-2. Summary of All Database (Baggett 1993).
| |
| Device Code' Device Definition
| |
| * Number of Devices' 31.5D Gamma Gauges 24,679 31.5E Beta Gauges 18,336 j 31.5H General Neutron Source Applications 110 j 31.5I Calibration Sources, A > 30 mci 317 31.5N lon Genera: ors, Chromatography 9,474
| |
| $1.50 Ton Generators, Static Eliminators 37,084 31.5R Gas Sources 11,146 l 31.5S Foil Sources 827 31.5T Other 5,277 31.5U X-Ray Fluorescence 3,763 31.5W Self-Luminous Light Source 310,667 31.5Y Calibrators 577 31.7W Self.I-innne Light Source 71,315 TOTAL 493,572 t,
| |
| ' Attachment to letter dated Nov. 2,1993, from S.L. Baggett to DJ. Strom, entitled " Summary of all Database."
| |
| 'U.S. Nuclear Regulatory Comnussion Regulatory Guide 10.10, Appendix C j (NRC 1987).
| |
| l Taylor (1939) eshmaannt that there were "approximamly 30,000 general licenses in non-Agreement States using about 400,000 devices, and about twice this many in Agraament States." Using !
| |
| Taylor's overall factor of 3 and the figure of 493,572 provided by NMSS, one is led to conclude i I
| |
| that about 3 x 493,572, or roughly 1.5 million generally licensed sources exist in the USA.
| |
| Registration rates for the various Device Codes are shown in Table Task-3-3. Production figures ,
| |
| are presented in the " Enclosure 1"' and ORAU report (see Attachment 2, Table 1, from the i
| |
| 'The data identified as " Enclosure l' data, entitled " Estimated number of generally licensed l devices and materials," dated 4/21/87, have been ascribed by S.L. Baggett to "part of an older i Commission paper or NRC staff report." The exact provenance of this data table is uncertain, except that it was ultimately supplied to PNL by NMSS for this review. ;
| |
| E-15
| |
| | |
| i i
| |
| ! ORAU Report). Production appears to have remained relatively constant for the major categories
| |
| ] of sources identifuxi in the ORAU report. Thus, a constant annual production was assumed to e.timate the total number of remaining sources (with the assumption of decay for short-lived source %, .3 shown in Table Task-3-4. There is rema.nble agreement between the 1987-92 figures and the data reported in earlier work.
| |
| Tne data in Tables Task-3-3 and Task-3-4 do not include isotopes, dates, and activities, and thus, air of very limited use in risk analysis.
| |
| The 1987 production and ammmincian of total quantities of generally licensed sources is given in the " Enclosure l' table, as shown in Table Task-3-5. A coree between the " Enclosure 1" categories and the ORAU categories is given where possible.
| |
| Table Task-3-3. Annual device registration rates based on 6-year averages (1987-1992) from
| |
| " General License Database System Report for Peer Review."
| |
| Code Device Definition per year per 6 years D Gamma Gauges 1083 6498 E Beta Gauges 1289 7731 H General Neutron Source Applications 14 86 ;
| |
| I Calibration Sources, A > 30 mci 20 121 I N Ion Generator, Chromatography 866 5197 O Ion Generator, Static Elimmator 4820 28918 R Gas Sources 1525 10950 S Foil Sources 14 86 T Other 347 2084 U X-Ray Fluorescence 426 2553 W Self f nmi=>s Light Source 40466 242797 W7 31.7 Self inmi-2= Light Source
| |
| - 3087 18519 Y Calibrators 24 141 TOTALS 54280 325681 E-16 m , e e.--
| |
| | |
| s I
| |
| l i Table Task-3-4. Annual Production Rate and Numbers of Sources in Non Agreement States from ;
| |
| i the " Enclosure 1" table dated 4/71/87 and from Table 4 of the 1987 ORAU Report by ORAU i j Repon Category.
| |
| : )
| |
| ! " Encl. 1 " " Encl. i
| |
| { 4/21/87: l'
| |
| ; Anmini 4/21/87 ORAU J
| |
| ! Production Total as 1987 !
| |
| ! ORAU Category Race oJ 1987 Table 4 I
| |
| ! A-1 Static Riminatars: Hand-Held / Portable /Small Brushes 7000 20000 20000 l (10 CFR 31.3)
| |
| A-2 Static Eliminatars or Detectors: In Equipment or 80000 160000
| |
| ~
| |
| Process Line (Very High Toxicity Sources) (10 CFR 31.5) 170000 l A-3 Static Eliminators or Detectors: In Egi==>t or 120 9600
| |
| : Process Line (Iew Toxicity Sources) (10 CFR 31.5) 1 j B Gamma Gauges (10 CFR 31.5) 337 16000 4200 l C-1 Beta Gauges: Backscatter Type (10 CFR 31.5) 800 7000 8000
| |
| , C-2 Beta Gauges: Transmission Type (10 CFR 31.5) - -
| |
| D Gas Chromatographs (10 CFR 31.5) 8000 l E-1 X-Ray Fluorescence Analyzers: Very High Toxicity 90 720 i Sor.rces (10 CFR 31.5) 720 i E-2 X-Ray Fluorescence Analyzers: Moderate Toxicity - -
| |
| j Sources (10 CFR 31.5) s
| |
| ; F Sources for Checking Detector Operation or Calibration - - -
| |
| i and Analyncal Reference Sources (10 CFR 31.5)
| |
| G-1 Self Imminmie Devices (10 CFR 31.5) 20000 180000 180000
| |
| ] G-2 Self f nmiamin Devices in Aircraft (10 CFR 31.7) 7600 90000 90000 4
| |
| j H Analytical Instruments Ceiaia: Small Calibration or > 600 7000 7000 j Reference Sources (10 CFR 31.5) (Liq Scint)(Liq Scint)
| |
| ! I Sources for Checking Detector Operation or Cahbration 60 480 2000
| |
| / and Analytical Reference Sources (10 CFR 31.8)
| |
| J Small Quantities of Source Material (10 CFR 40.22) 70 kg 2000 kg -
| |
| K Sources for Cbxking Detector Operation or Calibration Pu-239 NA -
| |
| and Analytacal Reference Sources (10 CFR 70.19) .
| |
| TOTAL (except J,K) -
| |
| 506745 487920 h i
| |
| E-17 I
| |
| | |
| k I
| |
| i l
| |
| l Table Task-3-5. Estimated Number of Generally Licensed Devices and Materials (" Enclosure 1,"
| |
| ! 4/21/87). ORAU Categories are given where there is a clear coris; - '- E-w between the
| |
| ; " Enclosure 1" categories (i.e., Regulatory Guide 10.10; NRC 1987) and the ORAU caegories.
| |
| 4 1 l
| |
| 1 Number
| |
| ! of ORAU
| |
| ! Dev- Devices Total Caegory I j CFR ice Sold per Number of (if Sec. Code Device Type Year Devices applicable) l I
| |
| 31.3 Static Elimmawr 7000 20000 A-1 31.5 Aerosol Neutralizer 120 9600 A-3 l l
| |
| 31.5 E Beta Baciracaner Gauge 800 7000 C-1 1 31.5 N Electron Capture Detector 900 8000 D l 31.5 Electrostatic Voltmeter 890 3000 H j 31.5 Fill level Gauge 600 4200 B I 31.5 Fuel Densitometer Emitter 200 945 B i 31.5 D Gauging Devices (Part I) 337 16000 B j 31.5 'In Flight Blade Inspection Systems 200 1000 B
| |
| }
| |
| 31.5 Liquid Scintillarian Sman.5 600 7000 H !
| |
| 31.f W Self Inminniin Exit Signs 20000 180000 G-1 31.5 Static Eliminatars/ Meters 80000 160000 A-2 31.5 X-Ray Fluorescence Wwer 90 720 E-2 4 31.7 Self f nminmin Aircraft Signs 7600 80000 G-2 !
| |
| 31.8 Calibration or Reference Sources 60 480 l
| |
| 40.22 Source Material (Depleted Uranium) 140 2000 J j TOTAL 119397 497945 l
| |
| 'Ibe data provided by NMSS (Baggett 1993) shows 71,315 devices in Device Code 31.7W, self-lumianus devices in aircraft. This is in reasonably good agreement with the 10 CFR 31.7 entry in the "Fselannre 1" Table (80,000) and with the ORAU estimate for Caegory G-2 (90,000).
| |
| 7he " Enclosure l' Table, praarad in 1987, shows 497,945 devices, and the NMSS table, -
| |
| praared 6.5 years later in 1993, shows 493,572. This is a remarkable coincidence with the ORAU Table 4 Total of 487,920 devices. One could safely conclude that there are about a half a
| |
| ~
| |
| E-18
| |
| | |
| I i
| |
| l 1
| |
| 4 l ,
| |
| I i million generally licensed devices in existence under NRC General License, with perhaps twica 3
| |
| that many in Agreement States (Taylor 1989).
| |
| In general, it has not been possible to establish a correspondence among the ORAU categories. the
| |
| { ,:ategories in the " Enclosure 1" Table of 4/21/87, and the data provided by NMSS (Baggett 1993).
| |
| { ' Ibis has been confumed in the {{letter dated|date=March 30, 1994|text=letter dated March 30,1994}}, from S.L. Baggett to D.J. Strom. To
| |
| ! apply the ORAU risk analysis :"ethods, it will be necessary to resolve what numbers of devices of ,
| |
| what designs, sources, and activities are in use. This will require significant additional investigation and direct access to the database.
| |
| i i 4.0 TYPES, FREQUENCIES, AND RELATIVE SEVERITIES OF IMPROPER j TRANSFER / DISPOSAL OCCURRENCES NMSS inspection report summaries (Piner 1990, Wheaton 1993) and the operational experience reports and bulletins issued by the NRC's Offa.c of Analysis azul Evaluation of Operational Data (AEOD) provided information on tb types, frequencies, and relative severities of improper transfer / disposal occurrences having actual or potential public dose implications. These data were accessed through written reports and the Nuclear Regulatory Event Report (NRER) database files listed in Attachment 1. These files covered event reports during the period 1980 through 1992.
| |
| The NRER database was searched for occurrences of anything leading to events concerning ,
| |
| generally licensed materials. Such events were found with a code of " GEN" or "GL." in the !
| |
| LIC_NO field, and in a couple of other instances. Many of the entries, however, either did not j, involve generally-licensed devices or materials, or involved materials whose range of activities 1 enc = led the activity limits in the ORAU Report. The latter groups were categorized as "L," and ,
| |
| I tabulated with the other ORAU Report Categories. Several huadred entries under "NL" (not licensed) were also reviewed, revealing no events involving generally-licensed devices.
| |
| Results of the survey are summarized in Table Task-34 (number of incidents) and Table Task-3-7 (nwnber of sources or devices). Table Task-34 shows incident rates by rows labeled with ORAU ,
| |
| Category. Column entries are suunbers of incidents of each kind that the PNL reviewers found in l!
| |
| the NRER danh==a. Since a given incident may involve more than one source, Table Task-3-7 ji shows mimher of devices involved in the incidents listed in Table-34 by rows labeled with ORAU Category. Many event reports covered multiple sources, includmg various nuclides ("Co, '''Po and .2''Pb in one case, and d'Am, 2''Po, and "Kr in another case). By far the most common occurrences were in Category A2, static elimination sources, with 2''Po being the most commonly-involved nuclide.
| |
| In the tables, an incident can be classified in more than one category of outcome, such as lost and recovered, or damaged and leaking.
| |
| In most cases, the NRER database gives no information concerning severity. Sometimes a description like "significant event" occurs. On only a rare occasion is a dose to a person mentioned. Occasionally a description like " completely destroyed in a fire resulting in contamination" occurs.
| |
| E-19 l:
| |
| | |
| ... . -- .-. ..- - . . - - . - . - . . - - . - _ _ . . . . - . - . . . - . . .. .~. . - . -
| |
| i i
| |
| l g..
| |
| [ Table Task-3-6. Number of incidents involving generally-licensed devices. Data are from the i
| |
| NRER database files listed in the attachment, for the period 1980-1992. Category "L" involves sources whose activities exceed the upper limits on activity in the ORAU Report.
| |
| ORAU .Incid. Land- Scrap- Dam- Recover- Mis-Cat. ents* IAst Stolen fill yard aged Le*ing ed use Al 3 2 1 A2 49 28 2 5 1 9 12 2 1 A3 1 1 I B 6 4 1 1 1 1 1 C1 1 1 C2 1 1 1 i D 1 1 El 2 2 E2 2 1 1 F
| |
| G1 11 1 5 2 2 2 1 G2 1 1 H 3 1 1 1 1 I
| |
| J 3 1 1 ,
| |
| 1 K '
| |
| L- 30 8 4 1 5 6 2 5 4 !
| |
| Total 114 50 12 8 10 20 19 11 6 Total w/o L 84 42 8 7 5 14 17 6' 2
| |
| * Incidents may be less than row total because some reports are cataloged multiple times, e.g., " stolen and recovered."
| |
| ~*
| |
| E-20
| |
| | |
| a Table Task-3-7. Number of devices involved in the incidents listed in Table Task-3-6. Data are from the NRER database files listed in the attachment, for the period 1980-1992 (there were no entries for 1980-82, so this is a 10-year period). Category "L" involves sources whose activities exceed the upper limits on activity in the ORAU Report.
| |
| 4 ORAU Land- Scrap- Dam- Recover- Mis-Cat. Devices
| |
| * Lost Stolen fill yard aged Leaking ed use Al 14 9 5 -
| |
| A2 89 54 2 5 3 28 27 4 3 A3 1 ;
| |
| B 9 6 1 1 2 1 3 Cl i 1 C2 1 1 1 D 1 1 El 2 2 E2 2 1 1 F
| |
| G1 57 1 43 2 9 9 1 G2 1 1 6 II 5 1 1 3 1 I l!
| |
| <1 J 4 2 1 ,
| |
| t K
| |
| L 113 12 51 1 7 35 34 6 7 >
| |
| i Total 300 90 97 8 14 80 75 13 13 Total t w/o L 187 78 46 7 7 45 41 7 6 ,
| |
| I
| |
| * Devices may be less than row total because some reports are cataloged multiple times, !
| |
| e.g., " stolen and recovered."
| |
| There were 41 incidents tabulated in the ORAU report (Table 4 in Stabin et al.1987, p.13). Our i analysis, displayed in Tables Task-3-6 and Task-3-7, shows 114 incidents involving 300 devices ;
| |
| over a 10-year period, for rates of 11.4 incidents per year and 30 device-incidents per year. The l ORAU in idents were based on only a few years worth of data, and included data from agreement state reports. It is also unclear whether the ORAU Report included numbers of devices involved (see Table Tast-3-7.) Thus, the two totals are not strictly comparable. If the "L" incidents listed ,
| |
| in Table Task 3-7 are not counted, the rates become 8.4/ year and 18.7/ year. l i
| |
| , Incident breakdown by nuclide and ORAU source category is shown in Table Task-3-8. It is f
| |
| ' difficult to justify the relevance of the "L" category for this study. Ignoring L incidents, many of l
| |
| _.. j i
| |
| E-21 l
| |
| | |
| which may be improperly classified in the database, the bottom row gives the total number of incidents for each nuclide. The nuclides,2''Po and 2''Am, dominate with 44 and 17 occurrences, respectively, out of a total of 84 incidents. Clearly, row totals show that incidents are dominated by static elimination sources (49 of 84 generally-licensed incidents), with self luminous devices following at 11. -
| |
| Table Task-3-8. NRER Incident Breakdown by Nuclide and Source Category. Table entries are l number of incidents per 10 years. !
| |
| ORAU DU Cat. ''' Am '"Cd "Co "'Cs 2'*U 'H "Kr "Ni '''Pm 23.Po "Sr Z Total Al 3 3 i 40 5 49 !
| |
| A2 4 I
| |
| A3 1 1 B 1 5 6 ,
| |
| Cl 1 1 C'e. I 1 :
| |
| D 1 1 El 2 2 E2 2 2 F 0 G1 5 5 11 i G2 1 1 H I 1 1 3 ;
| |
| 0 I )
| |
| J 3 3 K 0 L 10 2 6 5 3 1 ?. 30 Total 17 2 2 12 3 10 7 2 5 44 1 9 114 Total w/o L 7 2 0 6 3 5 7 2 2 43 1 6 84 4
| |
| E-22 i
| |
| | |
| ==5.0 CONCLUSION==
| |
| S AND NEED FOR FURTHER WORK
| |
| 'Ihe NRER data base gives a good idea of how many incidents have been reported. It would be good to find a formal method of estimating the number of unreported incidents, and for this a survey of field regulatory personnel may be the best approach.
| |
| A better study of the actual numbers of sources in use and in storage would be desirable. A more formal survey of manufacturers would provide a good basis for risk estimates for short-lived nuclides, such as Po-210, which appears to dominate the numbers of sources.'
| |
| The readily available information examined here should provide an adequate basis for an order-of-magnitude risk analysis, which would be better than many risk analyses in non-radiation fields, e
| |
| i I
| |
| e E-23
| |
| | |
| 1 i i I u
| |
| | |
| ==6.0 REFERENCES==
| |
| 1 Ayers, J. 1985. " General License Study - Analysis of Hazard." Internal memo to Division Director through channels, March 1985, attributed to J. Ayers. 8 pp. Washington, DC: Scaled i Source Safety Section, U.S. Nuclear Regulatory Commission.
| |
| 1 Baggett, S. 1987. GeneralUcense Study Report. vii + 127 pp. . Washington, DC: Sealed Source Safety Section, U.S. Nuclear Regulatory Comnussion. -
| |
| h Baggett, S. L. 1993. Attachments to letter addressed to Daruel J. Strom, dated November 2, i 1993. 'Ihe four hardcopy reports from " General License Database System" included one computer printout of six pages (plus a cover sheet) dated 10/22/93; one computer printout of two pages j!
| |
| dated 10/20/93; one undated typewritten page titled " Summary of All Database;" and one undated
| |
| {
| |
| 33-page report listing license numbers, model numbers, type numbers, and registration numbers for sources. )
| |
| )
| |
| I Dean, C. M., M. S. Lawrence, and H. D. Lester. 1991. Report on Survey of General Ucensees Under 10 CFR 31.5. NRC FIN D 2554-0. ICF Inc., Fairfax, Virginia.
| |
| "GRAU Report:" see Stabin et al.1987.
| |
| 'Piner, E., and B. Smith. 1990. Conformance with Regulatory Requirements. Also known as
| |
| " Inspection Report Summaries."- 50 pp. Washington, DC: Scaled Source Safety Section U.S.
| |
| Nuclear Regulatory Commission.
| |
| Scaled Source Safety Section and Policy and Procuiures Section, U.S. Nuclear Regulatory Commission. 1992. Program Code Descriptions Used in NRC Ucensing and Inspection Programs. 56 pp. Revision 2. Washington, DC: U.S. Nuclear Regulatory Commission.
| |
| Stabin, M., K. Paulson, and S. Robinson. 1987. Improper Transfer / Disposal Scenariosfor Generally Ucensed Devices. "The ORAU Report" produced under NRC FIN B0299. Oak Ridge Associated Universities, Oak Ridge, Tennessee. .
| |
| Taylor, J. M, 1989. StafInitiatives on the GeneralUcense Program. Memo to the Commissioners of the U.S. NRC. SECY-89-289. Includes 4 attachments. Washington, DC:
| |
| U.S. Nuclear Regulatory Commission.
| |
| l Wheaton, A. 1993. Analysis of Selected Incidents. Also known as " Operational Experience l Reports." Revision 2. Previous editions by E. Piner and B. Smith. Washington, DC: Scaled l Source Safety Section, U.S. Nuclear Regulatory Comnussion.
| |
| 1 U.S. Nuclear Regulatory Commission (NRC). 1987. Guide for the Preparation of Applications for Radiation Safety Evaluation and Registration of Devices Containing Byproduct Material.
| |
| Regulatory Guide 10.10. U.S.N.R.C., Washington, DC. -
| |
| E-24
| |
| | |
| l i
| |
| . i l
| |
| TASK 3-A'ITACHMENT 1: DOS FILES RECEIVED AND REVIEWED BY PNL l
| |
| J l The following list contains the most recent versicas of files reviewed by PNL staff for this task. '
| |
| ADDRESS DBF 176066 06-03-93 3:33p D:\NMSS\SSD 1 j B-BITXT CSV 413152 04-07-93 4:56p D:\NMSS\ INCIDENT i
| |
| j BYCODE NTX 14336 06-03-93 2:41p D:\NMSS\SSD CATALOG CAT 439 03-17-92 10:17a D:\NMSS\NREP,
| |
| ! COMPANY 'NTX ' 67584 06 43-93 2:41p D:\NMSS\SSD '
| |
| i CUSTOM DBF 25149 06-03-93 12:17p D:\NMSSiSSD i
| |
| CUSTOMI NTX 5120 06-15-93 4:49p D:\NMSS\SSD 1 l CUSTOM 2 NTX 10240 06-15-93 4:49p D:\NMSS\SSD
| |
| , EXP DBF 2?'458 05-03-93 4:29p D.\NMSS\NRER i j EXP XLS 158633 08-25-93 10:23a D:\NMSS\NRER j
| |
| ! LIC NDX 49664 06-24-92 4:48p D:\NMSS\NRER )
| |
| t NRER ZIP 926512 01-14-93 11:08a D:\NMSS\NRER NRERO DBF 186138 05-14-91 10:40a D:\NMSS\NRER NRER1 DBF 250880 07-19-92 8:47p D:\NMSS\NRER NRER2 DB 587469 04-20-93 9:34a D:\NMSS\NRER NRER2 DBF 306150 01-11-93 2:05p D:\NMSS\NRER NRER2 DBT 195584 01-11-93 2:05p D:\NMSS\NRER NRER8 DBF i19808 05-23-90 2:06a D:\NMSS\NRER NRER80-4 DBF 939378 08-20-92 6:07p D:\NMSS\NRER-NRER85-7 DBF 769898 05-03-93 4:25p D:\NMSS\NRER NRER9 DBF 178818 01-04-80 1:48a D:\NMSS\NRER PRINUSE DBF 1398 12-21-91 9:57a D:\NMSS\SSD REGNUM NTX 73728 06-03-93 2:41p D:\NMSS\SSD REGNUM 1 NTX 552 % 06 43-93 2:41p D:\NMSS\SSD REGNUM 2 NTX 47104 06-03-93 2:41p D:\NMSS\SSD SSD ~DBF 2706829 06-03-93 3:33p D:\NMSS\SSD EXE 324608 06-15-93 4:59p D:\NMSS\SSD SSDS TEMP DBF 1155 04-26-93 7:18a D:\NMSS\SSD l TEMP NTX 152576 08-20-93 1:16p D:\NMSS\SSD UNTITLED CAT {
| |
| 439 03-17-92 10:17a D:\NMSS\NRER i YR86 DBF 38778 03-17-92 10:40a D:\NMSS\NRER l
| |
| O E-25 i
| |
| | |
| TASK 3 ATTACHMENT 2: Table 1 from the ORAU Report (Stabin et al.1987).
| |
| 2 i TABLE 1 CLASSES OF DEVICES FOR SCENARIO DEVELOPMENT
| |
| * 4 APPLICABLE REGULATORY RADIONUCLIDES AND SECTION** CLASS DEVICE MAXIMUM ACTIVITIES
| |
| ; 31.3 A-1 Static Eliminators: Po-210 - 0.50 sci (18.5 MBq)
| |
| ' Hand-Held / Portable /
| |
| Small Brushes j
| |
| 31.5 A2 Static Eliminators Po 210 - 100 mci (3700 MBq) or Detectors: In Equipment or Process An 241 - 0.0005 mci (0.0185 MBq)
| |
| Line (Very High Toxicity) Ra-226 - 0.0005 act (0.0185 MBq) 31.5 A-3 Static Eliminators H 250 mci (9250 MBq) or Detectors: In Equipment or Process Kr 2 aci (74'MBq)
| |
| Line (Low Toxicity) 31.5 B Camma Cauges Co 10 sci (370 MBq)
| |
| Cs-137'- 20 act (740 MBq) j An-241 - 20 sci (740 MBq)
| |
| Ra-226 - 10 aci (370 MBq)
| |
| ~
| |
| 31.5 C1 Beta Cauges: Sr 0.025 mci (0.925 MBq)
| |
| Backscatt:r Type T1-204 - 0.10 mC1.(3.7.KBq)
| |
| Ru-106 - 0.02,5 act (0.925 MBq)
| |
| .Pm-147 , 0.050 mci (1.85 MBql C-14-0.050 sci (1.35MBk)
| |
| Pb-210 - 0.010 sci (0.37 MBq) 31.5 C-2 Beta Cauges: Sr 20 mCL (740 MBq)
| |
| Transmission Type 31.5 D Cas Chromatographs Ni 20 mci (740 MBq)
| |
| H 1000 aci (37 CBq) 31.5 E-1 X-Ray Fluorescence Am-241.- 30 aci (1100 MBq)
| |
| ) Analyzers (Very High Toxicity) Cm 244 - 100 sci (3700 MBq)
| |
| E-26
| |
| | |
| i l
| |
| TASK 3 ATTACHMENT 2 (continued) l TABLE 1 CLASSES OF DEVICES FOR SCENARI DEVELOPMENT * - CONTINUED j l APPLICABLE l
| |
| RECULATORY RADIONUCLIDES AND '
| |
| SECTION' CLASS DEVICE MAXIMUM ACTIVITIES. !
| |
| l 31.5 E-2 X Ray Fluorescence cd 109 20 mC1'(740 MBq) )
| |
| Analyzers l (Moderate Toxicity) Te 100 sci (3700 MBq) e 31.5
| |
| . F C'alibration'or Cs 137 - 0.10 mci (3.7 MBq) ;
| |
| Reference Sources Co 0.01 r:1 (0.37 MBq) 4 Ra-226 - 0.004 mci (0.15 mBq)
| |
| J Sr-90 0.001 mci (0.037 MBq) i j 31.5 C-1 Self-Luminous H 5000 mCL (1J15 CBq)
| |
| Devices Kr 1700 mci (62.9 CBq)
| |
| C-14-0.10 mci.(3.7MBq) 1
| |
| . 31 7 ,
| |
| C-2 Self-Luminous. H 5000 sci (185 CBq)
| |
| ; Devices in Aircraft .
| |
| 4 Pm-147 - 300 aci (11 CBq)
| |
| , 31.8 H Analytical. Cs-137 - 0.040 mC1 (1.5 MBqi Instruments i containing Small Ni 15 mci (555 MBq) j . Calibration or Reference Sources .
| |
| ' I 31.8 I calibration or An 241 - 0.005 mC1.(0.185 MBq)
| |
| Reference Sources i 40.22 J Small Quantities of U 238 and Th-232 - 15 pounds at Source Material any one time, no more than 150 j pounds per calendar year l~
| |
| 70.19 K . Calibration or Pu-239 - 0.005 aci (0.185 MBq)
| |
| Reference Sources 4
| |
| * See Appendix for device descriptions
| |
| ** Code of Federal Regulations, Title 10 -
| |
| E-27
| |
| | |
| l APPENDIX F TO FINAL REVIEW OF THE 1987 REPORT BY !
| |
| OAK RIDGE ASSOCIATED UNIVERSnTES, j
| |
| " IMPROPER TRANSFER / DISPOSAL SCENAFIOS FOR l GENERALLY LICENSED DEVICFJ:"
| |
| TECHNICAL Lb'7TER REPORT: ,
| |
| TASK 6, DEVELOPMENT OF ADDinONAL PROBABILITY AND RISK l INFORMATION i
| |
| NRC JOB CODE L2536 j PNL No. 20278 ;
| |
| D. J. Strom!
| |
| R. L. Hill!
| |
| J. S Dukelow2 I
| |
| Health Protection Department 2
| |
| Nuclear Systems and Concepts Department .
| |
| Pacific Northwest I2boratory
| |
| -Richland, WashinLton 99352 January 31,1994 revised Febniary 4,1994 revised June 2,1994 Prepared for the U.S. nuclear Regulatory Commission Under Contract DE-AC06-76RLO 1830
| |
| | |
| i 1.0 EXECUTIVE
| |
| | |
| ==SUMMARY==
| |
| | |
| Task 6 of the project " Review of Improper Transfer / Disposal Scenarios for Generally Licensed Devices Study" is entitled " Development of Additional Probability and Risk Information." For this task, .. "the individual reviewers will develop additional probability and risk information in their area of expertise to assist the NRC staff's decision making regarding the need for regulatory action. This information may include refm' ed estimates of probabilities associated with selected disposal scenarios, assessments of the sensitivity of -
| |
| consequence and risk calculations to different assumptions and inputs, and quantitative l
| |
| estimates of individual and population risk resulting from selected improper disposal activities."
| |
| Two areas of additional risk information have been developed by the PNL reviewers. The first area of additional risk infomiation that was identified in PNL's preliminary review of )
| |
| the 1987 ORAU Report is the need for a mathematical framework or " formula" for the risk i
| |
| of radiological accidents. This framework should address two items of methodology missing I from the ORAU Report; the probabilities of initiation of accident sequences, and the sue of historically-derived probability distributions of accident consequences (which include worst cases as their extremes).
| |
| 1 The second area of additional risk information is a quantitative characterization of relevant '
| |
| historical accidents with sources, whether generally licensed or not. The quantitative characterization results in three numerical factors for each person involved in each accident for whom intake, whole body dose, and local dose values are published. The numerical values are " fraction taken in," "whole body time-and-proximity factor," and " local dose time- l and-proximity factor." These factors depend only on human behavior and accident circumstances, not on the amount, Kind, and quantity of radioactive material involved in the accident.
| |
| The factors can be used to predict more realitticr'ly the radiological consequences of future accidents than the use of " worst case" factors. Furthermore, over 10 accidents involved exposures to two or more people, resulting in distributions, rather than point '
| |
| estimates, of values. Such distributions can be used as inputs to modem probabilistic risk j calculations. '
| |
| i 2.0 PROBABiLISTIC MATHEMATICAL FRAMEWORK The first area of additional risk information that was identified in PNL's preliminary review of the 1987 ORAU Report is the need for a mathematical framework or " formula" for the risk of rdiological accidents. This framework uses individt'al and collective radiation dose as surrogates for risk, and considers both the magnitude and probability of occurrence of various doses. The PNL reviewers have identified two probability considerations as missing from the ORAU Report. The probabilities of initiation of accident sequences, and the likely l (rather than worst case) consequences are included. Using historically-derived probability I distributions of accident consequences (which include worst cases as their extremes) enhances
| |
| *the realism of risk estimates calculated from postulated accidents.
| |
| F-1
| |
| | |
| i
| |
| [
| |
| 2.1 ASSESSMENT OF THREE KINDS OF DOSES FOR USE AS SURROGATES FOR !
| |
| HUMAN HEALTH RISKS ,
| |
| Risk is conventionally defined as Risk = Probability x Severity. (1)
| |
| L
| |
| 'Ibere may be several separate components to the probability term: probability of an accident happening, probability of a given dose resulting when the accident happens, and probability ;
| |
| of that dose resulting in a stochastic health effect, for example. For human health risks due to radiation exposure, various dose quantities multiplied by suitable health risk coefficients - ;
| |
| may be used as surrogates severity (ICRP 1991). At low doses, severity may connote the l likelihood of a severe effect such as cancer occurring in an individual. At higher doses exceeding thresholds for deterministic effects, " severity" has a more conventional meaning .
| |
| for an individual, such as how serious a burn is.
| |
| Severity in Equation 1, for incidents involving sealed sources, can be defined both as individual doses and collective doses, that is, the sum of all doses accruing to all individuals in a given incident.
| |
| An individual tissue or organ dose equivalent, if below 50 rems, carries no risk of deterministic (formerly "non-stochastic") health effects (such as radiation burns, develor-dal abnormalities, etc.), but represents some degree of risk for stochastic effects (i.e., cancer and heritable ill-health). Such individual tissue or organ dose equivalents from internal and external exposure can be combined to form a total effective dose equivalent (TEDE), which is a modem surrogate for stochastic risk in individuals. Current risk estimates are on the order of 4 x 10-2 per sievert (4 x 104per rem) for adverse stochastic health outcomes in workers, and perhaps 5 x 10-2 per sievert (5 x 104 per rem) in the general public (ICRP 1991).
| |
| Under the linear, non-threshold dose response hypothesis for stochastic effects used for i radiation protection purposes, individual TEDE values can be summed to make collective I total effective dose equivalent (" collective dose"). Collective dose is a surrogate for l collective risk of adverse stochastic health outcomes in populations.
| |
| For improper transfer and disposal scenarios for generally licensed devices, it is also necessary to consider the possibility of individual tissue or organ doses that exceed thresholds for deterministic effects. Such doses result in certain injury, whether sub-clinical, mild, !
| |
| severe, or fatal, to the individual receiving the dose. Doses above a few tenths of a sievert l (a few tens of rems) should be expressed in absorbed dose units, i.e., grays (or rads), l specifying the radiation type, since the relative biological effectiveness of high linear energy transfer (LET) radiation (e.g., neutrons and a particles) at such dose levels is significantly l
| |
| : less than the quality factor used for limitation of stochastic effects (ICRP 1991). !
| |
| Furthermore, no clinical effects may occur whatsoever from protracted irradiation F-2 1
| |
| | |
| l significantly exceeding the traditional deterministic threshold of 0.5 Gy, since significant repair can occur between damaging events on a microscopic scale. For this reason, committed doses of long-lived, tenaciously-retained radionuclides will be poor predictors of deterministic effects.
| |
| 1 Thus there are three dose endpoints that should be considered in a risk analysis:
| |
| Distributions of individual TEDEs, the collective TEDE, and distributions of individual tissue j or organ dose above thresholds, such as 0.5 Gy for acute irradiation and perhaps 1 Gy or more for protracted irradiation.
| |
| Individual Sotchastic Risk = Probability x TEDE l l
| |
| Collective 5tochastic Risk Probability x Collective TEDE
| |
| =
| |
| {2)
| |
| Individual Deterministic Risk a Probability of Dose Above a Threshold j x Dose Efect Function The probabilities of various doses being received from a given improper transfer / disposal scenario are related both to the probability of the scenario occurring and the probability distribution of doses resulting from the scenario. Finally, risks are summed over all scenanos.
| |
| Ecological risk is the risk to ecosystems, habitats, and potential loss of access and usability of land and environmental resources. Al' hough improper transfer and disposal scenarios for generally licensed sources may result in er.ological risks, ' hey are not considered here.
| |
| 2.2 PROBABILISTIC RISK METHODOLOGY Since 1987, many changes have occurred in probabilistic risk methodologies. Recent summaries of these techniques are provided by IAEA (1989), Finkel (1990), and Morgan and Henrion (1990). In addition, the advent of user-friendly Monte Carlo simulation software for probabilistic health risk analysis, such as Crystal Ball ("Decisioneering, Inc., Denver, CO),
| |
| makes it feasible to perform probabilistic risk assessments for this kind of work.
| |
| 2.1.1 Risk Networks The draft ORAU report contains, for each class of Generally Licensed Sources, a " risk network" connecting the Initial Events (listed in Table 2 of the report) with the Final Status of Device conditions (listed in Table 3). Figure 1 is an example risk network from the draft report. Each of these networks starts with the assumption that a device of that class has been improperly transferred / disposed. Along the left side of the network is a collection of Initial Events, or states that the device can be found in after improper transfer / disposal. Along the right side of the network are the Final Status of Device conditions. In between is a
| |
| , collection of transition conditions and a collection of paths leading (from left to right) from the Initial Events, perhaps through one or more of the transition conditions, ending in one of F-3
| |
| | |
| . . - - > + . . - ~ . - ~ _ _ .-s..- .a - __ _ n.- . .. . .+. . _ . . . _= _ . - _ = - , . a.
| |
| CLASS A-2 EQUIPMENT STATIC ELIMINATORS e- u
| |
| , h P8"** W ''*8 mo
| |
| ''**** .o,panee meheve e
| |
| eJ eeresW G.1 ees.on ,.se
| |
| ~ e.e wien e e h,e,o,m ,,,c on,=. . . sed se nosendhetend os =m. ,
| |
| beerteuse h peemenadesi w g.)2a !
| |
| - - - _ _ "_^ !
| |
| >Het 9.S 9.1 onehe pene
| |
| [
| |
| Drotee seenwered h stooege se S..e t p.,e,,o ana.e,ana 8 88 89 4.3 essveyo ye e h essemer g ,g Seteege S.j: ) 8883*'
| |
| testets g ,3,3 "M 0.33 e gg e.e 3 ''**I h ,
| |
| pu edue en # 827 i Dowtue dheterded .1 r.o. ..l
| |
| --~
| |
| to erowterwi==d g '3
| |
| . 15 4.19 9 bis
| |
| '' ' I Cea .eenen , ggg gg smetaleto i
| |
| o,eno. e.ie
| |
| __ wies w de mase . . _ , .
| |
| tenerse !
| |
| P. . .e owb ..
| |
| Figure 2.2.1 An example of a risk network from the ORAU Report (Figure 3).
| |
| F-4
| |
| | |
| the Final Status of Device conditions. Associated with each of the path segments between two of the condition boxes is the conditional probability of the transition from the left box (i.e., the left end of the path segment) to the right box. Finally, along the right side of the network, associated with each of the Final Status of Device conditions is the conditional probability of ending up in that state, obtained by adding up the probabilities associated with each of the distinct paths through the risk network that terminate in that state.
| |
| d 2.2.1 Adequacy of ORAU Report Risk Networks We feel that the conditional probabilities assigned (using engineering judgment) to the path segments by the authors of the ORAU draft report are generally reasonable. There is, however, a structural aspect of these risk networks with which we take issue: all of the networks have a pathway leading from " trash handler" to " incinerator" and thence on to either " salvage dealer" (with conditional probability 0.5) or to the Final Status of Device condition " Device buried in sanitary landfill" (with conditional probability 0.5). This seems to presuppose that the trash handler sifts the incinerator ashes for metal / ceramic slag which is sent on to the salvage dealer, and that half the time (i.e., probability 0.5) the remains of the source are incorporated in that slag. The probability 0.5 of " source-in-slag" seems high. In addition, for those cases in which the bulk of the source is volatilized in the incinerator and released to the atmosphere, the box " incinerator" has effectively become a Final Status of Device condition for which health consequences should be assessed. Several potentially significant scenarios, such as an intact source out of a shield, and potentially significant consequences, such as doses to workers (rather than the public), have been omitted. Finally, one could argue that the incinerator box ought to be " downstream" of the salvage dealer, with incineration as one of the salvage dealer's options for dealing with items that
| |
| ! incorporate both salvageable (i.e., metals) and non-salvageable materials.
| |
| What is missing from this picture, if we desire to estimate the risks associated with the improper transfer / disposal of various types of Generally Licensed Devices? Missing is the probability of entering the risk network in the first place (i.e., the probability that a, device of that type will be improperly transferred / disposed) and the consequences associated with each of the Final Status of Device conditions.
| |
| 2.2.2 Censoring of Scenario Probabilities by Under-Reporting Table 4 of the draft ORAU report contains a computation of the fraction of devices of various types that have been improperly transferred /disnmed, based on very limited data.
| |
| These values cannot be directly used as improper transrer/ disposal probabilities because of an obvious censoring problem. Improperly transferred / disposed devices show up in Table 4 only if the improper transfer / disposal was detected in some fashion. This detection could occur if an appropriately labelled device was found somewhere it didn't belong, if an inspection of records and device inventory discovers that a device is missing, or if someone's 3 medical symptoms can be tied directly to the improper transfer / disposal of a specific device.
| |
| If a device was buried in a landfill or incinerated improperly, that improper transfer / disposal F-5
| |
| | |
| ~ _ _ - . - . - . - - - - . - . __.- - .- .-_- - - . - - . . . - - . .
| |
| i 1
| |
| 4 is unlikely to be detected, since inspections and audits of holders of General Licenses seem l
| |
| j to sample only a small fraction of the total population of General Licensees.
| |
| 2.2.3 Correcting Probabilities for Under-Reporting of Incidents i
| |
| i One way of dealing with this censoring would be to obtain an estimate for the probability i P(d) of the event d, where d denotes the detection, in some fashion, of the improper
| |
| ! transfer / disposal of a particular Generally Licensed Device.
| |
| i We can then take the fraction of devices of that class known to be improperly transferred / disposed, F (from Table 4), and increase it by the factor 1/P(d) to obtain the total fraction of devices of that type irsproperly transferred / disposed. That is, we assume that Fis telling us only about that subset of the set of improperly transferred / disposed devices for which the improper transfer / disposal is detected in some fashion; that subset is only the P(d)-
| |
| the part of the whole set of improperly transferred / disposed devices.
| |
| In other cases (see, for instance, the discussion of industrial process line static eliminators in Section 3.2 of the ORAU report), there is enough information to directly estimate the fraction of devices that are improperly transferred / disposed, without reference to Table 4. In this case, all of the devices are out on lease and the distributor simply offers to charge for an additional year's lease if the device isn't returned for legally authorized disposal.
| |
| Suppose we are going to estimate P(d). We might consider d to be the union or " sum" of.
| |
| two events:
| |
| a= the event that the improper transfer / disposal of the source is detected by observation of the label, by radiation detection, by inspection of records and inventory, etc.; and !
| |
| b= the event that radiation sickness or mjury is recognized and tied back to the i improper transfer / disposal of the source. .
| |
| We can then calculate P(d) = P(a) + P(b) - P(ab), W where ab is the set intersection of the events a and b. If P(a) and/or P(b) are relatively small, say 0.1 or less, then the term P(ab) will be second order and can be ignored, i
| |
| The probability P(a) will depend on a variety of factors, including the durability and intrusiveness of the labelling, the likelihood of " fortuitous" radiation detection, the ;
| |
| pervasiveness of inspections and audits, the unwieldiness of the individual sources, etc. By l J
| |
| fortuitous radiation detection, we mean something similar to discovery of high indoor radon -
| |
| - levels in the Reading Prong because a Pennsylvania nc' ear plant worker set off one of the portal detectors as he was arriving for work or the outside world's first knowledge of the F-6
| |
| | |
| i l
| |
| i l! Chernobyl accident when area detectors outside a Swedish nuclear plant started alarming. !
| |
| We could also reasonably expect P(a) to vary by type of source. !
| |
| i ,
| |
| Reviewing the history of improper transfer / disposal of sealed sources, even " strong",
| |
| specifically licensed sealed sources used in radiography or radiotherapy, we see dozens of i incidents in which a sealed source improper transfer / disposal is first detected by recognition I of radiation sickness or injury, followed by an investigation to determine the cause of that
| |
| , sickness. These detections are noted by the entry "Med" for " medical" detection of the
| |
| ; incident with the subsequent re-establishment of control over the source. '
| |
| 1 l j We know of no such incidents involving Generally Licensed sources or equivalent amounts j of radioactivity. Because of this, P(b) is likely to be significantly smaller than P(a) for most
| |
| ' generally licensed sources. It would be reasonable to model P(b) as directly proportional to the wornt-case and/or the average human health consequences of improper transfer / disposal of that t"pe of source.
| |
| We propose a model for P(b) of the following .ype:
| |
| P(b) =
| |
| k, f(H r, max) + (1 -k,) g(Hr.m) - (4) l where HT,ma, = dose to the critical organ for the maximally exposed individual; Hr.aq = dose to the critical organ for the average individual; ;
| |
| O 17 Hr.: u s Hg; L
| |
| H -H i f(Hr. max)
| |
| * ~ ' * ''"^^ "
| |
| -H g ,
| |
| 1 17 H T , max 2H, er
| |
| ~ '
| |
| Hg is the average individual threshold dose for developing clinical symptoms;
| |
| #c7 is the critical threshold dose for certain diagnosis of clinical symptoms; I g(H T) is defined similarly to f, using values Hthe, pop appropriate for failure to detect and Her, pop for sure detection of the incident in a whole population exposed to a given dose Hr; and k, = an appropriately defined weighting factor for balancing between detection l based on a symr'omatic maximally-exposed individual and detection based on symptoms in a r plation; 0 s k, s 1.
| |
| 2.3 RECOMMENDATIONS FOR IMPLEMENTING THE PROBABILISTIC RISK METHODOLOGY g To carry out the program defined in the preceding paragraphs, we need to do the following:
| |
| F-7
| |
| . .-- 2
| |
| | |
| I l
| |
| l 4
| |
| I I) For each Generally Licensed sealeo source type x, reassess the improper j transfer / disposal fractions, Fx, defined in Table 4 of the ORAU draft repon, on the basis of additional data, as available. The annual rates of incidents of improper transfer or disposal as a function of source category, including probabilities of incidents not being reported were not adequately assessed in the ORAU Report.
| |
| Improved data will be available from the revised Task 3 report.
| |
| : 2) For each Generally Licensed sealed source type x (i.e., A-1, A-2, A-3, B, etc.) and i for each radionuclide that can be utilized for devices of that type, pmduce estimates of either P(Mx) directly or of P(ax) and P(br). If P(ar) and P(bx) are estimated, then P(dx) = P(ax) + P(bx) and P(Mx) = Fx/P(dr). If Mx is the event that a device of j Type x is improperly transferred / disposed, we can use the probability P(Mx) to
| |
| " enter" the corresponding risk network (whetaer that probability is estimated directly l or obtained by estimating the probability P(d) and factoring up the fraction, F, fron.
| |
| Table 4). The probability P(Mx) will propagate through the network and each of the Final Status probabilities (from the ORAU draft) will be multiplied by P(Mx) to give the probability that a particular device of that class will end up in that Final Status of l Device. We can then multiply.that probability by the associated health consequences, Cx,fs, for device type x and that Final Status /s, to obtain the risk with a single device of that type ending up in that particular Final Status. Those risks can be summed for all of the Final Statuses to obtained the risk associated with a single device of that type, which can then be multiplied by the number of such devices extant to obtain the total risk to the public associated with that type of device.
| |
| : 3) For each Final Status of Device end state, produce estimates of the human health consequences resulting from a device ending up in that state, either usingm ' formation in the draft ORAU report or additional information, as necessary.
| |
| : 4) Define the risk associated with a particular type x of device in a particular end-state as:
| |
| =
| |
| R,,f, P(M,) p,,,-(consequences associated with the device x final state fs), (6) where px,fs is the end state probability from the corresponding ORAU draft report risk network.
| |
| : 5) Define the total risk, Rr, associated with a single device of type x as the sum of all l the Rx,fs over the set of final states o. Finally, the total risk a.mociated with devices of type x is the product of the number of devices of type x and Rx,fs.
| |
| : 6) We can do a simple uncertainty analysis by replacing all of the estimated probabilities and the probabilities given in the ORAU draft report by probability distributions and
| |
| '; using random variable aritlunetic to propagate those distributions through the calculations described above. The quick and dirty part would use just the mean and j i
| |
| F-8 l l
| |
| | |
| t variance of the distribution (or the mean and variance of the log-transformation of the distribution) and the Central Limit theorem to replace sums of random variables by the normal distribution with the appropriate mean and variance and replace products .
| |
| of random variables by the lognormal distribution with the appropriate mean and '
| |
| variance. These approximations give good msults for random variables with a central tendency and arithmetic calculations consisting of several sums and products (roughly '
| |
| speaking the more sums and products, the better the approximation).
| |
| 3.0 QUANTITATIVE ASSESSMENTS OF HISTORICAL IMPROPER TRANSFER / DISPOSAL INCIDENTS AND ACCIDENTS FOR INPUT TO PROBABILISTIC RISK METHODOLOGY Many new risk assessment tools have been developed since 1987 when the ORAU Report was finalized. Furthermom, the PNL reviewers believe that the pmbabilistic nature of risk assessment results rather than worst case scenarios need to be emp:msized. We have analyzed historical accidents involving sources to yield quantitative characterizations of human behavior in accident situations as input to probabilistic risk assessments. Nuclear weapons, nuclear fuel cycle, and criticality accidents and accidents involving accelerators or ,
| |
| x-ray machines have not been analyzed, since it is difficult to determine the relevance of these to improper transfer / disposal scenarios for generally licensed devices.
| |
| 3.1 DISTRIBUTIONS OF RESULTS RATHER THAN POINT ESTIMATES j We have performed a quantitative characterization of relevant historical accidents with sources, whether generally licensed or not, for those accidents for which necessary data are available in the literature. The _ quantitative characterization results in three numerical factors for each person involved in each accident for whom intake, whole body dose, and local dose values are published. The numerical values are "fractior uken in," "whole body time-and-proximity factor," and " local dose time-and-proximity factor " For the most part, these j factors depend only on human behavior and accident circumstances, not on the amount, kind, ;
| |
| and quantity of radioactive material involved in the accident. The factors can be used to mom realistically predict the radiological consequences of future accidents than the use of
| |
| " worst case" factors. Furthermore, over 10 accidents involved exposures to two or more l people, resulting in distributions, rather than point estimates, of values. Such distributiom can be used as inputs to modern probabilistic risk calculations.
| |
| 3.2 TIME-AND-PROXIMITY FACTORS FOR WHOLE-BODY AND LOCAL EXTERNAL IRRADIATION
| |
| )
| |
| There is a need for ris'K analysis for accidents involving single radionuclide radioactive I sources (as opposed to nuclear reactor or nuclear weapons accidents). Such risk analysis I requires knowledge of the probabilities and severities of such accidents. Historically, g accidents have involved anywhere from one to several thousand people.
| |
| F-9
| |
| | |
| 1 i
| |
| 1
| |
| .\
| |
| l Information for risk analysis of small, generally-licensed sources can be derived from accidents, usually involving large sources, that have already happened.
| |
| : The objectives of this study were to Develop generalized, quantitative descriptions of human behavior and interactions l with radiation sources from study of historical accidents; I
| |
| Evaluate applicability, advantages, disadvantages, and limitations of the approach; and 4
| |
| Identify additional information needed to apply these risk estimates to quantitative risk -
| |
| assessments for informed regulatory decision making.
| |
| i
| |
| . Historical records of accidental human interactions with radiation sources are used to develop
| |
| ]
| |
| distributions of extemal radiation exposure factors. These factors retain information about
| |
| ! human behavior from the accidents, but are independent of source strength or the radionuclide(s) involved. For external exposures, we define a " time-and-proximity" exposure
| |
| ' factor, F p, with and without provision for breach of source shielding, for each person involved in an accident. The distributions of exposure factors can be used to ;-redict ranges i
| |
| ' of possible radiation doses to individuals from a variety of accident scenarios involving different radionuclides, activities, and device designs. Use and limitations of the exposure factors and their distributions are discussed below.
| |
| l One limitation is the possibility that the exposure factor distributions based on accident information are altered when the accidents are discovered. For example, deaths or symptoms
| |
| : of acute radiation syndrome may lead to an investigation, discovery of an accident, and i termination of exposure. Distributions of facton: from such an accident may be of limited applicability to accidents that go undiscovered. I i
| |
| 3.2.1 Mathematical Description of Time and Proximity Factors 4 .
| |
| i For an unshielded point source, dose equivalent H depends on exposure time t (hours),
| |
| i distance r (meters), source strength A (activity in Ci) and isotope (through P in rem /hr m2 /Ci
| |
| } or Sv/h m2/Bq):
| |
| 3 H= 'Adt
| |
| -l 1
| |
| , =
| |
| P A dt j .
| |
| f(f)2 M i
| |
| 4
| |
| ! = PA12 r l t
| |
| l For each individual i, exposed in an incident for whom the dose equivalent H is known, the
| |
| ; F-10 4
| |
| 3 i
| |
| | |
| source isotope and activity are known, one can calculate 3 time-and-proxuutty factor for the incident:
| |
| H. g Fp,f = = , (g)
| |
| This factor is independent of both source strength and radionuclide involved.
| |
| Distributions of time,and-proximity factors can then be applied to similar accidents to determine external exposures, even those involving a radioactive source of different isotope and activity. One simple definition of F is p the number of hours one would have to spend at one meter from the source to receive a dose equivalent of H. j If the source remained partially or wholly shielded, then an additional factor should be introduced:
| |
| . Hg g
| |
| = =
| |
| F,*i ,
| |
| {9; F,l'A p,,2 where F, is the fraction transmitted through a shield, a number less than 1. F, can be taken as the dose rate at 1 m from the source (in its shield) to the unshielded dose rate at 1 m.
| |
| When sources are sutxiivided, the full activity is still used in the calculations because the human interaction is what we want to characterize, not the immediate sourre. So doses from the 1984 Mexican accident, for example, are atuibutable to the entire source.
| |
| ~
| |
| 3.2.2 Theoretical Limits Are Not Useful The theoretical upper limit on Fp is 1 hfetune .
| |
| (10) l (very close)2 ,
| |
| where (very close) represents a small distance from the source, e.g., 0.01 m 2 . For a weak 8 8 l source, this represents about 10 to 10 hours at one meter, a quantity so large as to be l useless. However, in tens of historical accident cases, a person (and in two of the worst
| |
| ; accidents, in Mexico in 1%2 [5 fatalitiesJ and Morocco in 1984 [8 fatalities], a child) has found an industrial radiography source and put it in a " hip pocket." In many cases, the
| |
| - "very close" is I cm or so, and exposure thnes have been up to several months. When large sources are involved, such as industrial radiography sources, these cases result in local radiation burns. The ratio of the average bone marrow dose to the dose at the site of the
| |
| , radiation burn is dependent on "how far the bone marrow is away from the hip pocket." In
| |
| ; ; many cases, the bone marrow to burn site dose ratio is over 100, sometimes over 1000. .In other words, one needs to look at tmly potential exposures which may result in a high dose, 4
| |
| l F-ll i
| |
| s
| |
| --.m. ,-- - .-, - :.-_- -
| |
| _r,.-
| |
| | |
| - - - - . --. . . - . .- .---- . -- - .. . _. .~ ,
| |
| 2 4
| |
| 1 i
| |
| but are realistic.
| |
| 3.2.3 Historically-Derived Values of Time-and-Proximity Factors !
| |
| Table 3.2.1 shows an analysis of 42 incidents involving external exposure. Accidents are
| |
| ! characterized by year of occurrence, by nuclide, source type and acdvity. The number of j people involved (broken down as public, workers, and cleanup workers) is given.~ Each
| |
| , accident is characterized by whether it involved brief or protracted whole-body irradiation,
| |
| , . brief or protracted localized irradiation, whether there were intakes, whether the source was
| |
| ! removed from the shield, and whether the source was damaged. References to the literature 1
| |
| are given. Specific dose equivalent rate constants are tabulated, along with dose rates at I
| |
| ! meter from the sources. The table contains average and geometric mean values of whole-body time-and-proximity factors, as well as minimum, maximum, and standard deviation and geometric standard deviation values, where appropriate. Also tabulated are maximum values
| |
| ! of skin dose or local irradiation time-and-proximity factors.
| |
| l These 42 incidents were chosen because of the availability of data on source identity, source j- activity, whole body and/or local doses to individuals, incident descriptions. The NRC's i Office of Analysis and Evaluation of Operational Data (AEOD) incident database does not, in l t general, contain the information needed for this kind of analysis. Nuclides include 241Am(1
| |
| ; accident), "Co (20),137Cs (4; note that Goiania is listed 4 times for various analyses), 331I
| |
| ! (1), and 192 Ir (16). Accidents included 21 industrial radiography sources, 8 sterilization
| |
| ; facilities, 5 teletherapy sources, 3 experimental sources, 2 brachytherapy sources,1 defense
| |
| ; incident, and 1 " radiation station." Members of the public we e involved in 14 incidents, in numbers ranging from 1 person to 4000 persons, totalling roughly 7000. Four accidents F (Mexico 1983 [4000], Morocco 1984 [28], Brazil 1987 [2800], and Pennsylvania 1992 [94])
| |
| { account for virtually all of this total 7000 people. There were 29 incidents involving exposure at work, and 6 incidents involved exposure to cleanup or recovery workers. ;
| |
| l i The weighted average of the whole body is 1.28 hours at a meter. This is far bek,w the j 3360 hours exposure time assumed as a worst case in the ORAU Report (i.e.,20 weeks x 7 j days per week x 24 hours per day = 3360 [page i, Stabin et al.1987). Ignoring the Texas child-abuse case of 1972, whole-body time-and-proximity factors ranged from essentially zero to 686 hours at a meter, the latter deriving from the 1%2 Mexican incident (which 5
| |
| formed the basis for the ORAU Report value of 3360). For only accidents exposing the
| |
| ; public, time-and-proximity factors averaged 117 hours at a meter (avera'ging over accidents) ;
| |
| 4 and 1.37 hours at a meter (using a weighted average over all accident victims, for which !
| |
| accidents with many victims dominate the average).
| |
| The 16 available maximum local time-and-proximity factors formed a highly skewed j distribution ranging from 0.05 to 24000 hours at a meter, with an average of 3100, a
| |
| ; standard deviation of 6900, and a geometric standard mean of 91. The average of 3100 is j . comparable to the 1%2 Mexico "Co accident whole body factor.
| |
| ~'
| |
| l -
| |
| 3 F-12 i
| |
| i 4
| |
| - . , - _ _ .r.- - - , -- _, . . , . _, e -----
| |
| | |
| Four accidents in particular are especially intere: ting, due to their nature and the potential for generalization to improper transfer and disposal scenarios for generally licensed devices.
| |
| These are the November 1992 Indiana, Pennsylvania accident; the 1983-84 Mexican "Co accident, and the 1987 Goiinia, Brazil accident, and the 1990 Korean source shipment accident.
| |
| ; The November 1992 Indiana, Pennsylvania case of a medical misadministration is particularly instructive for improper transfer and disposal scenarios for generally licensed
| |
| , devices (NRC 1993a, NRC 1993b). Doses are listed in the Appendix. This was a case of complete loss of control of the source, and the persons exposed < d not know about it until the source was essentially "out of harm's way." The 192 Ir source activity was 3.7 Ci 2
| |
| (1.37E11 Bq). The specific dose equivalent rate constant, P, is 0.48 rem /hr m /Ci. The central (i.e., halfway between the minimum and maximum doses) estimates of doses ranged from 0.0006 to 18.9 rems, with an averagc of 3.0 4.1 rems, a geometric mean of 0.29 rems 5 25 (i.e., a GSD of 25). The time-and-proximity factors ranged from 0.0003 to 8.6 for the 85 cases that the NRC (1993b) evaluated. A preliminary investigation of the distribution of time-and-proximity factors for this incident shows that, while the distribution is very broad and skewed, it is not lognormal (Figure 3.2.1). There were a number of persons who evidently received roughly the same doses due to similar duties involving patient care, so these may need to be treated separately from doses to the others involved.
| |
| The 1983-84 Mexican "Co accident (IAEA 1989) was also a complete loss of control of a large source (1.7E13 Bq (450 Ci) of"Co; P = 3.6E-13 Sv/hr m 2/Bq (1.3 rem /hr m2 /Ci)).
| |
| This incident resulted in an estimated 4000 persons exposed. From the data, the largest F p was 1.2, the mean of lognormal fits (with GSDs of 13.8 (uniform weighting) and 22.5 (Finney weighting)) were 0.004 to 0.008. In this situation, symptoms of acute radiation syndrome would have prevented much larger F p values, since the F p = 1.2 corresponds to 700 rems.
| |
| The 1987 Goiinia, Brazil accident presents four different populations for analysis, and it is difficult to determine whether some individuals may appear in more than one of these populations. Depending on how many persons are included, the time-and-proximity factors for the public plus the 9 workers involved with the source average 0.20, 0.11, or 0.0034 hours at a meter, with the former population including 97 persons, the next,249 persons, and the last,2812 persons. The maximum whole-body time-and-proximity factor was 1.46, and mmimum factors on the order of 10'9 The maximum local irradiation time-and-proximity factor was 4.2 hours at a meter. Figure 3.2.2 shows a histogram and cumulative density function for whole-body time-and-puimity factors from the GC. inia accident for the 46 highest dose cases, with a roughly logarithmic horizontal axis.
| |
| The 1992 Korean Accident (NUREG 1405, NRC 1990) includes dose estimates for 24 individuals. These are clearly lognormally-distributed as shown in Figure 3.2.3. Doses are
| |
| . listed in the Appendix. The average time-and-proximity factor was 1.3 3.7 hours at a
| |
| ' meter, with a geometric mean of 0.11 y 7.9 (GSD = 7.9), and a range of 0.0063 to 14.7.
| |
| F-13 l
| |
| l
| |
| | |
| A Time-and-Proximity Factor analysis has been performed for 40 accidents involving 231 individual doses. The complete results (231 lines of data) are p, resented in the Appendix.
| |
| The results are shown in Figure 3.2.4 and given in Table 3.2.2.
| |
| 3.2.4 Censoring of factors for high-dose accidents by the appearance of clinical symptoms of acute irradiation In many cases, accidents were discovered by the appearance of clinical symptorw of acute irradiation. In many cases, persons stopped receiving any more dose because they died. In such cases, it can be concluded that the accident did not run its normal course. A plot of time-and-proximity factors versus source strength, I'A, in sieverts per hour at im from an unshielded source, is shown in Figure 3.2.5. It shows a decreasing relationship between the time-and-proximity factor and soume strength. In most cases, the incidents involving a source with activity greater than 10,000 Ci resulted in fatalities. For many of the strong source accidents, medical symptoms were the first sign of an accident. The column in Table i 3.2.1 labeled "How Terminated" shows "Med" if medical symptoms appeared, and " Rad" if the accident was discovered by other means, usually by radiation measurements or by l discovery of n essing sources through inventory or malfunction. l Only a few accidents appear to have run their full course, that is, delivered all the dose they l ever would have. The 1992 Indiana, PA accident and the 1990 source shipment from Korea j fall into this category. In each case, the accident was essentially over when it was i discovered. These accidents are particularly valid for assessment of improper transfer and disposal scenarios.
| |
| 3.2.5 Collective Time-and-Proximity Factors for Accidents The collective dose can be determined using a Collective Time-and-Proximity Factor for an accident. Values of Time-and-Proximity Factors are given in Table 3.2.6 for the accidents at Goiania, Brazil; Indiana, PA; and the Korea-USA incident based on collective dose due to external irradiation (UNSCEAR 1993; NRC 1992; NRC 1990). Notice that'the Indiana, PA accident was an order of magnitude more serious than the Brazilian accident, and the Korea-USA shipment incident was a factor of 3 more serious, in terms of Time-and-Proximity Factors. That is, had the Brazilian source been involved in the Indiana or Korea accidents, the collective doses would have been 10 times and 3 times higher, respectively, than they were at Goiania. Using a similar rationale, if any of these three accidents had occurred with a small, generally licensed source, the collective dose would have been lowest for the l Goiinia-like accident, and highest for the Indiana, PA-like accident. j 3.2.6 Conclusions Distributions of Time-and-Proximity Factors from historical accidents can be used in
| |
| ; probabilistic risk analyses for both whole-body and local irradiation from external sources.
| |
| An analysis of 42 accidents for which source identity and strength are available show that the F-14 1
| |
| i l
| |
| | |
| 1 l
| |
| l i I
| |
| ' average accident victim gets a whole body dose equal to that from being at I meter from the !
| |
| accident's unshielded source for an hour. The average accident is characterized by a value of 46 hours at a meter. In other words, the population-weighted average is about I hour at a meter, while the accident-weighted average is 117 hours at a meter. Clearly, the accidents t with large numbers of victims (e.g., Goiania and Juarez) dominate the former average. The maximum value seen for whcle-body doses is about 700 hours at a meter. The average, '
| |
| geometric mean, and maximum values for local irradiation are 3100, 60 and 24,000 hours at a meter, respectively.
| |
| Such distributions should be used in probabilistic risk analyses to determine likely distributions of risks or doses from improper transfer and disposal scenarios for generally licensed devices.
| |
| I l
| |
| Accidents that were terminated due to the appearance of clinical symptoms of acute ,
| |
| irradiation have less value for risk analyses than accidents that were terminated by other l means, or never terminated.
| |
| f The current AEOD incident database does not contain the kinds of information needed to perform this analysis. It is recommended that the database either be modified to include this l information, or a separate database be created. There is a great deal of work to be done to refine these preliminary analyses, extend them to additional accidents, and develop the logical framework for extrapolating to other kinds of sources and scenarios.
| |
| d s
| |
| l 2
| |
| F-15
| |
| | |
| Table 3.2.1. Radiation Accidents involving external exposures. See text.
| |
| t T
| |
| S E 2 o $
| |
| - g $3 m I E U *
| |
| : 8. h ! $ $ E i
| |
| # 4 4 e a l t E e e ? 5 E 4 8 $a 5 E H 3 1 2 1 B 8
| |
| ,5 5 B .2 g g g 8 8 % E B E Cide T $ $ k N I 5 5 3 3 5 $ S Reference E
| |
| y ALG78 78 fr-192 IndRad 25 9.2SE+ 11 7 0 0 Med 1 1 1 Jammet et al.1980s,1980b 1.60E-13 AUS70 70 tr-192 IndRad 22 8.14 E+ 11 0 2 0 Rad 1 Brown and McNeill 1971 1.60E-13 BAN 85 85 tr-192 IndRad 50 1.85E+12 0 1 0 Med 1 JaIII and Molla 1989 1.60E-13 Ramalho et al.1988 Tables 1E2 (97 BRA 87 87 Cs-137 Tete 1400 5.18E+13 89 9 Dersons forwhom cytogenetic dosimetry Med 1 1 1 1 1 1 was done) 9.25E-14 (AEA 1988: legnormal fit to 50 individual BRA 87 87 Cs-137 Tele 1400 5.18E+13 doses from Fig. 9 and 199 zeroes (249 240 9 Med 1 1 1 total, p.117) 9.25E-14 BRA 87 87 Cs-137 Tete 1400 5.18E+13 2803 9 Med 1 1 1 1 1 1 Lushbaugh et al. lAEA-CN-51-92 p401 9.2SE-14 IAEA 1988 p.116 External doses for 583 BRA 87 87 Cs-137 Tele 1400 5.18E+13 583 Med 1 1 1 Cleanup Workers CA79 79 tr 192 indRad 28 1.04 E+12 0 11 0 Rad 9.25E-14 1 1 Ross 1980 CZE66 66 l-131 Medical 2.25 8 33E+10 0 0 16 Rad 1.60E-13 1 1 1 Carach et al.1967 7.63E-14 CZE73 73 Co40 Tele 2973 1.10E+14 0 2 0 R.id 1 1 Klener et al.1986 3.70E-13 FRG68 68 Ir-192 IndRad 7.8 2.89E+11 0 1 0 Med 1 1 Chone et al.1970 1.60E-13 FRG72 72 Ir-192 IndRad 29.7297 1.10E+12 0 1 0 Rad 1 UNSCEAR 1988 p. 416 1.60E-13 FRG81 81 Co-60 Tele 2594.59 9.60E+13 0 2 0 Rad 1 1 Stephan et al.1983 IND68 68 Ir-192 IndRad 1.4 5.18E+10 0 1 0 Rad 1 3.70E-13 1 1 Annamalal et al.1978 1.60E-13 ISR90 90 Co-60 Steril 340541 1.26E+16 0 1 0 Med 1 1 IAEA 1993 ITA75 75 Co-60 Steril 36000 1.33E+15 0 1 0 Rad 1 3.70E-13 1
| |
| Parmentler et al.1980 3.70E-13 JCH59 59 Co-60 IndRad 1.75 6.48E+10 1 0 0 Rad 1 1 1 Elliott 1960 JPN71 71 Ir-192 IndRad 5.26 1.95E+11 5 1 0 Rad 1 1 3.70E-13 1
| |
| KOR90
| |
| * Hirashima et al.1980 1.60E-13 90 Ir-192 IndRad 4 1.48E+11 O 19 5 Red 1 1 1 NUREG-1405 KY76 76 Ir-192 IndRad 78 2.89E+12 0 1 0 Rad 1 1 1.60E-13 1
| |
| Jacobson et af.1977 1.60E-13 LA78 78 tr'-192 IndRad 100 3.70E+12 0 1 0 Med 1 Scot; 1980 1.60E-13 F-16
| |
| | |
| Table 3.2.1. (cont.) Radiation Accidents involving external exposures. See text. .
| |
| y :::
| |
| 31 5 e i n h e 9 E v $ E
| |
| - o = g .: e g 2 2 f 4 #
| |
| E 5 t S E i $ E M S S 8 B
| |
| (g .
| |
| .{
| |
| $a t E
| |
| S g
| |
| $ 3 $ $ I E
| |
| 8 8 E
| |
| 8 E
| |
| 8 5
| |
| E C:de s $ $ < k % k W I 3 3 5 $ $ Reference y MEX62 62 Co-60 IndRad 5 1.85E+11 6 0 0 Med 1 1 1 Andrews 1963 3.70E-13 MEX83 83 Co-60 Tele 430 1.59E+13 4000 100 ? Rad 1 1 Listar 1984; IAEA 1989 3.70E-13 MOR84 84 Ir-192 IndRad 16.2162 6.00E+11 28 0 0 Med 1 1 1 Marshall 1984 1.60E-13 NJ74 74 Co-60 Steril 120000 4.44E+15 0 1 0 Rad 1 Barlotta 1980 3.70E-13
| |
| 'NJ77 77 Co-60 Steril 500000 1.85E+16 0 1 ~0 Rad 1 Barlotta 1980 3.70E-13
| |
| .40R82 82 Co-60 Steril 65720 2.43E+15 0 1 0 Med 1 Flatby et al.1983 3.70E-13 NY83 83 Co-60 IndRad 25 9.25E+11 0 10 0 Rad 1 1 1 NRC I&E Notice 83-16 3.70E-13 PA92 92 tr-192 Brachy 3.7 1.37E+11 94 0 0 Red 1 1 1 1 NUREG-1480 1.60E-13 PRC63 -03 Co-60 IndRad 10 3.70E+ 11 7 0 0 Med 1 Gen-yao et al.1980 3.70E-13 PRC80 80 Co-60 Sterli 53000 1.96E+15 0 1 0 Med 1 1 Gen Yao Ya et al.1990 3.70'E-13 PRC85 85 Cs-137 IndRad? 10 3.70E+11 3 0 0 Med 1 1 1 Gen Yao Ya et al.1990 9.25E-14 PRC86 86 Co-60 (1) 6888 2.55E+14 2 0 0 7 1 Gen Yao Ya et al.1990 3.70E-13 PRC87 87 Co-60 Sterli 89000 3.29E+15 0 1 0 7 1 Gen Yao Ya et al.1990 3.70E-13 PRC92 92 Co-60 Expt 12 4.44E+11 18 0 0 Meo 1 1 1 Nenot 1993 3.70E-13
| |
| , SAF77 77 tr-192 IndRad 6.75676 2.50E+11 3 0 0 ?Med 1 1 1 UNSCEAR 1988 p. 416 1.60E-13 SA1.89 89 Co-60 Steril 18000 6.66E+14 0 3 0 ?Med 1 1 Uttlefield et al.1991 3.70E-13 SCO69 69 Ir-192 IndRad 25 9.2SE+11 0 1 0 Rad 1 1 Harrison et al.1973 1.60E-13 TN71 71 Co-60 Expt 7700 2.85E+14 0 1 0 Red 1 Wade 1972, Vodopick and Andrews 1980 3.70E-13 TRK76 76 Co-60 Tele 2260 8.36E+13 0 5 0 Rad 1 1 1 Yalcintas et al.1980 3.70E-13 TX72 72 Cs-137 IndRad 4 1.48E+ 11 '1 0 0 Med 1 1 1 Collins 1980 9.25E-14 UK77 77 Ir-192 IndRad 21.6216 8.00E+11 0 1 0 Rad 1 UNSCEAR 1988 p. 416 1.60E-13 UK81 81 Cs-137 Brachy 0.12 4 14E+09 0 0 1 Rad 1 Heaton and Murray 1982 9.25E-14 WA76 76 Am-241 Ofnse 343.337 1.27E+13 0 1 10 Rad' 1 1 1 McMurray 1983 7.57E-15 Wl81 61 Co.40 Expt 200 7.40E+12 .
| |
| 0 1 0 Rad i 1 1 Rossi et al.1962 3.70E-13 (1)* Radiation Station" I F-17 i
| |
| | |
| ~
| |
| Tabin 3.2.1. (cont.) Radiation Accidents involving external exposures. See text.
| |
| S 5 e
| |
| .s 2 m
| |
| }m ~
| |
| 2
| |
| - 'E
| |
| , & E E y 2 $ g ? -
| |
| g 3 u e - E 1
| |
| & E' o S E .
| |
| O O s 8 $ e i s E l f i c a a !
| |
| C;de Minimum '
| |
| z*
| |
| O.12 4A4E+09 k
| |
| 0 0 0 h f m Reference 0
| |
| Maximum 500000 1.85E+16 4000 100
| |
| * 'IO 583 Average 2.8E+04 1.029E+15 166 4.545 3.7E-13 15 Std Dev 9.1E+04 3.358E+15 726.9 15.24 90.97 1.24E-13 GeoMean 1.9E+02 6.85E+12 Medlan 50 1.85E+12 0 1 0 i Mide 1400 5.18E+13 0 1 0 45 45 44 7E-Number 44 41 31 15 16 5 3 28 8 45 D
| |
| s F-18
| |
| | |
| Table 3.2.1. (cont.) Radiation Accidents involving external exposures. See text.
| |
| TIME AND PROXIMITY FACTORS, F(t,p) Whole Body DOSES (SV) Skin DOSE
| |
| ,E E e E e
| |
| @ n O E - -
| |
| T E E g g g 5 d
| |
| & 5 5 '
| |
| o R R i @ @
| |
| e 5 'E R R R 2 y :
| |
| 6 6 ~ j $ 'E 8 7, e > E E, E y . u. E E 3
| |
| -E 9
| |
| m 3o %
| |
| u E F $ $ ,E E E. [ S $ ,E O 3 5 * , 3 a 2 2 g 3 i a E
| |
| 2 U.
| |
| Code e ?c 5 2 2 O O < m 2 2 O 2 M s O 2 2 ALG78 0.148 88.108 88.176 6.7568 270.27 47.297 4 13.04 13.05 40 7 1 4 675.68 100 AUS70 0.13024 1.6892 1.2285 3.5319 2.083 0.22 0.16 0.46 0.271293 BAN 85 0.296 8.4459 6.7568 10.135 8.2753 2.5 2 3 2.44949 81.081 24 BRA 87 4.7915 0.2024 1.9927 0.0209 1.4609 0.0605 4.7 0.97 9.5 0.1 7 0.29 4.7 4.1741 20 BRA 87 4.7915 0.1119 3.2975 2E-06 1.4609 0.0038 13 0.536 15.8 1.02E-05 7 0.0182 13.49 4.1741 20 BRA 87 4.7915 0.0034 0.6261 1E-09 1.4609 0.0163 3 7.00E-09 7 4.1741 20 BRA 87 4.7915 0.0002 0.0003 SE-06 0.0033 0.0001 2.8 1.07E-03 1.48E-03 2.426E-05 0.016 6.2E-04 2.82 4.1741 20 CA79 0.16576 0.7264 1.5601 0.0121 5.2787 0.1647 5.8 0.*204 0.2586 0.002 0.875 0.0273 5.83 24131 4000 CZE66 0.00635 0.0236 0.0079 0.0394 0.0176 0.00015 0.00005 0.00025 0.000112 CZE73 40.7 0.0184 0.0025 0.0344 0.0092 0.75 0.1 1.4 0.374166 2.457 100 FRG68 0.04618 32.484 21.656 43.313 30.627 1.5 1 2 1.414214 4331.3 200 FRG72 0.176 1.7045 0.3 FRG81 35.52 0.0084 0.0056 0.0113 0.008 0.3 0.2 0.4 0.282843 IND68 0.00829 156.85 .
| |
| 1.3 15685 130 ISR90 4662 0.0032 0.0021 0.0043 15 10 20 ITA75 492.84 0.0284 14 0.0487 24 JOH59 0.02396 1.0435 0.025 375.67 9 JPN71 0.03114 13.167 15.222 3.2114 42.711 8.4138 2.7 0.41 0.474 0.1 1.33 0.262 2.66 KOR90 0.02368 1.2965 3.7162 0.0063 14.717 0.1098 7.9 0.0307 0.088 0.00015 0.3485 0.0026 7.87 KY76 0.46176 2.0573 0.8663 3.2484 1.6775 0.95 0.4 1.5 0.774597 32.484 15 LA78 0.592 0.0845 0.05 l 168.92 100 F-19
| |
| \
| |
| | |
| Table 3.2.1. (ciont.) Radiation Accidents involving external exposures. See text.
| |
| TIME AND PROXIMITY FACTORS, F(t,p) lWhole Body DOSES (Sv) Skin DOSE
| |
| $ 1 E
| |
| O ^
| |
| e y
| |
| $ E M e $- S .$ 6 E
| |
| = -
| |
| e o u -
| |
| e e .n % a
| |
| , e
| |
| ~
| |
| s .E e s E e o r n ~
| |
| u O O a
| |
| * O o
| |
| E" ka *k o
| |
| !E E M
| |
| E Y
| |
| W E A ! b 3 Y e E M a Code O
| |
| ! 3 m 2 5 .E 2
| |
| 2 O O
| |
| : 2. 2
| |
| < "5 m 2 5
| |
| .E E
| |
| E 8 8 %
| |
| a 2 O O s s MEX62 0.06845 446.17 184.51 175.31 686.63 408.04 1.7 30.54 12.63 12 47 27.93 1.7 MEX83 5.8867 0.0041 0.1276 8E-09 1.1891 0.0001 14 0.024 0.751 5.0E-08 7 7.7E-04 13.8 MOR84 0.096 123.02 75.417 10.417 260.42 '92.188 2.6 11.81 7.24 1 25 8.85 2.6 ,
| |
| NJ7/. 1642.8 0.0025 4.1 NJ77 6845 0.0003 2.1 NOR82 899.707 0.0445 40 NY83 0.34225 0.0044 0.0015 0.0058 0.0029 0.0015 0.0005 0.002 0.001 PA92 0.0219 1.3313 1.8718 0.0003 8.6286 0.1329 25 0.02916 0.041 5.6E-06 0.189 0.00291 25.3 I PRC63 0.1369 170.2 227.17 14.609 584.37 75.237 4.1 23.3 31.1 2 80 10.3 4.1 PRC80 725.57 0.0072 0.0072 5.22 5.22 PRC85 0.03423 262.97 29.218 233.75 292.18 261.88 9 1 8 10 8.962809 1.12 PRC86 94.2967 0.0323 0.0276 0.0371 0.032 3.05 2.6 3.5 3.016621 i PRC87 1218.41 0.0011 0.0011 0.0012 0.0012 1.35 1.35 1.46 1.403923 PRC92 0.16428 22.583 33.479 1.887 121.74 11.748 2.9 3.71 5.5 0.31 20 1.93 2.89 i SAF77 0.04 11.925 14.825 2.5 29 6.75 3.6 0.477 0.593 0.1 1.16 0.27 3.63 2500 100 SAL 89 246.42 0.0215 0.0108 0.013 0.0337 0.0198 1.6 5.3 2.67 3.2 8.3 4.89 1.62 SCO69 0.148 4.0541 0.6 1351.4 200 :
| |
| TN71 105.413 0.0247 2.6 i TRK76 30.9394 0.0001 0.0001 2E-06 0.0003 3E-05 10 0.00346 0.00411 0.000047 0.00978 0.000945 10.1 TX72 0.01369 730.46 10 UK77 0.128 0.7813 0.1 I UK81 0.00041 0.487 0.0002 WA76 0.09617 0.052 -
| |
| 0.005 Wl61 2.738 0.9131 2.5 6
| |
| F-20
| |
| ..m___ _ . _ .______...._______.___.__.______._._____.__._m.__._ . _ _ _ . _ . _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ . _ _ _ _ _ _ . _ . _ _ _ . _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ . _ _ _
| |
| | |
| Table 3.2.1. (cont.) Radiation Accidents involving external exposures. See text.
| |
| TIME AND PROXIMITY FACTORS, F(t,p) Whole Bod'r DOSES (Sv) Skin DOSE
| |
| .y -
| |
| E .
| |
| y e
| |
| e
| |
| ^
| |
| E E E c - c Y N* E -
| |
| ea ea &
| |
| ~
| |
| e -
| |
| g g e a e a
| |
| .0 w $a e a
| |
| *@ , .c
| |
| ~ 5 7 o .E
| |
| $ 1 d 8
| |
| : k.
| |
| * n
| |
| ~
| |
| E o o a a k Q
| |
| * E " 3* Y E E c \ % 3 E & B 3 E & "4 a 3 % 3 2 o E E 62 2 " E E E m a 8 8 a
| |
| 5 s 3 IE l E S I $ E 3 Code e < m 2 2 0 0 < m 8 Y M M Minimum 0.00041 0.0001 0.0001 1 E-09 0.0003 3E-05 1.6 2 2 o e s s 0.00015 0.00148 7E-09 0.00025 0.000112 Maximu 6845 730.46 227.17 233.75 686.63 408.04 25 40 1.12 0.0487 "9 31.1 12 80 27.93 25.3 Average 3.8E+02 46 38 82 24131 4000 17 34 6.7 4.62 5.76 1.57 10.21 3.07 6.37 3084.5 317.6 Std D iv 1.2E+03 133 67 53 176 90 6.3 8.47 8.19 3.08 17.94 5.79 6.26 6848.8 984.1 GeoMea 1.3E+00 0.32 .0.43 0.018 1.42 0.19 4.45 91.2 59.7 Median 3.0E-01 0.49 3.51 0.01 3.25 0.1213 4.1 0.95 1.835 0.16 2 0.332083 4 125 62 Mode 4.8E+00 6.76 1.46 2.5 0.1 7 4.2 100 Number 45 45 18 29 29 28 16 45 18 29 29 28 17 16 ~3 i
| |
| t F-21
| |
| | |
| 10 .: : 100 S Frequency Ti 8 - - -
| |
| 80 Cumulative Mi R
| |
| + Frequency I g 6 -
| |
| L./ F tt 60 m ,
| |
| : c. *
| |
| \.
| |
| o .
| |
| 1 g.
| |
| e p$g p -
| |
| g E
| |
| : u. 4 -
| |
| l yo de 7@k < _
| |
| 40 ?.
| |
| y g rs 7 . ,
| |
| 2 -
| |
| 20 lffe ??
| |
| u ,, g y a .
| |
| 0 : - 0 o 0.01 0.02 0.04 0.07 0.1 0.2 0.4 0.7 1 2 Time-and Proximity Factor (h at 1m)
| |
| Figure 3.2.1. Frequency distribution of time-and-proximity factors from the 46 highest dose ~
| |
| cases from the 1987 Goiania accident on a roughly logarithmic horizontal scale.
| |
| (IAEA,1988a; Figure 9) 12 Frequency v 10 -
| |
| { .
| |
| ..e t
| |
| ~
| |
| D 5 k -
| |
| 8 : : i s* s I
| |
| :: 6 - e -
| |
| I E i s$.- 5 i I i it f M i i ! E 4 -
| |
| : R i e s P $ $ f II (p
| |
| I $ T S 3 l N ll h 2 s j p g g gi n e p- ; k j [-
| |
| - E. p p
| |
| ' 4. p e e p 4 t -
| |
| 0 N
| |
| * 0 b O N E $ I I 0.0001 0.001 0.01 0.1 1 to Time-and Proximity Factor (h ct 1m)
| |
| Figure 3.2.2. Frequency Distribution of Time-and-Proximity Factors for the 1992 Indiana, PA mIr therapy misadministration accident on a logarithmic horizontal scale.
| |
| (NRC, NUREG-1480,1993b) -
| |
| )
| |
| 6 F-22
| |
| | |
| 1 5 - : : : : 100
| |
| ../
| |
| l l g--
| |
| ^
| |
| 4 -
| |
| f -
| |
| 80
| |
| , O Frequency
| |
| + umulatNe Frequency h'3 - , - 2 -
| |
| 60 m I
| |
| a l ! ! ! ! $ ;
| |
| 2 f 40 b f f u 7 3 g
| |
| l ,e ! $
| |
| ]ff j
| |
| f iij -
| |
| : f. - -
| |
| 20
| |
| [
| |
| 1 .
| |
| s (>; i & g -
| |
| s
| |
| ;i '
| |
| 0 : -
| |
| =
| |
| 0 i
| |
| . 0.001 0.01 0.1 1 10 100 l Time-and Proximity Factor (h at Im) i 1
| |
| Figure 3.2.3. Frequency and Cumulative Frequency of Time-and-Proximity Factors for the 1990 Korea-USA n2Ir source shipment accident on a logarithmic horizontal scale.
| |
| (NRC, NUREG-1405,1990) l i
| |
| 60 -
| |
| : 100
| |
| , 5 Frequency ( l i 50 -
| |
| -e- Cumulative Frequency j -
| |
| 80'
| |
| : :i
| |
| :l h, 40 -
| |
| 60 ' y Ex o-
| |
| - B e
| |
| 2 -
| |
| 40 E
| |
| *
| |
| * 20 -
| |
| 20.
| |
| 10 .
| |
| 0 0 0.00001 0.0001 0.001 0.01 0.1 1 10 100 1000 Time-and Proximity Factor (h at 1m)
| |
| Figure 3.2.4. Analysis of 40 radiation accidents involving scaled sources, involving 231 .
| |
| individuals for whom doses were available. The Goiania and Juarez accidents are excluded because the large numbers of victims of those accidents would dominate the graphs.
| |
| i a F-23
| |
| | |
| l Table 3.2.2. Summary of Whole Body Time-and-Proximity Factors from a survey of 40 accidents involving sealed sources. The accidents at Goiinia, Brazil, and Juarez, Mexico are j not included. The very large numbers of persons involved would dominate the summary i statistics.
| |
| Quantity Value Units i Number of accidents 40 -
| |
| Number of individual doses 231 -
| |
| Average 30 hours at a meter Standard Deviation 100 hours at a meter i
| |
| Geometric Mean 0.37 hours at a meter l Geometric Standard Deviation 43 -
| |
| Median 0.46 hours at a meter Mode 0.00068 hours at a meter Minimum 0.000112 hours at a meter Maximum 730 hours at a meter l
| |
| Table 3.2.3. Collective Time-and-Proximity Factors for several scaled source accidents.
| |
| Collective Source Source Collective Time-and-External Dose Activity Strength (Sv/h Proximity Factor Population Group (Person-Gy) (Bq) at 1 m) (hours at 1 m)
| |
| All persons in Goiinia 56.3* 5.09E13 4.71- . 11.%
| |
| 4 who died in Goiania 14.9* 5.09E13 4.71 3.17 85 persons 2.5 1.37E11 0.022 113 l Indiana, PA 24 persons, 0.74 9.2E10 0.024 31.2 Korea - USA
| |
| ' Collective doses from 1172, UNSCEAR 1993. These doses are probably too low by a factor of 2, from comparison with Figure 9 of IAEA 1988b (74 person-Gy listed there among 50 most highly exposed).
| |
| F-24 1
| |
| | |
| 1000, ,
| |
| 8 - F = p0.4(rA)472
| |
| ' a 100 r ' 85
| |
| -
| |
| * Acute Irrad. Symptoms T %
| |
| * 's , a Term. by Other Means 10 r
| |
| ~
| |
| Indiana, PA e a' . --- 5 Sv lsodose Une a e .
| |
| a: 1 r a em u', ,
| |
| tO 0.1 r K ''
| |
| a N
| |
| 'g . e *
| |
| ~
| |
| $ m Goiania E 0.01 r u -
| |
| s * .ap g' a
| |
| 0.001 "
| |
| Juarez * .
| |
| 0.0001 ' ' " " ' ' ' ' " " " ' ' ' " " " ' ' ' " " " ' ' ' ' ' '" ' ' ' ""'''"""''''"
| |
| 0.0001 0.001 0.01 0.1 1 10 100 1000 10,000 Source Strength, PA (Sv/h at 1 m)
| |
| Figure 3.2.5. Average whole-body time-and-proximity factors (F,) for 41 sealed source accidents as a function of source strength on a lagarithmic horizontal scale. Most points shown above the 5-Sv (500 rem) isodose line were fatal accidents.
| |
| 3.3 INTERNAL EXPOSURES There are substantial differences between cases involving only external irradiation and cases where internal contamination has resulted. First, it is relatively rare for internal contamination to occur in non-reactor-type accidents. Second, early clinically recognizable symptoms of the internal radiation exposure are rarely seen in internal contamination cases.
| |
| In most instances, any deterministic effects, and certainly the stochastic effects, will display themselves only at a time in the distant future, if at all. As a result, accidents may go unreported. Third, medical treatment procedures can be used to induce decorporation of the internal radionuclide burden, resulting in a reduction in radiation dose.
| |
| 3.3.1 Theoretical Basis for Fraction-Taken-In Radioactive material may gain entry to the body by inhalation, ingestion, or absorption through skin or wounds. The amount of radioactive material, or fraction of the total radioactive material available, that enters the body is a major factor in the radiation dose
| |
| - resulting from the intake. The approach taken in this report is to perform a quantitative, probabilistic characterization of historical radiation accidents in which the fraction-taken-in is known or can be calculated and then to generalize the results to generally licensed radioactive F-25
| |
| | |
| l l
| |
| i l
| |
| 1 devices. 1 It should be noted that since, in many cases, it is generally not known that an accident has occurred, the majority of early biological data needed to calculate radiation doses from the incident will not have been collected. In some cases, the lack (J early biological data has hampered the use of historical records for detennining the fraction-taken-in values.
| |
| For internal contamination cases, Brodsky (1980) has proposed the fraction-taken-in approach t to chancterizing internal contamination cases. Depending on how the data are reported in the literature, the fraction-taken-in may be available directly, or may be calculated from reported doses. In the former case,
| |
| =
| |
| A' (II)
| |
| F,'f _ _ ,
| |
| A where F,,f denotes the fraction-taken-in for the ith person; l
| |
| A fdenotes the activity taken in by the ith person; and A denotes the source activity.
| |
| l If activities taken in are not published, the fraction-taken-in may be calculated from 50-year ,
| |
| committed effective dose equivalent to the ith person using l
| |
| H50''- = ''' 5 rems ,
| |
| (12)
| |
| ALI where ALI is the 1980 stochastic annual limit on intake (EPA 1988); and 5 rems is the 50 year committed effective dose equivalent resulting from an intake of 1 ALI.
| |
| Solving for F,,f, we have Hso,i ALI y,', ,
| |
| (13) \
| |
| 5 rems A where H yo d
| |
| ,j enotes the 50 year committed effective dose equivalent received by individual i; ALI denotes the annual limit on intake; and 5 rems is the committed effecdve dose equivalent associated with an intake of a 1980 ALI.
| |
| If 1991 ALIs (based on 2 rems effective dose) are used, F ,; is i Ei ALI
| |
| = (14)
| |
| F'''. 2 rems A
| |
| 'where E idenotes the effective dose to the ith person from the incident.
| |
| F-26
| |
| | |
| i i
| |
| i
| |
| ! 3.3.2 Results of Analysis of Accidents for Fraction-Taken-In
| |
| ; The activity of the radioactive source at the time of the accident is an essential piece of information. The information needed for this analysis was generally unavailable in the NRC/AEOD incident database. A literature review of available data on accidents involving L radioactive materials was conducted. The results of this review are presented in Table 3.3.1. ,
| |
| l The historical cases reported in this table are only those for which the source activity data 4 were obtainable from the literature sources available, and are the same as those included in
| |
| ! Table 3.2.1. For each incident, the locate, date, nuclide, source activity, incident type, and
| |
| } reference were reported, as well as whether or not external or internal exposure (s) were l involved, the number of deaths resulting from such exposure (s) and the calculated fraction-taken-in (using Eq.11).
| |
| Of the 60 radiation accidents reviewed and reported in Table 3.3.1, only 17 incidents, or..
| |
| (28%), involved cases of internal. contamination. The low percentage of internal
| |
| : contamination cases is most probably due to the fact that many of the reported radiation j accidents involved sealed, radiographic sources that maintained their structural integrity
| |
| , during the accident. As a result, no radioactive material was available for intake into the i body, and the fraction-taken-in is zero. Of the 4364 individuals involved in the accidents
| |
| : listed in Table 3.3.1, 90 individuals (2.1%)
| |
| 1 received internal contamination as a result of
| |
| , the indicated accident. The calculated fraction-taken-in for these individuals ranges from two j in ten-thousand to twenty in a billion (2E-4 to 2E-8). In only one case of an industrial.
| |
| j accident involving an unsealed source of tritium (Lloyd 1986) was the fraction-taken-in j greater than this range (in that case it was 2E-2).
| |
| i I
| |
| i
| |
| ! In most cases, the distribution of the fraction-taken-in, and thus the resultant dose, is
| |
| ! lognormally distributed. This is best illustrated in Fi re 3.3.1 using data from the 1987
| |
| ! Goiania, Brazil accident which involved a 1,375 Ci p7Cs source (Brandao-Mello 1987). The 1
| |
| i log probability analysis of the 20 highest-exposed individuals indicated in Table 3.3.1 L indicates that the distribution of fraction-taken-in values is lognormal with a mean of 2.6E-6 l and a geometric standard deviation (GSD) of 5.4 (Figure 3.3.1). When a total of 77 i individuals who had internal contamination from this accident were considered (IAEA j 1988b), the distribution of fraction-taken-in values was lognormal with a mean of 6.2E-6 and a GSD of 21 (Figure 3.3.2). The committed effective dose equivalents determined for these
| |
| ; 77 individuals are also lognormally distributed with a mean of 0.3 Gy and a GSD of 11.2 4 (Figure 3.3.3). When 194 of the cleanup workers at Goiinia were evaluated, their fraction-i taken-in values were estimated to range from 7E-11 to 2E-15.
| |
| l In the most recent publication of the UNSCEAR report (UNSCEAR 1993), the collective dose for all persons involved in the Goiania incident and for the 4 individuals who died as a l
| |
| ! 1 For the 77 cases of internal contamination resulting in the Goiania accident (Brandao-Mello i 1991b), only 20 highest internal exposures are listed.
| |
| F-27 l
| |
| 1
| |
| | |
| i l result of the incident were given. As shown in Table 3.3.2, this data can then be used 2: a i
| |
| direct application of Eq.13. By using the reported collective internal doses of 3.7 aed 2.3, respectively, the resulting fraction-taken-in values are 9E-6 and SE-6, respectively. The
| |
| } estimated collective doses for this incident are within the range of values calculated for individuals.
| |
| 1 Figure 3.3.4 provides an illustration of the frequency of values for fraction-taken-in for the i
| |
| 60 radiation incidents reviewed. In addition, both the 0.3 maximum value and th.9 realistic range (10E-6 to 10E-5) used in the ORAU report are include on the plot. .It is interesting to j note that the so called ' realistic' range of values used in the ORAU report (obtained from published literature on transportation accidents) falls in the range of values observed from
| |
| { actual accidents. However, the maximum value of 0.3 used by ORAU falls above the
| |
| ! observed range, Thus, the ORAU approach to calculation the doses from their selected scenarios might be applicable today, if two things are done: 1) the more realistic values for i
| |
| fraction-taken-in should be used, and 2) CEDE and TEDE should be calculated instead of
| |
| ; just organ doses.
| |
| 1
| |
| ; 3.3.3 Conclusions and Generalizations of Results to Imoroner Transfer and Disposal Scenarios for Generally Licensed Devices
| |
| ' When considering all reported non-reactor-type radiation accidents listed in Table 3.3.1, the fraction-taken-in was found to range from 2E-4 to 2E-8. The internal contamination cases upon which this range of values is based resulted largely from either industrial accidents i involving unsealed sources of radioactive material or intentional destruction of licensed I j radiation sources.
| |
| I i
| |
| When generalizing the results of this historical review, it is difficult to support the l " arbitrarily chosen" value of 0.3 for the fraction-taken-in for generally licensed devices used 4 in the ORAU Report. Thus, the most defensible values to use for the fraction-taken-in are j distributions in the range of 2E-4 to 2E-8, except for cases involving sealed tritium sources
| |
| ! in generally licensed devices.
| |
| i
| |
| -i 1
| |
| 1 e
| |
| 4 i
| |
| 4 F-28 i
| |
| ,- ,, - - , , -- m
| |
| | |
| TABLE 3.3.1. Summary of Radiation Accidents Used to Determine Fraction-Taken-In Values
| |
| * lo O '
| |
| l1*Ee . n e
| |
| is a y z SOURCE e,
| |
| b
| |
| ! 3 d
| |
| g g
| |
| a iE 6
| |
| 8 O
| |
| CODE UK44 LOCATP)N UK DATE HUCLf0E ACTfvfTY 1944 Re-226 0 005 CI INCIDENT TYPE Pr45- vi B
| |
| X s ;
| |
| 2 e
| |
| 0
| |
| ! .. . ..:: ~
| |
| 0 SmRh 1962 JOH59 JOH 1969 Co 80 2 Cl trusustrial Aeeksort X -
| |
| 1 0 0 wig 1 Emaet 1960 W1 1961 Co80 200 Cl Egerime1af Aceksort X - 1 0 0 Roost 1962 MEX62 Menks 1962 Co80 5 Cl Incbetrial Amh$ert X - S 4 0 PA63 Pha. PA 1963 S-35 A.- u 1963 1.27 CI E-+ _ _- Vint X X 1 0 1.0E45 Meees 1953 PRCS3 PRofCNne 1983 Co80 10 Cl test eeurtee X - 6 2 O Mettler 1990 USAS4 U.SA 1964 Ar%241 0 05 Cl test eewee X X 2e 0 1.6E46 Cohen 1979 USAS4 USA 1984 ArN241 0.05 CI Leet source X X 2b 0 7.2E47 Cohen 1979 FRO 64 CA, 1964 H-3 ? Cl Trtisted Petre u ; - X 4 1 2 " n; gger CZE86 Crocemolowskie 1988 1-131 2 25 C1 --IrAtereft eecWent X X 16 0 0 Carsch 1E61 NY6' Urtv of Raaer 1967 tr-192
| |
| * 2 Cl Het ese leak X X 2a 0 1,4E44 Coat 1982 NYS7 UnP,of RA;er 1967 tr 192 2 Cl Het est best X X 2b 0 2.1t 44 Cool 1962 USA 67 USA 1957 Ir-192 70 Cl Pulustrief AceWort X - 2 0 0 Shusen 1973; Manflekf 1969 FRG68 C- ;a.; + 1900 tr-192 8 Cl IndustMe! AccWert X - 1 0 0 Chene 1970 fND68 India 1989 fr-192 1 Cl Industrial Aechtert X - 1 0 0 A e-;.Isl1973 ARoe8 A. y .;;, . 1989 Co-137 14 Cl Industrial Acehsert X - 1 0 0 8ertnaen t969 FRG68 Cw r 1989 Ir 192 ? Cl Industdel Accidert X - 1 0 0 SeneWor 1969 gNoe9 inces 1969 fr-192 Cl
| |
| % Industrial AW X - 1 0 0 Scoop Annammtal1978 SCO 1989 Ir-192 25 Cl M AccWort X - 1 0 0 HerMean t973 UKs0 UK 1989 fr-192 - 24 Cl industrial Actddert X - 1 0 0 HerMmen 1973 AU370 AUS 1970 Ir-192 22 Cl trufustrief Aecksort X - 2 0 0 UK70 UK 1970 Brown and McNeit 1971 Ir 192 22 CI trufustrial Accidert X - 1 0 0 Pumst 1972 JPN71 CNt:e, Japart 1971 fr-192 S C1 industrial Accidert X - 8 0 0 Kurtstart 1977; MerJer 1990 TN71 Oet Rkfge, TN 1971 Coe0 7,700 Cl EW.,a AceWort X 0
| |
| - 1 0 Velopick 1974; 7,_:. 1974; Fry 1990 UK71 UK 1971 fr 192 S Cl Industrial Acekfert X - 1 0 0 Purnit 1973 8UL72 Bulgerte 1972 Co 137 6.3 Cl SuicWe X - 1 1 0 09eerta1967 TX72 Temme 1972 Co-137 4 Cl trtettional W X - 1 0 0 Ce;line 1960 FRG72 Germore, 1972 tr-192 30 Cl Industrini Accident X - 1 0 0 Seget1975 CZE73 Cee levsk$a 1973 Co40 2.973 CI Med. T% y, Aeddert X - 2 0 0 10ener 1986 ME74 "- ;J Eser 1974 Ir 192 ? Cl Industrial AceWort X - 1 0 0 Purret 1976 NV74 EPA-Lee Vogen 1974 H9-203 0.03 Cl Acclef. fnad. seletsization ? X 2a 0 3.0E46 Brown 1975 NV74 EPA-Lee Vegas 1974 j's203 0.03 CI Aceks. fned. setoglanden ? X 2b 0 2.5E45 Brown 1975 l F-29
| |
| | |
| ~
| |
| i t
| |
| TABLE 3.3.1. (cont.)
| |
| 5 a z
| |
| i o
| |
| s = g a
| |
| CODE N.f74 LOCATION Peregeny, NJ - 1974 Co 40
| |
| ==ca DATE NUCUDE ACTMTY 13,000 l
| |
| Cl INCIDENT TYPE MeenalStat. AmeMut l lv l X -
| |
| 9
| |
| =
| |
| 0 3
| |
| w REFERENCE 1 0 Siehesy 1979; Fry 1900 IRQ75 traq 1975 k-182 82 Cl IndumMul AseMart X -
| |
| 1 0 0 usyd1979 ITA75 emir 1975 C*eo 3s 000 Cl Inessarial Accadert X -
| |
| 2 1 0 {
| |
| Jernmut tes0; Pennerater tee 0; oeverte tee 7 UM76 USA 1978 Ir-102 95 CI trutustrial Aostemet X -
| |
| 1 0 0 USA 7e USA 1978 Co40 105 NUREG62219771NUREGL400441977 Cl inessertal AmeWort X -
| |
| 0 1 0 NUREG4221977iNUREC400441977 WA79 Hentwd 1978 h241 340 Cl 8 .C ^
| |
| X X 1 0 3.2E48 Held 1979; Meekspror 1983; Fry 1000
| |
| * KY76 leereustry 1979 Ir-192 78 Cl InesstrialAmeMont X -
| |
| 0 ~
| |
| 1 0 Jacobsen 1977 '
| |
| TRK7e Turher 1978 C+40 2.200 Cl Med.Therepr Aceldert X 5 0 0 Yetoirene 1000 UK77 UK 1977 tr-182 22 O Industrial AmeMont X - 1 0 0 usyd 1979 sA77 soumA9km 1977 Ir-te2 7 Cl IndusertalAschtere X -
| |
| 0 1 0 usyd 197s; sessen tee 0 NJ77 T- 1 NJ 1977 C+40 000,000 Cl MedoelStart. Accedere X -
| |
| 0
| |
| . 1 0 sestsey 1973; Fry tee 0 ALG70 W 1978 W-182 25 Cl I.est heusertaleeuree X ? 22 5 0 Morear 1980 LAyg taugsgge - 978 Ir132 100 Cl InductitelseeMart X - 1 0 0 Best 1900 CA79 Caelernte 1079 Ir-tet 28 Cl IndusMut essedent X -
| |
| 11 0 0 Rene1g30 (A7s tendelene 1979 tr-tet ? Cl IndusMuleneident X - 3 0 0 i
| |
| Seatt000 PRC40 PRetChha 1980 Ce40 53,000 CI IndusMalesehtent X -
| |
| 0 1 0 Ye 1980 MAgo SU Med Center 1980 To.Seen 0.04 C1 Seged W to C- - - 1 0 0 EwouteneR 1000 UOG1 UK 1981 Co.137 0.12 Cl GreelgWierapy X 0
| |
| - 1 0 Hesten and "_n,1982 NORS2 L --, 1982 C640 M.000 Cl hesuMui Use X - 1 1 0 Regen 19e0 MET 83 Juares, Muene 1983 Co40 m Cl LastineauMuiseuree X -
| |
| 4000 5 0 Unter 1900; IAEA 190ste MOR44 Moroese 1984 tr-132 10 CI Lost trutusMut sowoe X X 28 8 ? tesleer 1000; t" O tee 0 Swee5 Suenertung 1986 H3 30 Cl enesuMaloesteert - X 5e 0 2 0EM usyd iges awsag summenand 1985 S3 90 Cl IndumMuleseldet X Sb 0 S.4E44 Usyd1988 3Weg5 suemartend 1985 S3_ , . 80 C1 IndustreuleseMurt - X Sc 0 7.0E4 Usyd iges SWse5 Sutesnand 1985 H3 80 Cl Industrialeseldad - X 5d 0 7.0E6 usys tees SWse5 Seenestand 1985 m3 30 Cl hduettelsesidert - X 5e 0 7.0E 05 Usydites BANOS SAN 1985 tr-te2 80 Cl heusMuleseldent X - 1 0 0 Jos and Moss tese PRCOS PRetCNne 1985 Co.137 10 Cl Lost inesmertal Ssence X - 3 0 1 Yetes0 PRCes PR et CNne 1988 Co40 0.008 Cl heseMelaccMurt X - 2 0 0 Ye 1980 F-30
| |
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| 3 9
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| b b h h b h b h b b h b b b b h b h b b 9 1 1 t t 1 t 1 t 1 1 t 1 1 1 1 t 1 t 1 1 1 0 3 9 9 e e 9 e 0 e 9 e 9 0 9 9 e e
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| 9 e e
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| 9 9 A 9 9 9 9 e e 9 1
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| 1 o m s o o o o o o e e-o e o o o c- o > o 9 R s o 0 G G s e m s o s e e E d d F n n e m m m m s e e m d d d d d d d d dn dn d n n n a n n n n n n a d
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| B B t B B B t a B 5 8 8 8 8 8 8 8 7 7 7 7 7 7 T 8 0 -
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| 0- m Em E4 4 4 4 4 4 0- 0 4 0-0- 0 0 4 0 E E E E E
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| 0 0 0 0 0 0 2 0 0 0 0 0 0 2
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| 0 20 1 3 1 2 9 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2
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| . X X X X X X XX X X X X X X X X X X X X - - - - -
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| ) o%i- .
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| t n O$i- &t l a X X X X X X X X X X X X X X X X X X XX X X X X X o
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| C 1 1 e e e e e e e e e e e e e c c 3 3 e e e e e e e e e c c c r m o e o c c c c r u
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| r u d -
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| o e c c c c c o F 3 Nw T o e o o w w w w w w w o o o o w 3u u o
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| w u o o e o o S S S t
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| S w w a S S r r u u ru u e e o o o d r S S S r t n e o r s e r d y E N S S S S S S S S S ? ic e n ae c i
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| c p
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| l l e e a a l l l a
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| l e le le e l l a
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| l l n o e mr e are c o ic e e do do c e n c c e D e o e c c ls do do e e e e e s
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| B se do
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| d o e t t e h C e e o e e e A Ms ks s t t r y N M M M M M t M M M M M M M M MM M M M e e t
| |
| i s hc T I t t t s s s s s s s s t t t t t s s t t t s s s s t t t t s s s s t t t a s du hs tu t u d a a La La La La La La L a L a o a a a a L L L L L I e a a a l L l
| |
| 'a L L a n s S I I n Br L
| |
| l l l l l l i l l l l l C C C C lC C C C C C C C C C C C C C C C C C C C C l l l l l l I l l l l l h5 EV 0 0 1 CT 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 7 7 7 7 7 7 7 7 7 7 7 7 7 0, 0, 5 5 5 0 0 4 5, 4 RM UT 7 7 7 7 7 7 7 3, 3, 3, 3, 3, 3, 3, 3, 3, 3, 3, 3, 3, 3, 3, 3, 3, 3, 1 1 1 1 1 1 3, 3, 1 1 0 8 0 4
| |
| 4 3
| |
| OC SA 1 1 1 1 1 1 1 1 1 1 1 1 1 3
| |
| 7 7 7 7 7 7 7 7 7 7_7 7 7 7 7 7 7 7 7 7 0 0 0 2 2 M 3 3 3 3 3 3 3 3 3 3 3 1
| |
| 3 1
| |
| 3 1
| |
| 3 1
| |
| 3 1
| |
| 3 1
| |
| 3 1
| |
| 3 1
| |
| 3 1
| |
| 3 1 8- 8 - 4 91 9 L 1 1 1 1 1 1 1 1 1 1 o o o 1 o- o o o o o o o o o o o o o o o o o o o C C C r r M
| |
| t C C C C C C C C C C C C C C C C C C C C I I N
| |
| E 7 7 7 7 7 7 7 7 7 7 7 7 7 7 7 7 9 9 0 0 2 7 7 7 7 8 8 8 8 8 8 8 8 8 8 8 8 9 9 9 T 8 8 8 8 8 8 8 8 8 8 9 9 9 9 9 9 A 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 D 1 1 1 1 1 1 1 1 ss ss sa sa g g s s u sa 's sa g s g d a
| |
| sa sa sa sa a n a a z n s a i u a e e e o ra a e e re e e a r N a r a
| |
| r e
| |
| r e
| |
| r a
| |
| r r e
| |
| r e
| |
| r r r r r r r r r r r n e B, B B, B B, B, N d O B, B B, B, B, B t, B, B, B, B, ia, a a s B, B, B, e a a a C e n is I
| |
| T a a, s a ia e s ia s a s ,
| |
| A in in n i
| |
| in n i
| |
| n n n i i n in in n n in i i n in in in n in fo he l a a _
| |
| n le a n le in la le lo a in n a e e n e C a a io i l o
| |
| n e lo la o io a lo io a o o a o io lo la lo io R S e e Ko F r
| |
| r A
| |
| O o G l
| |
| L G G G G G G G G G G G G G G G G G G G P E l 7 7 7 7 7 7 7 7 7 7 7 7 7 7 9 0 0 2 E 78 7 7 7 7 7 7 8 8 8 8 8 8 9 8 8 8 8 8 8 8 8 8 8 8 9 9 A A A A A A A A A A C LA P R A 8 8 D A 8A 8A A A A A A A A R R R R R R R R R R S 5 O P O R R R R R R R R R R R B B B B B B B B B P 1 K
| |
| C B B B B B B B B B B B
| |
| | |
| \
| |
| TABLE 3.3.2 Calculation of Fraction-Taken-In Using Collective Internal Doses From the G ..
| |
| l t
| |
| f Collective Source ALI ;
| |
| Population Internal Dose Activity hr Cs-137 !
| |
| Group (Person-Gy) (Bq) Collective Fraction-(Bq) Taken-In l All persons in 3.7 ;
| |
| 5.09E13 6E6 ;
| |
| Goiania 8.7E-6 "
| |
| 4 who died 2.3 5.09E13 i
| |
| 6E6 5.4E-6 '
| |
| I i
| |
| I i
| |
| f
| |
| [
| |
| t I
| |
| i
| |
| [
| |
| i F-32 l
| |
| I I
| |
| ~ . .
| |
| L m.____ _ _ _ _ . _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ . _ _ ______.__._____.______m___._._m__.___. -
| |
| _ _ - .__. _ . _ _ _ _ __ _ _ _ _ . . _ _ _m.__
| |
| | |
| 10'4 .
| |
| m Data Fin.ney c 10-5 c ---- Uniform a
| |
| ~~
| |
| 10~6 a
| |
| 9 -m<
| |
| E 10-7 iii i i i i i i i i i iii 10-8 3 5 10 20 30 40 50 60 70 80 90 95 97 Cumulative (%)
| |
| FIGURE 3.3.1. Log Probability Analysis of Fraction-Taken-In for 20 Highest-Exposed Individuals with Internal Contamination from the 1987 Goiinia '"Cs Accident. Regression lines are uniform- or Finney-weighted.
| |
| F-33 i
| |
| | |
| .._-_1 l
| |
| 4 m 10 r 5
| |
| E r
| |
| ,c -
| |
| ~ :
| |
| O I x 10-6 r r
| |
| l t
| |
| c o 5 a Data E 10~8 r i
| |
| 2 5 Finney 12- p :
| |
| Uniform 1 0-10 h l
| |
| ~
| |
| I I I I I 'l i I I i 0.6 2 5 10 20 50 80 90 95 98 91.4 C,umulative Probability (%)
| |
| l FIGURE 3.3.2. Log Probability Analysis of Fraction-Taken-In for 77 Individuals with Internal Contamination from the 1987 Goiania "'Cs Accident. Regression lines are uniform- or Finney-weighted. (IAEA,1988a, from Figure 13)
| |
| ! t F-34 v
| |
| 4
| |
| [
| |
| | |
| 101 -
| |
| 1 r .- s S i
| |
| ~
| |
| m 0.1 e a' .-
| |
| * m :
| |
| o -
| |
| 0 0.01 e o E a Data m -
| |
| : g. 0.001 [: Finney i E .
| |
| $ 0.0001 ---- Uniform P
| |
| O.00001 O.6 2 5 10 20 50 80 90' 95 98 99.4 Cumulative Probability (%)
| |
| FIGURE 3.3.3. Log Probability Analysis of Committed Dose for 77 Individuals with Internal Contamination from the 1987 Goiania D'Cs Accident. Regression lines are uniform- or Finney-weighted. (IAEA,1988a, from Figure 14) ,
| |
| F-35 l
| |
| | |
| 160 3,ggg3g,337,qqqggggn;;3 1
| |
| 140 v -v y BRA 87w H l BRA 871 l
| |
| " *
| |
| * V 120 l WA76 l i 100 Y V y h
| |
| c l NV74 l
| |
| . , mesoamcAu
| |
| $ 80 -
| |
| y y i..M... .... ...!.D....E.. D. ..,.I.y it i USAS4 l 60 _ ; i RAqg; _
| |
| y y iloW!!
| |
| l NY67 l i 40 - y
| |
| -------------Y PA63 l l ,
| |
| 20 - - -
| |
| +
| |
| 0-1 E-15 1E-13 1E-11 1E-09 1E-07 1 E-05 1 E-03 1 E-01 Fraction-Taken-in :
| |
| FfGURE 3.3.4. Frequency of Fraction-Taken-In For 60 Radiation Accidents F-36
| |
| \
| |
| i
| |
| | |
| i i !
| |
| | |
| ==4.0 CONCLUSION==
| |
| S i
| |
| ! Work should be done in concert with NMSS sta.f to develop a risk framework that meets the 1
| |
| NRC's regulatory decision making needs. Such a framework is proposed in this report, but j should be fine-tuned to meet the staff's requirements.
| |
| i Such a framework should include the probabilities of accidents occurring. The probabilities j
| |
| ; should be expressed on a per-soume, per-year basis, and include summation over accident types and multiplication by the number of soumes in use. Potential harm from accidents should be ammwi using distributions of coefficients, not point estimates, derived from !
| |
| j historical accidents, including fractions of soume activity taken in, and probable external l ql doses (both local and whole-body) based on analyses such as those presented above.
| |
| l Time-and-proximity factors for 231 individuals involved in 40 accidents reviewed by PNL l (excluding the large population accidents in Goiania Brazil and Juarez Mexico) averaged 30 ,
| |
| hours at a meter, with a standard deviation of 100 hours at a meter, a geometric mean of !
| |
| l 0.37 hours at a meter with a geometric standard deviation of 43. Collective time-and-proximity factors were 12,113, and 31 for the Goiania, Indiana PA, and Korea-USA .
| |
| accidents, respectively.
| |
| Fractions-taken-in were found to t's in the range of 2 x 10-8 to 2 x 104for many accidents reviewed by PNL, many involving miscaled sources. Cleanup workers in Goiania had ,;
| |
| fractions-taken-in in the range of 2 x 10-15 to 7 x 10-11, while to le, including those directly involved in the incident, had fractions-taken-in ave 2.ging 6 x 10 . Roughly 98%
| |
| of individuals in the 60 accidents surveyed had fractions-taken-in of zero.
| |
| 2 i
| |
| i F-37
| |
| | |
| i i l i
| |
| : i
| |
| : i
| |
| | |
| ==5.0 REFERENCES==
| |
| i l' 10 CFR 20. 1991. U. S. Nuclear Regulatory Commission (USNRC), " Standards for Protection Against Radiation." U.S. Code of Federal Regulations. 56 FR 23360 23474.
| |
| l i
| |
| Amended and corrected 56 FR 61352 (3 December 1991) and 57 FR 57877 (8 December i 1992).
| |
| 4 i j 10 CFR 30. 1993. U.S. Nuclear Regulatory Commission (USNRC), " Rules of General
| |
| ' Applicability to Domestic Licensing of Byproduct Material." U.S. Code of Federal Regulations, February 26,1993.
| |
| i i i 10 CFR 31. 1992. U.S. Nuclear Regulatory Commission (USNRC), " General Domestic j Licenses for Byprtxluct Material." U.S. Code of Federal Regulations, November 30,1992. !
| |
| ^
| |
| l 10 CFR 32. 1992. U.S. Nuclear Regulatory Commission (USNRC), " Specific Domestic i Licenses to Manufacture or Transfer Certain Items Containing Byproduct Material." U.S.
| |
| ; Code of Federal Regulations, November 30,1992.
| |
| ' 10 CFR 835, 1993. U.S. Department of Energy (DOE), " Occupational Radiation Protection." U.S. Code of Federal Regulations, December 14, 1993.
| |
| : 40 CFR .190. U.S. Environmental Protection Agency (EPA). U.S. Code of Federal ;
| |
| j Regulations. '
| |
| 4 Accidental Nuclear Excursion. Recuplex Operation 234-5 Facility. Final Medical Report, i
| |
| : From the Health Operation, General Electric, Hanford Atomic Products Operation, Richland, l
| |
| i Washington, April 7,1%2. "
| |
| . Allen, W. R. 1966. " Radiation Injury to the Hand." J. Kansas Med. Soc. 67:447-453, i
| |
| l Alves, R. N. 1990. " Contamination control in the 137 Cs accident in Goiania. In: Ibn Mair.: m.i. for RMiatian Acd'- r Prena-+i= ;; II - CI:=:c=I kr: *=e mM Follow-un j- Sine
| |
| * 1979." Prne-liaan of the heaarl Intarantiaa=I REAC/TS 6.i----e on the MMica! i b i. for Radiarian Accidaar Fin-=i =;. Eds. R. C. Ricks and S. A. Fry. Elsevier l Science Publishing Co., New York, New York.
| |
| 4 Andrews, G. A., B. W. Sitterson, A. L. Kretchmar, and M. Brucer. 1959. " Accidental Radiation Excursion at the Oak Ridge Y-12 Plant IV. Preliminary Report on Clinical and j I2boratory Effects in the Irradiated Employees." Health Phys. 2:134-138,
| |
| ( i
| |
| )
| |
| Andrews, G. A. 1963. " Mexican # oCRadiation Accident." Isotopes Radiat. Technol.
| |
| 1:2:200-201. .
| |
| Andrews, G. A., K. F. HQbner, and S. A. Fry. 1980. " Report o/ 21-year Medical F-38
| |
| | |
| 2 :
| |
| ?
| |
| l 1
| |
| i 1
| |
| ; Follow-up of Survivors of the Oak Ridge Y-12 Accident," pp. 59-79. In: The Medical l' Basis for Radiation Accident Prea= redness. ProceMinas of the REAC/TS International i
| |
| Conference: The Madic=I Basis for Radiation A cident Pre; Mness. Eds. K. F. Hilbner and S. A. Fry. Elsevier, New York.
| |
| j Annamalai, M, P. S. Iyer, and T. M. R. Panicker. 1978. " Radiation Injury from Acute l Exposure to an Iridium-192 Source: Case History." Health Phvs. 35:387-389.
| |
| i Auxier, J. A. 1%5. " Nuclear Accident at Wood River Junction." Nucl. Saf. 6:3:298-300.
| |
| i I Barlotta, F. M. 1980. "The New Jersey Radiation Accidents of 1974 and 1977" 1980. In:
| |
| ;[
| |
| ' The MMical R==3= for Radiarian Accidaat PraanrMaaee. PraceM*= of the REAC/TS far,,natinant Conferanca: The Madical Basis for Radiatian AccWnt Pikarets . Eds. K.
| |
| j F Hahner and S. A. Fry. Elsevier, New York: pp. 151-160.
| |
| ^
| |
| Baron, J. M., S. Yachnin, R. Polcyn, et al. 1%9. " Accidental Radiogold (198Au) Liver l Scan Overdose with Fatal Outcome," pp. 399-414. In: Proceedians of a Symoosium on the l j Handline of Radiation Accidents, International Atomic Energy Agency /World Health
| |
| ; Organization, Vienna, Austria, May 19-23, 1 % 9.
| |
| ! Basson J. K., A. P. Hanekom, F. C. Coetzee. 1980. " Health physics evaluation of an acute i 4
| |
| over-exposure to a radiography source." In: Radiarian Protaccian: A Sveta' antic Anaranch to Safety. Pr = t= of the Fifth Canoress of the f= -d.ral Radiarian Pratacrian jl l Association. Jerunnlem,1980. Pergamon Press: 64-68. l Beck, H. L. 1982. "Special Composition of the Gamma-ray Fraa=n e Rate Due to Noble i
| |
| : Gases Released During a Reactor Accident." Health Phys. 43(3):335-343.
| |
| i
| |
| } Becker, J. and T. Rosen. 1987. " Acute Radiodermatitis from Occupational Exposure to
| |
| ; 1921r." haith. Med. J. 82:12:1561-1563.
| |
| I
| |
| ; Benioson, D., A. Placer, and Vander Elst E. 1969. "Estudio de un caso de irradiaci6n humaan accidental," pp. 415-429. In: Proceedings of a Symposium on the Handling of j Radiation Accidents. International Atomic Energy Agetry/World Health Organization, Vienna, Austria,19-23 May 1969.
| |
| Berteig, L. and J. Flatby. 1984. "The Radiation Accident at Institute for Energy Technology, September 1982. Some Technical' Considerations." J. Indust. Irradiat.
| |
| Technol. 2:3 and 4:309-319,1984.
| |
| Bertelli, L., J. L. Lipsztein, C. A. N. Oliveira, and D. R. Mello. 1990. " Internal 137CS :
| |
| Contamination in the Goiania, Brazil Accident," in: The M: dical Basis for Radiation Accidaat Preamrada-en II - Clinical Fraarience and Follow-up Since 1979. Pracaadiaen of the Second International REAC/TS Conference on the Medical Basis for Radiation Accident F-39
| |
| | |
| t i t Preparedness. Eds. R. C. Ricks and S. A. Fry. Elsevier Science Publishing Co., New York: pp. 342-252. l 1
| |
| Bhushan, V. 1974. "Large Radiation Exposure." !a. Prv4cdines of the Third International
| |
| .l Congress of the International Radiation Protection Association. CONF-730907: Washmgton- 1 769-772. j Brahams, D. 1988. " Radiotherapy Overdose." Lancet.
| |
| Brantian-Mello, C. E. 1991a. " Personal Insights into the. Goiinia Radiation Accident."
| |
| Health Phys. 60:3-4.
| |
| Brandao-Mella, C. E., A. R. Oliveira, N. J. Valverde, R. Farina, and J. R. Cordeim.
| |
| 1991b. " Clinical and Hematological Aspects of Cs-137: The Goiinia Radiation Accident." l Health Phys. 60:31-39.
| |
| " Brazilian Radiological Accident Traced to Abandoned Radiotherapy Device." 1987. !
| |
| Nuclear Waste News 3C5-304, October 22,1987.
| |
| Breitenstein, B. D. Jr... and H. E. Palmer. 1989. " Lifetime Fehw-up of the 1976 !
| |
| Americium Accident Victim." Radiat. Prot. Dosim. 26:317-322.
| |
| Breitenstein, B. D. Jr, S. A. Fry, and C. C. Imshbaugh. 1990. "DTPA Therapy: The U.S. Experience 1958-1987," pp. 397-403. In: The Medical Basis for Radiation Accident ne=*= II - Clinic =1 Runerience mad Follow-co Siace 1979. Praeaadi=== of the Leond International REAC/TS Conference on the Medical Basis for Radiation Accident Preparedness. Eds. R. C. Ricks and S. A. F.y. Elsevier Science Publishing Co., New York.
| |
| Brodsky, A. 1980. Re a aa== ion Factors and Probabilities of Intaire of Material in Process (or "Is 10 4a Magic Number in Health Physics?"). Hea tb Phys. 39(6):992-1000.
| |
| l Brown, J. K. and J. R. McNeill. 1971. " Biological Dosimetry in an Industrial Radiography j Accident." Health Phys. 21:519-522. :
| |
| I Brown, K. W., J. C. McFarlane, and D. E. Bernhardt. 1975. " Accidental Inhalation of Mercury-203." Health Phys. 28:1-4.
| |
| Burson, Z. and C. C. Imshbaugh. 1990. "The 1983-1984 Ciudad Juarez, Mexico, "Co l Accident," 13-23. In: The Medical Basis fcr Radiation Accident Preparedness II - Clinical i Runeriaare mad Follow-un Since 1979. Prordiaen of the Leond I ;e : ont REAC/TS Conference on the Medical Basis for Radiation Accident Preparedness. Eds. R. C. Ricks and S. A. Fry. Elsevier Science Publishing Co., New York.
| |
| F-40
| |
| | |
| u 4
| |
| i I i
| |
| l
| |
| : i Carach, J., S. Csipka, M. Petrasove, T. Tronovec, and F. Minarik. 1%8.
| |
| i j
| |
| ! " Recovery Operations Followi:g An Aircraft Accident Involving Radioactive Cargo."
| |
| Health Phys. 15:280-282. )
| |
| l l l
| |
| 3 Catlin, R. J. and V. P. Bond. 1990. " Assessing the Risk to the General Population in i large Scale Radiation Accidents:
| |
| j A Review," pp. 291-315. In: The Medical h=is for j Radiation Picwi=; of Accident the %.- :Preamrednese II - Clinical Exnerirr-e and Follow-un Since 1979.;
| |
| j -
| |
| Ice. s: anal REAC/TS Ccia a-+ on the M-Micm3 hele for Radiarian Ace''== Ps -mar- =. Eds.~ R. C. Ricks and S. A. Fry. Elsevier Science
| |
| . Publishing Co., New York. j 1
| |
| i Catsaros, N., and A. Vassiliou. 1987. "An A==*== ment of the Average Shielding Factor for the Population of the Attica Basin Using the Shield-f Code." Radiat.' Prot. Danim.
| |
| 21:97-102.
| |
| Chone, B., G. Schneider, and D. Fehrentz. 1970. "lMiridium-Kontaktbestrahlung und Folgeerscheinungen, English Summary." Strahlentheranie 140 1:113122.
| |
| Cohen, N., T. L. Sasso, and M. E. Wrenn. 1979. " Metabolism of Americium 241 in Man; An Unusual Case of Internal Contamination of a Child and His Father." Science 206.
| |
| Collins, V. P., and M.E. Gaulden. 1980. "A Case of Child Abuse by Radiation Exposure."
| |
| In: The Medical Basis for Radiation Accident Preparedness. pp. 197-203. Eds. K.F.
| |
| Hubner, S.A. Fry, Elsevier Science Publishing Co., New York.
| |
| Collins, D. L. 1992. " Behavioral Differences of Irradiated Persons Associated with the Kyshtym, Chelyabinsk, and Chernobyl Nuclear Accidents " Military Med. 157:548-552.
| |
| Collins, V. P. and M. E. Gaulden. 1980. "A Case of Child Abuse by Radiation Exposure,"
| |
| pp.197-203. In: The Madical hei= for Radiarian Accidant Pren=ral.===. Prcwira of the REAC/TS Internarianal Conference: The Medical haie for Radi=*ian_ Au!.+e Preparedness. Eds. K. F. Habner and S. A. Fry. Elsevier Science Publishirig Co., New York.
| |
| ' Cool, D. A. 1982. " Committed Dose Equivalents from an Accidental Inhalation of Insoluble Ir-192." Health Phys. 42.
| |
| Cronkite, E. P. 1990. " Clinical Aspects of Accidents Resulting in Acute Total Body Irradiation (Continued)," pp. 75-78. In: The Medical baie for Radiarian Accident Prenard.2= II - Cliniemi Furgg,ce and Follow-un Since 1979. Pic-Winom of the Second
| |
| 'I=;-.._:: anal REAC/TS Conference on the Medies! R==3* for Radiatian Accident Preparedness. Eds. R. C. Ricks and S. A. Fry. Elsevier Science Publishing Co., New York.
| |
| F-41
| |
| | |
| i l
| |
| L l I
| |
| E i da Silva, C. J., J. U. Delgado, M. T. B. Luiz, p. G. Cunha, and P. D. de Barros. 1991. i
| |
| " Considerations Related to the Decontamination of Houses in Goiania: Limitations and
| |
| ; Impacations." Health Phys. 60:8/-90.
| |
| de Oliveira, A. R. 1987. Un r6pertoire des accidents radiologiques 1945-1985.
| |
| j Radioprotection 22(2):89-135.
| |
| Desrosiers, M. F. 1991. "In vivo Assessment of Radiation Exposure." Health i Phys 61:859-861. i
| |
| ! I l Dodd, B. and L. L. Hemphries. 1988. " Hazards Amnessment of Worst Case Transportation j Accidents Having Typical Radioactive Material Shiamante. f%alth Phys. 55:963-983.
| |
| 1 i Dodd, B. and L. L. Humphries. 1989. " Errata -- Hazards Annenament of Worst Case l Transportation Accidents Having Typical Radioactive Material Shipments." Health Phys.
| |
| j 56:974.
| |
| ! ED. 1984. "NRPS - R166 'Joses in Radiation Accidents Investigated by Chromosome.
| |
| j Aberration Analysis. XIV A Review of Cases Investigated: 1983' by D. C. Lloyd, J. S. l
| |
| . Presser, J. E. Moquet and P. Finnion." J. Radiat. Prot. 4:189-190.
| |
| i l i ED. 1985. "Juarez Incident,1984." J. Radiat. Prot. 5:145-147.
| |
| Elliott, G. A. 1960. " Accidental Acute Irradiation from Cobalt-60." S. Afr. Med. J.
| |
| ; 34:524-529.
| |
| ! Evdokimoff, V. N.1980. " Hot Lab Radiopharmaceutical Accident: Potential Airborne
| |
| ! Release." Health Phys. 39:573-574.
| |
| i l Farina, R., C. E. Brnadan-Mello, and A. R. Oliveira. 1991. " Medical Aspects of Cs-137 l- Decorporation: 'Ibe Goiania Radiological Accident." Health Phys. 60:63-66.
| |
| 4
| |
| ! Finkel, A. M. 1990. Canfmatine Uacenniary in Rick Mane ===c A Gaida for Decision-Makers Center for Risk Management, Resources for the Future, Washington, DC.
| |
| g Fisher, D. R., K. L. Kathren, M. J. Swint. 1991. " Modified Biakiaat i c Model for
| |
| ! Uranium from Analysis of Acute Exposure to UF6." Health Phys, 60:335-342.
| |
| j Flatby, J., T. Hendnksen, and H. Host. 1983. The Radiarian Accident at the Institute for Faerry Tachnalory. Kieller. Norway. hae*=her 2.1982. D u =a ;c Evah>=tiaan.
| |
| : National Institute of Radiation Hygiene, Statens Instituit for Stralebygiene Report.
| |
| j Fry, F. A. 1976. "Imog-term Retention of Americium-241 Following Accidental Inhalation." Health Phys. 31:13-20.
| |
| ]
| |
| F-42 4
| |
| ~
| |
| ~ -.r - ,
| |
| | |
| i.
| |
| i 4 .
| |
| 4
| |
| ! Fry, S. A., L. G. Littlefield, C. C. Lushbaugh, et al. 1990. " Follow-up of Survivors of !
| |
| i Serious Radiation Accidents in the United States," pp. 373-3%. In: The Medical Basis for i
| |
| Rs :: tion Accidaar PrensrMaass II - Clinical E rerience and Follow-uo Since 1979.
| |
| ]- PracaMinen of the Secand International REAC/TS Coaference on the Medical R= sis for
| |
| [ Radsation Acgident Preparedness. Eds. R. C. Ricks and S. A. Fry. Elsevier Science i Publishing Co., New York.
| |
| Fuqua, P. A., W. D. Norwood, and S. Mvks. 1%5. " Biologic Effects of Human l Radiation Exposure; Report of a Criticality Accident." J. Occuo. Med. 7:3:85-93.
| |
| 1 I
| |
| Gilberti, M. V. 1980. "The 1%7 Radiation Accident Near Pittsburgh, Pennsylvania, and a Follow-up Report," pp. 131-140. In: The Medic =I R==le for Radiarian Accidaat Pra= =taa*= . Prcraadinen of the REAC/TS International Confer *aca: The Madical Rncic for j Radiarian Accidaar Preamredaaes. Eds. K. F. Habner and S. A Fry. Elsevier Science Publishing Co., New York.
| |
| l
| |
| : Hatamermaier, A., E. Reich, and W. Bogt. 1988. "An Evaluation of Chemiluminescence from Pharmaceuticals and OQer Solids for Neutron Accident Dosimetry." Radiat. Prot.
| |
| Dosim. 25:97-100.
| |
| Harrison, N. T., P. C. Escott, G. W. Dolphin, et al. 1974. "The Investigation and Reconstruction of a Severe Radiation Injury to an Industrial Radiographer in Scortand," pp. .
| |
| 760-768. In: Prar**diaen of the Third International Coneress of therae,ra eia-i nadineian Protection Association. Ed. W. S Snyder. U.S. Atomic Energy Commission, Washington, D.C.
| |
| Hashimme, T. Y. Kato, T. Nakajima, H. Yamaguchi, and K. Fujimoto. 1972. " Emergent Dose, Enrimation of Non-Occupational Perscas Accidentally Exposed to IR-192 Gamma Rays." Health Phys. 23:855-7.
| |
| Hasterlik, R. J. and L. D. Marinelli. 1956. " Physical Dosimet4y and Clinical Observations on Four Human Beings Involved in an Accidental Critical Assembly Excursion." In:
| |
| Praca diaen of an I--r--d I Cnafercara on the Pa=c-Ful Uses of Ata ain Faarirv. United Nations, Geneva, Switzerland.
| |
| Harrison, N. T., P. C. Escott, G. W. Dolphin. 1974. "The Invesdgatica and Reconstruction of a Severe Radiation Injury :c, an Industrial Radiograpt'4r in Scotland." pp.
| |
| 760-768. In: International Congress of the IRPA, Washington,1973. CONF-730907.
| |
| Heaton, B., and A. A. Murray.1982. " Experience Gained from tie Recovery of Imst Radioactive Sources." J. Radiat. Prot. 2:34-35.
| |
| Heid, K. R., B. D. Breitenstein, H. E. Palmer, B. J. McMurray, N. Wald. 1979. Ibs.
| |
| 1976 Hanford Americium Accident. TID-28938. Pacific Northwest laboratory, Richland, F-43
| |
| | |
| l i :
| |
| i F l Washington, j l
| |
| i Heio, K. R., B. D. Breitenstein, H. E. Palmer, et al. 1980. "The 1976 Hanford Americium Accident," pp. 345-355. In: Th.e Medical Basis for Radiation Accident l'
| |
| Prenaredness. PraceMines of the REAC/TS International Conference: The MMici! Basis for Radiation Accident Preamredne_u. Eds. K. F. Habner and S. A. Fry. Elsevier Science Publishing Company, New York.
| |
| Hirashima, K., H. Sugiyama, T. Ishihara, et al. 1980. "The 1971 Chiba, Japan Accident:
| |
| Exposure to Iridium 192," pp.179-195. In: The Medical Basis for Radiatian AccWnt Pi==*=. Pracaadinem of the REAC/TS Interantiaani Conf:r-ace: 'Ihe MMicil Basis for Radiation Accident Preparedness. Eds. K. F. Hahner and S. A. Fry. Elsevier Science Publishing Company, New York.
| |
| Holly, F. E. and W. L. Beck. 1980. " Dosimetry Studies for an Industrial Radiography Accident," In: The Medical Basis for Radiation Accident Preoaredness. Proceedinas of the REAC/TS International Conference: The Medical Basis for Radiation Accident Preparedness.
| |
| Eds. K. F. Hahner and S. A. Fry. Elsevier Science Publishing Company, New York. pp. I 265-277.
| |
| Hurtado, R. M., R. Sccin, M. Marquez, R. C. Ricks, M. E. Berger. 1991. "The Radiological Accident in El Salvador: Psychological Aspects." In: The Medical Basis for Radiatian AccWat PranarMaacs III. Pracaadinen of the 'Ibird REAC/TS Interantian=3 Conference 'Ihe MMical R==is for Radiarian AccWat P_renaredness. Eds. R.C. Ricks, M.
| |
| E. Berger, and F. M O'Hara. Elsevier Science Publishing Company, New York. pp. 187-191.
| |
| I Hurst, G. S., R. H. Ritchie, F. W. Sanders, P. W. Reinhardt, J. A. Auxier, E. B. Wagner, ;
| |
| A. D. Callihan, and K. Z. Morgan KZ. 1%1. " Dosimetric Investigation of the Yugoslav l Radiation Accident." Health Phys. 5:179-202.
| |
| l International Atomic Energy Agency (IAEA).1%2. "The Vinca Dosimetry Experiment."
| |
| Technical Report Series No.6, International Atomic Energy Agency. IAEA, Vienna, Austria.
| |
| International Atomic Energy Agency (IAEA). 1%9. " Handling of Radiation Accidents."
| |
| Proceedings Series, IAEA, Vienna, Austna. '
| |
| International Atomic Energy Agency (IAEA). 1974. " Evaluation of Radiation Emergencies and Accidents: Selected Criteria and Data." Technical Report Series No.152, IAEA, Vienna, Austria.
| |
| International Atomic Energy Agency (IAEA). 1977. " Handling of Radiation Accidents 1977." Proceedings Series, IAEA, Vienna, Austria.
| |
| i F-44
| |
| | |
| i 1
| |
| l International Atomic Energy Agency (IAEA) 1986. " Derived Intervention Levels for i Application in Controlling Radiation Doses to the Public in the Event of a Nuclear Accident !
| |
| or Radiological Emergency: Principles, Proced-res and Data." Safety Series No. 81, IAEA, Vienna, Austria.
| |
| ]
| |
| International Atomic energy Agency (IAEA). 1988a. " Medical Handling of Accidentally Erpaa~i Individuals." Safety Series No. 88, IAEA, Vienna, Austria.
| |
| International Atomic Energy Agency (IAEA).1988b. "The Radiological Accident in Goiania." IAEA, Vienna, Austria.
| |
| International Atomic Energy Agency (IAEA). 1989a. " Evaluating the Reliability of Predictions Made Using Environmental Transport Models." IAEA Safety Series No.100, )
| |
| IAEA, Vienna, Austria. '
| |
| International Atomic Energy Agency (IAEA).1989b. "Ernergency Planmng and Preparedness for Accidents Involving Radioactive Materials Used in Medicine, Industry, !
| |
| Research and Trainmg." Safety Series No. 91. pp. 51-79. IAEA, Vienna, Austria.
| |
| International Atomic Energy Agency (IAEA).1990. "The Radiological Accident in San Salvador." IAEA, Vienna, Austria.
| |
| International Atomic Energy Agency (IAEA). 1992. " Application of Exemption Principles to the Recycle and Reuse of Wierials from Nuclear FacilMies." IAEA Safety Series No.
| |
| 111-P-1.1, IAEA, Vienna, / astria.
| |
| International Atomic Energy Agency (IAEA). 1993. "The Radiological Accident in Soreq."
| |
| IAEA, Vienna, Austria.
| |
| International Comminion on Radiological Protection (ICRP). 1977. " Recommendations of the International Comminion on Radiological Protection." ICRP Publication 26, Pergamon Press, Oxforti.
| |
| International Comminion on Radiological Protection (ICRP). 1991. "1990 Recommendations of the International Commission on Radiological Protection." ICRP Publication 60, Pergamon Press, Oxford.
| |
| Jacob, P. and R. Meckbach. 1987. " Shield.ag Factors and External Case Evaluation."
| |
| Radiat. Prot. Dosim. 21:79-85.
| |
| Jacob, P., H. G. Paretzke and H. Rosenbaum. 1988. " Organ Doses from Radionuclides on the Ground. Part II. Non-trivial Time Dependence." Health Phys. 55:37-49.
| |
| Jacobson, A., B. W. Wilson, T. E. Banks, and R. M. Scott. 1977. mIr Over Industrial F-45
| |
| | |
| _ _. . _ . _ _ _ _ _ _ . _ . _ _ ~ _ _ _ _ . _ . . - . _ _ _ _ _ _ . _ _ _ . _ _ _ . . _ . _ _ _
| |
| i i
| |
| i l Radiography." Health Phys. 32:291-293, 1 i
| |
| j Jalil, A. and K. A. Rab Molla. 1989. "An Overexposure in Industrial Radiography Using
| |
| : An Ir-192 Radionuclide." Health Phys. 57:117-119.
| |
| l l Janunet, H., G. Mathe, B. Pendic, et al. 1959. " Etudes de six cas d' irradiation totale j accidenselle." Revue Francaise d' Etudes Cliniques et Biologiques 4:210-225.
| |
| l
| |
| ! Jammat, H. P., R. Gongora, R. Le Go, et al. 1966. " Observation Clinique et Traitement l l d'un cas d' irradiation globale accidentelle." Proceedings of the First International Congress i i of Radiation Protection (IRP). IRP Health Physics, Roma, Italy. 5-9 Lf der 1966. i 1
| |
| ! Jammet, H., R. Gongora, P. Pouillard, et al. 1980. "The 1978 Algerian Accident: Four l
| |
| ! Cases of Protracted Whole-body Irradiation," pp. I13-129. In: 'Ibe Medical Basis for l Radiation Accident PrenarMnaes. Praemiinn of the REAC/TS Inearnatianal Canremnce: )
| |
| l The Medical Basis for Radiation Accident Prenaredness. Eds. K. F. Hahner and S. A. Fry. l i Elsevier Science Publishing Co., New York, i Jammet, H., R. Gongora, P. Jockey, et al. 1980. "The 1978 Algerian Accident: Acute !
| |
| 4 Local Frnaeme of Two Children," pp. 229-245. In: The Medical Basis for Radiation i Accident Preparedness. Proceedmas of the REAC/TS International Conference: The Medical 2- Basis for Radiation Accident Preparedness. Eds. K. F. Habner and S. A. Fry. Elsevier d i Science Publishing Co., New York.
| |
| 1 J
| |
| i Johnson, J. R. 1978. " Summary of Bionssay and Thyroid Monitoring Results Following an Accidental Exposure to I-125." Health Phys. 34:106-107.
| |
| Tohnson, J. R., D. W. Dunford, and G. H. Kramer. 1983. "9=-- y of a Strontium-89 i Conramination Case." Radiat. Prot. Dosu' n. 5:247-249.
| |
| Kathren, R. L. and R. H. Moore. 1986. " Acute, Accidental Inhalatian of U: A 38-year Follow-up." Health Phys. 51(5):609-619.
| |
| Kerciakes, J. G. 1978. " Handling of Radiation Accidenta 1977" Med. Phys. 5:457.
| |
| Klener, V., R. Tuscany, J. Vejlupkova, J. Dvorak, and P. Vlkovic. 1986. "Long-term Follow-up After Accidental Gamma Irradiation from a Cc40 Source." Health Phys.
| |
| 51:601-601.
| |
| Kirchmam, R., G. S. Gerber, E. 'Fagniart, C. M. Vandecasteele, and M. van IIe:s. 1986.
| |
| " Accidental Release of Elemental Tritium Gas and Tritium Oxides: Models and In Situ Experunents on Various Plant Species." Radiat. Prot. Dosim. 16:107-110.
| |
| Kumatori, T., K. Hirashima, T.1stuhara, et al. 1977. " Radiation Accident Caused by an F-46
| |
| | |
| l i
| |
| . 1 i Iridium-192 Radiographic Source," pp. 35-42. In: Eroceedines of a Symoosium on Hnadline j Radiarian Accidente. International Atomic Energy Agency, Vienna, Austria.
| |
| 1 1 j !
| |
| Ian7l, L. H., M. L. Rozenfeld, A. R. Tarlov. 1%7. " Injury Due to Accidental High-Dose '
| |
| Exposure to 10 MeV Electrons." Health Phys. 13:241-251.
| |
| I j
| |
| Im Grand, J., J. C. Croize, T. de Dorlodot, and Y. Roux. 1987. "Simrinrical Survey of the i ,
| |
| Housing Characteristics and Evaluation of Shielding Facters in the Surrounding of French j !
| |
| Nuclear Sites." Radiat. Prot. Dosim. 21:87-95. I levanon, I.- and A. Pernicit. 1988. "The Inhalation Hazard of Radioactive Fallout." Health j Ehy.L 54:6:645-657.
| |
| 1 i t 1 i Lipsztein, J. L., P. G. Cunha, and C. A. N. Oliveira. 19912. "The Goiania Accident:
| |
| 4 Behind the Scenes." Bealth Phys. 60:5-6.
| |
| a Lipsztein, J. L., L. Bertelli, D. R. Melo, A. M. G. F. Azeredo, L. Juliao, and M. S. )
| |
| Santos. 1991b. " Application of in-vitro Bioassay for Cs-137 During the Emergency f aase !
| |
| i of the Goiinia Accident." Health Phys. 60:43-49. I i
| |
| l j' Upsztein, J. L., L. Bertelli, C. A. N. Oliveira, and B. M. Dantas. 1991c. " Studies of Cs j Perentian in the Human Body Related to Body Parameters and Prussian Blue -
| |
| i
| |
| } Administration." Health Phys. 60:57 41.
| |
| l Uster, B. A. J. 1986. " Contaminated Mexican Steel Incident." J. Radine. Prot. 6:48.
| |
| i
| |
| ; Littlefield, L. G., E. E. Joiner, S. P. Colyer, R. C. Ricks, C. C. f =hhanurh, and R. .
| |
| j Hurtado-Monroy. 1991. "The 1989 San Salvador Co-60 Ra 'iation Accident: Cytogenetic
| |
| ; Dosimetry and Follow-up Evaluations in Three Accident Victims." Radiat. Prot. Dosim.
| |
| i 35:115-123.
| |
| i.
| |
| l Uoyd, D. C., R. J. Purrott, J. S. Prosser. 1978. Da== in radiarian accidaar invcA r.1
| |
| ! by chran=ama=== mherreeian annivsis. VIII: A rev:ew of c== inv =:. r- i - 1977. NRPB-
| |
| ] R70, United Kingdom.
| |
| j Ucyd, D. C., R. J. Purrott, J. S. Prosser. 1979. Doses in naintion accidant inva=rientad 4
| |
| by chrnmaname sherration annivsis. IX: A review of emaae inveetiea'ad - 1978. NRPB-R83.
| |
| 2 United Kingdom.
| |
| ; Doyd, D. C., A. A. Edwards, J. S. Presser, A. Auf der Maur, A. Etzweiler, U.
| |
| ; Weickhardt, U. Gossi, L. Geiger, U. Noelpp, and H. Rosier. 1986. " Accidental Intake of j Tritiated Water: A Report of Two Cases." Radiat. Prot. Dosim. 15:191-1 % .
| |
| ) Lubenau, J. O., J. S. Davis, D. J. Mcdonald, and T. M. Gerusky. 1%9. " Analytical i
| |
| !; F-47
| |
| )
| |
| | |
| i i
| |
| I
| |
| 'i. '
| |
| I
| |
| ! X-ray Hazards: A Continuing Problem." Health Phys. 16:739-746.
| |
| L 1
| |
| Lubenau, J. O. and T. M. Gerusky. 1971. " Radiation ancients Registry - Pennsylvania ;
| |
| { Experiena." Health Phys. 21:605-7. i 1
| |
| l Imbenau, J. O. and D. A. Nesshnamer. 1986. " Comment of 'A Possible Hazard: Pressure
| |
| ! Build-up in Scaled Ampoules of Radionuclides in Aqueous Solution'."
| |
| Health Phys. 51:147-148.
| |
| Imshbaugh, C. C., S. A. Fry, K. F. Habner, and R. C. Ricks. 1980. " Total-body Irradia: ion: A Historical Review and Follow-up," pp. 59-79. In: The Medical Basis for P=diarian Accidaat Prmanradaa==. Pra-dia=< of the REAC/TS I=- d-- I Cr a ew:
| |
| 'Ihe Medical basis for Radiarian Accidaat Pr===inaes. Eds. K. F. Hahner and S. A. Fry.
| |
| Elsevier Science Publishing Cc., New York.
| |
| Imshbaugh, C. C. 1981. " Management of Persons Accidentally Contaminated with ,
| |
| Radionuclides, NCRP Report 65" by National Council on Radiation Protection and Measurements. Med. Pnys. 8:525-526.
| |
| Imhhaugh, C. C., S. A. Fry, R. C. Ricks, et al. 1986. "HistorL1 Update of Past and ;
| |
| Recent Skin Damage Radiation Accidents." In: Radiation Damage to Skin: Fundamental and l Practical Aspects. Eds. H. Jammet, F. Daburon and G. B. Gerber. Bnt, inst. Radiol.
| |
| (Iandon):7-12.
| |
| Imshbaugh, C. C., R. C. Ricks, S. A. Fry. 1988. " Radiological Accidents. A Historical Review of Scaled Sources Accidents." Pmceedings of an International Conference on P=diarian Prat _actian in Nncianr Fnerry. IAEn-CN-51-92, laternational Atomic Energy Agency, Vienna, Austria.
| |
| Imshbaugh, C. C., S. A. Fry, A. Sipe, R. C. Ricks. 1990. "An Historical "w.pstive of Human Involvement in Radiation Accidents." In: Radiation Protection Today 'Ibe NCRP at Sixty Years. Prn-dia=< of the Twenty-Fifth Ananal Meeting April 54,1989. pp.171-188. NCRP Pracaadinen No.10, NCRP, Bethesda, Maryland.
| |
| Imshbaugh, C., G. Eisele, W. Burr Jr., K. Hubner, and B. Wachholz. 1991. " Current !
| |
| Status of Biolmgical indicators to Detect and Quantify Previous Exposures to Radiation."
| |
| Health Phys. 60(Supp 1):103-109.
| |
| Maass, A. R. and T. L. Flanagan. 1%3. " Accidental Personnel Exposure to Elemental S-35." Health Phys. 9:731-740.
| |
| Majborn, B. 1984. " Estimation of Accidental Gamma Dose by Means of
| |
| '1hermohnminescence from Watch Jewels. Health Phys. 46:917-919.
| |
| i F-48 !
| |
| | |
| _ . _ _ . _ _ _ _ . . _ _ . _ . _ _ _ _ _ . - . . _ . _ m ._ _..
| |
| 4 i
| |
| !' Marshall, E. 1984a. "Juarez: An Unprecedent:d Radiation Accident." Science 73:1152-1154.
| |
| j Marshall, E. 1984b. " Morocco Reports lethal Radiation Accident " Science 225:395.
| |
| 1 I Martin, J. B. 1991. "'The Radiological Accident in San Salvador' by IAEA." l Health Phys. 61:578. !
| |
| : I l Matetskos, C. J. and C. C. Lushbough. 1991. "The Goiinia Radiation Accident." Health
| |
| ] EhxL 60:1. -,
| |
| i
| |
| { Matthews. J. D. 1970. " Accidental Extremity Exposures from Analytical X-ray Beams."
| |
| s Health Phys.18:75.6.
| |
| 3 Maxfield, WS, and GH Porter. " Accidental radiation exposure from iridium-192 camera."
| |
| l 1969. In: Handline of Radiation Accidents. IAEA, Vienna: 459-467.
| |
| l McMurray, B. J. 1983. "1976 Hanford Americium Exposure Incident: Accident Description." Health Phys. 45(4):847-853.
| |
| ,. Mettler, F. A., and R. C. Ricks. 1990. " Historical Aspects of Radiation Accidents." In:
| |
| j Merib=1 Manneement of Radiation Accidents. Ed. F.A. Mettler, C.A. Kelsey, R. C. Ricks. l
| |
| ] CRC Press, Boca Raton, Florida. pp.17-32. !
| |
| i j Mexican Cobalt-60 Incident-1984. Aerial Measuring System (AMS) Surveys, Nenot JC. ;
| |
| ; Overview of the Radiological Accidents in the world, Updated December 1989. Internat. J.
| |
| j Rad. Biol. 57:1073-1085, 1990.'
| |
| ! Minder, W. 1%9. "Interne Kontamination mit Tritium." Strahlentberpic 137:700-704.
| |
| i i
| |
| Morgan, J. 1989. ' Medical Handling of Accidentally Ernaead Individuals' IAEA Safety Series No. 88. J. Radine. Prot. 9:232.
| |
| l Morgan, M. G. and M. Henrion. 1990. Uncertainty. A Guide to Dealine with Uncertainty l in On=atie=*ive nad Policy Risk Analysis. Cambridge University Press, New York. j i Mossman, K. L. 1989. " Health Physics Annotated Bibliography" by Charles A. Willis.
| |
| l Health Phys 57:790.
| |
| I a
| |
| : National Academy of Sciences (NAS), National Research Council. 1988. Health Risks of
| |
| ;- P=daa mad Other Internally Dana =ited Alnha Fmitters: BEIR IV. National Academy Press,
| |
| , Washington, DC.
| |
| National Academy of Sciences (NAS), National Research Council. 1990. Health Effects 'of
| |
| ! F-49 i
| |
| i b
| |
| | |
| i 4
| |
| d i
| |
| Fraamre to Iow 12vels ofIoni7ine Radiation: BEIR V. National Academy Presa; Committee on the Biological Effects oflonizing Radiation, Washington, DC.
| |
| 4 i Nenot, J. C. 1990. Review: Overview of the Radiological Accidents in the World, Updated j December 1989. Int. J. Radiat. Biol. 57(6):1073-1085.
| |
| i j
| |
| Nenot, J. C. 1993a. Bilan des accidents d' exposition interne. Radioprotection 28(3):265-j 277.
| |
| )
| |
| l Nenot, J. C. 1993b. Accident d' irradiation en Chine. Radioprotection 28(4):453-455.
| |
| j Newman, H. F. 1990. "The Malfunction 115411 Accelerator Accidents 1985, 1986, 1987,"
| |
| l pp.165-171 In: The Medical R==ie for Radiation Accidaat Pream.rednaee II Clinic =1 i
| |
| F2r-ikc-* mad Follow-un Since 1979. Pir udines of the Lcc.r i frta..utic. sal REAC/TS Conference on the Medici! Ramic for Radiation Accident Prenaredr- ss. Eds. R. C. Ricks
| |
| !- and S. A. Dv. Elsevier Science Publishing Co., New York.
| |
| Newton, D., J. Rundo, and J. D. Fakins. 1981. "IAng-term Retention of The-228 Following Accidental Intake." Health Phys. 40:291-298.
| |
| Oberhofer, M. and J. L. Bacelar Izao. 1988. The Radiolonical Accidaat at GoiAnin.
| |
| International Atomic Energy Agency, Australia, i i Oliveira, A. R. 1987. "Un Repertoire des Accidents Radiologiques 1945-1985."
| |
| : Radioprotection 22(2):89-135.
| |
| l Oliveira, A. R., C. E. Brandeo-Heilo, N. J. L. Valverde, R. Farma, and M. P. Curado.
| |
| 1991. "Iacalized Iesions Induced by Cs-137 During the Golama Accident." Health Phys.
| |
| 60:25-29.
| |
| ; Oliveira, A. R., J. G. Hunt, N. J. L. Valverde, C. E. Brandso-Mello, R. Farina. 1991.
| |
| j " Medical and 3 elated Aspects of the Goiinia Accident: An Overview." Health Phys.
| |
| i 60:17-24.
| |
| -m j
| |
| Oliveira, A. R., N. J. Valverde, C. E. Brandan-Mello, et al. 1990. " Skin Iesions i
| |
| Arscelated ivith the GoiAnia Accident," pp. 173-181. In: The Medical Basis for Radiation Accidaat Preansedna== II - Clinical Fwr' awe and Follow-un Since 1979. Prc =*= of the -W Ira. :ic-si REAC/TS Conimme on the Meacal Raele for Radiatian Accident Preparedness. Eds. R. C. Ricks and S. A. Fry. Elsevier Science Publishing Co., New York.
| |
| Oliveira, C. A. N., R. Farina, L. Bertelli, A. T. Natarajan, A. T. Ramalho, and B. M.
| |
| Dantas. 1991. " Measurements of Cs-137 in Blood from Individuals Exposed During the -
| |
| Goiinia Accident." Health Phys. 60:41-42.
| |
| F-50
| |
| | |
| ___. . _ _ . . _ _ ~ . _ _ _ - _ _ _ _ _ . . _ . _ _ _ . . _ _ _ . . _ _ _ . . _
| |
| k l
| |
| t j Oliveira, C. A. N., N. C. I.ouretico, B. M. Dantas, and E. A. 1991. " Design and -
| |
| j Operation of a Whole-body Monitoring System for the Goinnia Radiation Accident." Health Phn 00:51-55.
| |
| : "ORAU Report:" see Stabin et al.1987.
| |
| 1 1
| |
| ! Parmentier, N. C., J. C. Nenot, and H. J. Jammet. 1980. "A Dosimetric Study of the l Belgian and Italian Accidents," pp. 105-112. In: The Medical Basis for Radiation Accident l Preparedness. Proceedings of the REAC/TS International Conference: 'Ibe Medical Basis for
| |
| ; Radiation Accident Preparedness. Eds. K. F. Hubner and S, A. Fry. Elsevier Science Publishing Co., New York.
| |
| 3 i
| |
| j Parmentier, N. C., J. C. Nenot, and C. Pannemier. 1990. "Two Cases of Accidental
| |
| { Protracted Overexposure: Aspects of an Extensive Bone Marrow Study.," pp. 29-51. In:
| |
| j Ihe Medical Basis for Radiation Accident Prepared _ ness II Clinical Experience and Follow-uo l Since 1979. Proceedinn of the Second International REAC/TS Conference on the Medical l Basis for Radiation Accident I reparedness. Eds. R. C. Ricks and S. A. Fry. Elsevier Science Publishing Co., New York.
| |
| l f Petkau, A. and S. D. Pleskach. " Case of Accidental Aspiration of Sr-90 (CL 2)." Health j Ehyt 22:87-90.
| |
| 1 I Pow',ewell, D. S., and G. J. Ham. 1989. " Distribution of Plutonium and Americium in l Tissues from a Human Autopsy Case." J. Radiat. Prot. 9:159-164. !
| |
| s i
| |
| ! l Purrott R. J., G. W. Dolphin, D. C. Uoyd. 1972. The study of chromosome aberration
| |
| ! vield in hamna tv5'-m as an indieaear of radiatian da . II: A review of e===
| |
| l invec:r- ' - 1970-71. NRPB-RS, United Kingdom.
| |
| } Purrott R. J., G. W. Do4 bin, D. C. Uoyd. 1973. The study or chromosome aberration sield in hmnan lymphievtes ** an indicaear of radiatian dam. II: A review of e=== l
| |
| ; w rir- ' - 1971-72. NRPB-RIO, United Kingdom.
| |
| i
| |
| ! Purrott R. J., D. C. Lloyd, J. S. Prosser. 1976. "Ihe study of chmmosome aberration vield l in hn=na Ivmahncytes as an indica'ar of radiarian dame. VI: A review of ca== investigated - )
| |
| j 127_1. NRPB-R41, United Kingdom. I i
| |
| ! Preston R. H., J. G. Brewen, N. Gengozian. 1964. " Persistence of radiation induced i chromosome aberrations in mannoset and man." Radiat. Res. 60:516-524.
| |
| i 1
| |
| " Radiation Accident Grips Goinnia. News and Comment." Science 238:1028-1031, November 1987.
| |
| 1 F-51
| |
| | |
| i i
| |
| Ramalho, A. T., A. C. M. Nescimento, and A. T. Netarajan. 1988. " Dose Assessments by i Cytogenetic Analysis in the Goiinia (Brazil) Radiation Accident." Radiat. Prot. Dacim.
| |
| j 23:121-154.
| |
| i Ramalhn, A. T., and A. C. H. Mascimento. 1991. "The Fate of Chromosomal Aberrations in Cs-137 Exposed Individuals in the Goiinia Radiation Accident." Health Phys. 60:67-70.
| |
| { Reitan, J. P., P. Stavem, K. Kett, et al. 1990. "The "Co Accident in f!orway,1982: A
| |
| ; Clinical Reappraisal," pp. 3-11 In: The Medical Baie for Rac'intian Accid-at Fr.=Ms=
| |
| 1 II - Cli=bal Rr=.'=+ =ad FoHaw-un Sine
| |
| * 1979. Prc-W'=; of the Lc- i I.c...ri .r.sl REAC/TS Conference on the MMic=1 hele for Radiation Accident Pie ='wish. Eds. R.
| |
| ] C. Ricks and S. A. Fry. Elsevier Science Publishing Co., New York.
| |
| i i
| |
| Ricks, R. C.1991. "The Scope of the Problem." In: The Medical baie for Radiation i
| |
| Accident PreserM== III. Pracadines of the Third REAC/TS Internations! Conferec-x:
| |
| { The Medical Basis for Radiation Accident Preaaredness. pp. 3-10. Eds. R.C. Ricks, M. E.
| |
| Berger, and F. M O'Hara. Elsevier Science Publishing Company, New York.
| |
| i j
| |
| Ricks, R. C., S. A. Fry, A. H. Sipe, M. E. Berger, F. H. Fong, C. C. Lushbaugh. 1992, 1
| |
| " History of Radiation Accidents." In: The Biological Basis of Radiatian Prataction Practice.
| |
| l pp. 218-225. Eds. K. L. Mossman and W. A. Mills. Williams and Wilkins, Baltimore, j Maryland.
| |
| Rosenthal, J. J., C. E. de Almeida, and A. H. Mendonca. 1991. "The Radiological Accident in Goiinia: The Initial Remedial Actions." Ha=Ith Phys. 60:7-15.
| |
| Ross, J. P., F. E. Holly, H. A. Zarem, et al. 1990. "The 1979 los Angeles Accident:
| |
| Fraa=2 e to Iridium 192 Industrial Radiographic Source," pp. 205-221. In: The Medical Basis for Radiariaa Accidaar Preaanwina==. Procadia== of the REAC/TS International Cr 5 ;--=e The Madient b=3= for Radiation Acci.'-e Fr =M==. Eds. K. F. Hahner and S. A. Fry. Elsevier Science Publishing co., New York.
| |
| Rossi, E. C., A. A. Thorngate, and F. C. Larson. 1%2. " Acute Radiation Syndrome Caused by Accidental Fraa=J e to Cobalt-60." J.12b. Clin. Med., 59:655-666.
| |
| Rubin, L. S. 1978. "The Riverside Radiation Tragedy." Columhas Monthly 52-66.
| |
| Saenger, E. L., J. G. Kerciakes, N. Wald, et al. 1974. " Clinical Course and Dosimetry of Acute Hand Injuries to Industrial Radiographer from Multicurie Scaled Gamma Sources."
| |
| Proceedings of the Thiri Internatianal Conrress of faterr.mtia==1 Radiatian_ Prneretian Association United States Atomic Energy Commission, Ofnce of Information Services (Tech Div.),1:773-782.
| |
| Sagell H. "Ein lehrreicher Strahlenunfall." 1975. Arbeitsmed, Sozialmed. Praventivmed F-52 l
| |
| 1
| |
| | |
| -. . - - . ~ - . - . - . ... - --.- - . - . - . . - . - . . . . . - ~ _-. . . -
| |
| i l
| |
| i-
| |
| ! 24.
| |
| i Sanders, S. M. Jr. 1974. " Excretion of Am-2% and Cm-244 Following Two Cases of
| |
| ; Accidental Inhalation. Health Phys. 27:359-65 j Schneider G. J., B. Chones, T. Blonnigen. 1%9. " Chromosomal Aberrations in a Radiation Accident: Dosimetric and Hematological Aspects." Radiat. Res.- 40:613-617.
| |
| Scott, L. M. and C. M. West. 1975. " Excretion of PO-210 Oxide Following Accidental l Inhalatian " Health Phys. 28:563-565.
| |
| Scott, E. B., Jr. 1980. "The 1978 and 1979 Louisiana Accidents: FQawre to Iridium l 192," pp. 223-262. In: The Medical Basis for Radiarian AccWat Pren=redaass.
| |
| Procesiines of the REAC/TS Irtemational Conferaare: The Medical Rmeie for radiation Accident Preparedness. Eds. K. F. Hilbner and S. A. Fry. Elsevier/ North Holland, New York.
| |
| l Smith, E. E. 1982. "The Recovery on 29th March 1944 of a Lost Smg Radium Tube."
| |
| Radiat. Prot. 2:39-40.
| |
| Smith, H. 1983. " Dose - Effect Relationships for Early Response to Total Body Irradiation." J. Radiat. Prot. 3:510.
| |
| Smith, N. 1985. "Assessumnt of and Therapy Followint, Contamination Through the Lungs or Gastrointestinal Tract." J. Radiat. Prot. 5:15-20.
| |
| Sowby, D. 1989. 'The Radiological Accident in Goinnia' by IAEA. J. Radiat. Prot. 9:78.
| |
| Sowby, D. 1990. 'The Radiological Accident in San Salvador' by International Atomic Energy Agency. J. Radiat. Prot.10:313.
| |
| Stabin, M., K. Paulson, and S. Robinson. 1987. Inaproper Transfer / Disposal Scenarios for Generally Licensed Devices. "'Ibe ORAU Report" produced under NRC FIN B0299. Oak Ridge Associated Universities, Oak Ridge, Tennessee.
| |
| Steidley, K. D. 1976. "A Co-60 Hot Cell Accident." Health Phys. 31:382-385.
| |
| Steidley, K. D, G. S. Zeik, and R. Ouellette. 1979. "Another Co-60 Hot Cell Accident."
| |
| Health Phys. 36:437-441.
| |
| Steinhausler, F. 1987. "'Ibe Effect of Fall-out Deposition on Indoor Gamma Radiation levels in Single-family Dwellings." Radiat. Prot. Dosim. 21:103-105.
| |
| Stephan, G., W. Hadnagy, C. H. Hammaier, and U. Imhof. 1983. " Biologically and F-53
| |
| | |
| i i
| |
| j Physically Recorded Doses After an Accidental Exposure to Co-60 Gamma Rays. Health ,
| |
| Phys, 44:409-411.
| |
| ]
| |
| 4 j Stott, A. N. B. 1980. "Co-60 Hot Cell Accidents." Health Phys. 39:363-364.
| |
| Straume, T., R.' G. langloi;, J. Lucas, R. H. Jensen, W. L. Bigbee, A. T. Ramlho, and C.
| |
| j E. Brandao-Mello. 1991. " Novel Biodosimetry Methods Applied to victims of the Goiinia j Accident." Health Phys, 60:71-76.
| |
| 4 l Taylor, D. M. 1989. " Biological Aneument of Occupational Fraa-re to Actinides -
| |
| Round Table Discussion." P=diat. Prot. Daeim. 26:391-394.
| |
| "The Medical Basis for Radiation Accident Preparedness." 1980. Procaadines of the .
| |
| REAClTS International. Conference: The Niedical basis for Radiation Accident Preparedness. I
| |
| {
| |
| : Eds. K. F. Habner, and S. A. Fry. Elsevier/Nonh Holland, New York. !
| |
| ?
| |
| l i "The Medical Basis for Radiation Accident Preparedness II - Clinical Experience and Follow-up Since 1979." 1990. Proceediaen of the Second International REAC/TS l
| |
| Conference on the Medici! Basis for Radiation Accident Preparedness. Eds. R. C. Ricks j and S. A. Fry. Elsevier Science Publishing Co, New York.
| |
| ! Thompson, R. C., editor-in-chief. 1983. "1976 Hanford Americium Exposure Incident:
| |
| l Multiple Reports." Health Phys. 45:4.
| |
| I United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR).
| |
| 1988. Sources. Efface =. =ad Pi=Ir= of Ionisiae Padiarian. United Nations Publications, New i York.
| |
| ! United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR).
| |
| 1993. Sources and Effects of Ionizing Radiation United Nations Publications, New York. i 1
| |
| (
| |
| U.S. Nuclear Regulatory Commission (NRC). 1977. Ninth Aa=al Occuantiaani Padiation l
| |
| ; Exposure Report. NUREG-0322. NRC, Washington, DC.
| |
| ! U.S. Nuclear Regulatory Commission (NRC). 1977. Fraart to Coneress on Abnormal Occurrt;nces. NUREG-0090-6. NRC, Washington, DC.
| |
| i '
| |
| ! U.S. Nuclear Regulatory Comminion (NRC). 1985. Coa'amiantad Mericia S***1 Incident.
| |
| ; 1 a-Ma of Staal lo;s the US that had haan landvertrativ Coatamina**d with "Co as a i Result of Scrancine of a Teletherapy Unit. NUREG-1103. NRC, Washington, DC.
| |
| ' i U.S. Nuclear Regulatory Comminion (NRC). 1990. Inadvenent Shioment of a
| |
| )
| |
| i Radiaernahic Source from Korea to Amersham CorDoration. Burlington. Mas ==chucatts. l i NUREG-1405. NRC, Washington, DC.
| |
| l F-54
| |
| ]
| |
| 4 ,
| |
| | |
| U.S. Nuclear Regulatory Commission (NRC). 1993a. The NRC Calendar XII(6), Week ending February 12; p.1.
| |
| U.S. Nuclear Regulatory Commission (NRC). 1993b. Ioss of an Iridium-192 Scurce and Thernov Mie=dmini=tralian at Indiana Renional Cancer Center Indiana. Pennsylvanin. on November 16.1992. NUREG-1480. NRC, Washington, DC.
| |
| Valverde, N. J., J. M. Cordeiro, A. R. Oliveira, and C. E. Brandao-Mello. 1990. "The Acute Radiation Syndrome in the 137 Cs Brazilian Accident,1987," pp. 89-107 In: Ihg Madie=1 Ranie for P=diation Accident Prea= redness II - Clinie=1 Frnerience and Follow-up Since 1979. Prac-linom of the Sacond Intemat anali REAC/TS Conference on the Medical R==3e for Padiarian Accident Preparedness. Eds. R. C. Ricke and S. A. Fry. Elsevier Science Publishing Co., New York.
| |
| Vandecasteele, C. M., J. P. Dehut, S. Van Laer, D. Deprins, and C. Myttenaere. 1980.
| |
| "Long-term availability of Tc deposited on soil after accidental releases." Health Phys.
| |
| 57:247-254.
| |
| Vodopick H., and G. A. Andrews. 1974. " Accidental radiation exposure." Arch. Environ.
| |
| Health 28:53-56.
| |
| Vodopick, H. and G. A. Andrews. "The University of Tennessee Comparative Ammal Research laboratory Accident in 1971." 1980. In: The Madical Basis for Radiation AccWat Pnmaradaa==. Proca-dines of the REAC/TS Internarbnal Conference: The Medical Racie for Radiation Accident Preparedness. Eds. K. F. Hahner and S. A. Fry.
| |
| Elsevier/ North Holland:141-149.
| |
| Wade, L., Jr. 1972. " Accidental "Co Exposure at the Univasity of Tennessee Atomic Energy Commission Agricultural Research Laboratory." Nuclear Safety 13:304-308.
| |
| Webber, C. E. and J. W. Harvey. 1976. " Accidental Human Inhalation of Ruthenium Tetroxide." Health Phys. 30:352-355.
| |
| Widmer, D. J., K. W. Iogan, S. M. Langhorst, and W. L. Kemedy. 1986. " Gamma Camera Measurement of Accidental Internal Radionuclide Deposition: Ir-192 and Sm-153."
| |
| Health Phys. 51(3):349-351.
| |
| Yalcintas, M. G., T. D. Jones, H. R. Meyer, H. Ozer, and S. Unsal. 1980. " Estimation of Dose Due to Accidental Exposure to a Co-60 Therapy Source." Health Phys. 38:187-191.
| |
| Ye, G. Y., Y. Lui, N. Tien, B. Chiang, F. Chien, C. Xiae. 1980. "The People's Republic of China Accident in 1%3." In: The Medical Basis for Radiation Accident Preparedness.
| |
| Proceedines of the REAC/TS International Conference: The Medical Basis for Radiation Accident Preparedness. Eds. K. F. Hahner and S. A. Fry. Elsevier/ North Holland:81-89.
| |
| i F-55
| |
| | |
| 4 Yoshizumi, T, T, and C. A. Chuprinko. 1989. " Accidental CT Scanning Without Al i Filtration and Its Dosimetry." Health Phys. 56:253-254.
| |
| l Zeevaert, T., C. M. Vandecasteele, and R. Kirchmem. 1989. " Assessment of Dose to Man from Releases of Tc-99 in Fresh Water Systems. Health Phys. 57:337-343.
| |
| 9 3
| |
| l l l
| |
| + i n
| |
| I
| |
| .4 d
| |
| 4 4
| |
| 1 .
| |
| O 1
| |
| 1
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| , i a
| |
| 1 1
| |
| )
| |
| F-56
| |
| | |
| 6.0 TASK 6 APPENDIX
| |
| . Ae Task-6-A1. Individual dose egoivalent salues for the Indana, PA accident.
| |
| Gamma Gcmma*Ac (Sv/h t (Sv/h @
| |
| at Indiana Regional Cancer Center [m a2/Bq1) im)
| |
| Indiana, Pennsylvania, on November 16,1993, NUREG-1480 1.6E-13 0.021904 1.37E11 Sq (3.7 Ci) Ir-192 mSv mSv mSv Sv F_p Low High Avg. Avo. (h @ 1 m) 1 Table 6.3 Physician A 1.82 7.21 4.515 0.004515 0.206127 2 TaNe 6.3 TWTT-A 8.2 0.0082 0.374361 3 Table 6.3 RTT-8 1.1 0.0011 0.050219 l 4 TaNe 6.3 RTR 1.4 0.0014 0.063915 l 5 Table 6.3 Medical Physicist A 1.2 0.0012 0.054785 ;
| |
| 6 Table 6.3 Nurse A 6.3 0.0064 0.287619 !
| |
| 7 Table 6.4 Patient C 0.09 0.1 0.095 0.000095 0.004337 8 Table 6.4 Patient D 0.09 0.1 0.095 0.000095 0.004337 9 Table 6.4 Patient E 0.09 0.1 0.095 0.000095 0.004337 10 Table G.4 Patient F 0.09 0.1 0.095 0.000095 0.004337 11 Table 6.4 Patient G 0.09 0.1 0.095 0.000095 0.004337 12 Table 6.4 Patient H 0.09 0.1 0.095 0.000095 0.004337 l 13 Table 6.4 Patient 1 0.09 0.1 0.095 0.000095 0.004337 14 Table 6.4 Patient J 0.17 0.99 0.58 0.00058 0.026479 15 Table 6.4 Phlebotomist 3.5 13.9 8.7 0.0087 0.397188 16 Table 6.4 Office Manager 0.04 0.09 0.065 0.000065 0.002967 17 Table 6.4 Medical Secretary 0.04 0.09 0.065 0.000065 0.002967 1 18 Table 6.4 Tumor Registrar 0.04 0.09 0.065 0.000065 0.002967 19 Table 6.5 Patient K 2.4 4.3 3.35 0.00335 0.15294 20 Table 6.5 Patient L 2.4 4.3 3.35 0.00335 0.15294 21 Table 6.5 Administrative Aide A 0.09 0.58 0.335 0.000335 0.015294 22 Table 6.5 Administrative Aide B 0.09 0.58 0.335 0.000335 0.015294 23 Table 6.5 Laboratory Employee 0.05 0.23 0.155 0.000155 0.007076 l 24 Table 6.6 Ambulance Driver 4.8 8.4 6.6
| |
| * 0.0066 0.301315 25 Table 6.6 Ambulance Aide 13.3 25.7 10.5 0.0195 0.890248 26 Table 6.8 RN A 25 42 33.5 0.0335 1.529401.
| |
| 27 Table 6.8 RN 8 89 137 113 0.113 5.158875 28 Table 6.8 RN C 38 63 ' 4.5 0.0505 2.305515 29 Table 6.8 RN D 3.6 5.5 4.55 0.00455 0.207725 30 Table 6.8 GPN A 88 136 112 0.112 5.113221 l 31 Table 6.8 LPN A 56 87 71.5 0.0715 3.264244 32 Table 6.8 LPNB 112 174 143 0.143 6.526488-33 Table 6.8 LPN C 56 87 71.5 0.0715 3.264244 i 34 Table 6.8 LPN D 28 43 35.5 0.0355 1.620709 35 Table 6.8 CNA A . 103 148 125.5 0.1255 5.729547 j 36 Table 6.8 CNA B 52 74 63 0.063 2.876187 l 37 Table 6.8 CNA C 155 223 189 0.189 8.628561 l 38 Table 6.8 CNA D 02 74 63 0.063 2.876187 ;
| |
| 39 Table 6.8 CNA E 103 148 125.5 0.1255 5.729547 40 Table 6.8 CNA F 52 74 63 0.063 2.876187 1 41 Table 6.8 CNA G 52 74 63 0.063 2.816187 42 Table 6.8 CNA H 3.6 5.5 4.55 0.00455 0.207725 43 Table 6.8 CNA1 4.3 6.2 5.25 0.00525 0.239682 ;
| |
| 44 Table 6.8 Maintenance Man A 19 38 28.5 0.0285 1.301132 45 Table 6.8 Dietician 3.6 6.3 4.95 0.00495 0.225986 ;
| |
| F-57 i
| |
| | |
| Table Task-6-Al, continued. Individual dose equivalent values for the Indiana, PA accident.
| |
| I I
| |
| 46 Table 6.8 Activities Director 4.8 22 13.4 0.0134 0.61176 j 47 Table 6.10 Relative A 54.4 166 110.2 0.1102 5.031045 l 48 Table 6.10 Relative B 29.2 42 35.6 0.0356 1.625274 49 Table 6.10 Relative C 36 5 64.5 50.5 0.0505 2.305515 i 50 Tab!e 6.10 Relative D 23.5 36.5 30 0.03 1.369613 51 Ta' ale 6.10 Relative E 21.9 31.5 26.7 0.0267 1.218955 52 Talde 6.10 Relative F 21.9 31.5 26.7 0.0267 1.218955 l 53 Tab'e 6.10 Friend A 23.5 92.8 58.15 0.05815 2.654766 54 Tatle 6.11 Resident B 37.6 128 82.8 0.0828 3.780131 55 Tattle 6.11 Resident C 63.9 197 130.45 0.13045 5.955533 56 Table 6.11 Resident D 23.1 84 53.55 0.05355 2.444759 57 Tatse 6.11 Resident E 11.2 15.4 13.3 0.0133 0.607195 58 Table 6.11 Resident F 27.5 31.7 29.6 0.0296 1.351351 59 Table 6.11 Resident G 18.3 22 20.15 0.02015 0.919923 60 Table 6.11 Resident H 34.2 39.5 36.85 0.03685 1.682341 61 Table 6.11 Resident i 21.6 25.9 23.75 0.02375 1.084277 62 Table 6.11 Resident J 14.4 24.3 19.35 0.01935 0.8834 63 Table 6.11 Res. dent K 17.3 22.8 20.05 0.02005 0.915358 j 64 Table 6.11 Resident L 12.2 27.5 19.85 '0.01085 0.906227 1 65 Table 6.11 Resident M 57.3 90.9 74.1 O.0741 3.382944 66 Table 6.11 Resident N 13.8 23.4 18.6 0.0106 0.84916 67 Table 6.13 Driver A 1.7 0.0017 0.077611 68 Table 6.13 Driver B 2.5 5.2 3.85 0.00385 0.175757 69 Table 6.13 Driver C 0.34 0.36 0.35 0.00035 0.0159.79 70 Table 6.13 Supervisor A 13 51.3 32.15 0.03215 1.467768 71 Table 6.13 Safety Technician A 34.3 89.5 61.9 0.0019 2.825968 72 Table 6.13 Safety Technician B 26.9 68.4 47.65 0.04765 2.175402 75 Table 6.13 Other BFI 1 0.01.s 0.017 0.0148 1.48E-05 0.000676 74 Table 6.13 Other BFI 2 0.013 0.01'r 0.014'd 1.48E-05 0.000676 75 Table 6.13 Other BFI 3 0.013 0.017 0.0148 '1.48E-05 0.000676 76 Table 6.13 Other BFI 4 0.013 0.017 0.0148 1.48E-05 0.000676 77 Table 6.13 Other BFI 5 0.013 0.017 0.0148 1.48E-05 0.000676 78 Table 6.13 Other BFI 6 0.013 0.017 .0.0148 1.48E-05 0.000676 79 Table 6.13 Other BFI 7 0.013 0.017 0.0148 1.48E-05 0.000676 do Table 6.13 Other BFI 8 0.013 0.017 0.0148 1.48E-05 0.000676 81 Table 6.13 Other BFI 9 0.013 0.017 0.0148 1.48E-05 0.000676 82 Table 6.13 Other BFI 10 0.013 0.017 0.0148 1.48E-05 0.000676 83 Table 6.13 Other BFI 11 0.013 0.017 0.0148 1.48E-05 0.000676 84 Table 6.13 Other BFI 1211/29 0.004 0.007 0.00564 5.64E-06 0.000257 85 Table 6.13 Other BFI 1311/29 0.004 0.007 0.00564 5.64E-06 0.000257 Collective 2478 2.478 113.139 Average 29.16 0.02916 1.33105 Standard Deviation 40.75 0.041 1.860 Minimum 0.00564 5.64E-06 0.000257 Maximum 189.00 0.189 8.628561 No. individuals 85 85 85 Geometric Mean 2.91 0.00291 0.13304 Geometric Standard Deviation 25.3 25.3 25.3 F-58
| |
| | |
| i 1
| |
| 4 Table Task-6-A2. Individual dose equivalents and Time-and-Proxim;ty Factors for the 1990 i
| |
| Korea-USA 192 Ir Shipment Accident. ;
| |
| I Inadvertent Shipment of a Radiographic Source form Korea to Amersham j Corporation, Burlington, Massachusetts, NUREG-1405,1990 l Ir 192 Korea: 1/18//90 - 2/11/90 i 1 INC S/N 1062 USA: 2/11/90 - 3/8/90 2 Gamma Gamma *Ac
| |
| ; (Sv/h L (Sv/h @
| |
| ~
| |
| 74.6 days im^2/Bal) 1m) ,
| |
| date days Act. (Ci) 1.60E-13 0.02368 l l
| |
| 1/18/90 55 4.00 1 3/8/90 6 2.54 3/14/90 0 2.40 2.49
| |
| /
| |
| i I F_p. hours ,
| |
| ! Source of data VAo mrem rem Sv at a meter l
| |
| *- Tab'e 5.2 Nova Truck Driver 60 0.04 0.0004 0.016892 i Tablo 5.2 Nova Cargo Unioader 1 330 0.33 0.0033 0.139358 i Table 5.2 Nova Carge Unioadct 2 330 0.33 0.0033 0.139358
| |
| ) Table 5.2 Nova Unloading Forklift Operator 50 0.05 0.0005 0.021115 j Table 5.2 Nova Unloading Checker 70 0.07 0.0007 0.029561 Table 5.2 Nova Shipping clerk 230 0.23 0.0023 0.097128 Table 5.2 No /a Shipping Supervisor 230 0.23 0.0023 0.097128 l !
| |
| Table 5.2 Nova Asst. Shipping Clerk 150 0.15 0.0015 0.063345 Table 5.2 Nova Receiving Clerk 400 0.4 0.004 0.168919 Table 5.2 Nova Loading Foridift Operator 200 0.2 0.002 0.084459 Table 5.2 Nova Cargo Loader 1 470 0.47 0.0047 0.19848 l
| |
| Table 5.2 Nova Cargo Loader 2 470 0.47 0.0047 0.19848 Table 5.2 Nova Loading Checker 220 0.22 0.0022 0.092905 i Tcbles 5.3 & 5.4 Covenant Senior Driver 34850 34.85 0.3485 14.71706 Tables 5.3 & 5.4 Covenant Driver Trainee 27560 27.56 0.2756 11.63851
| |
| {' 5600 5.6 0.056 2.364865 Table 5.4 Patriot Operator i Table 5.4 Patriot Warehouseman 1080 1.08' O.0108 0.456081 i Table 5.4 USCS inspector 810 0.81 0.0061 0.342061 Table 5.4 Patriot Truck' Driver 550 0.55 0.0055 0.232264
| |
| } 20 0.02 0.0002 0.008446
| |
| ; Table 5.5 Amersham Rad Tech A
| |
| . Table 5.5 Amersham Rad Safety Specialist 30 0.03 0.0003 0.012669 ;
| |
| i Table 5.5 Amersham Rad Tech B 20 0.02 0.0002 0.008446 Table 5.5 Amersham Rad Safety Officer 15 0.015 0.00015 0.006334 Table 5.5 Amersham Hot Lab Superviw 40 0.04 0.0004 0.016892 Collective 73765 73.765 0.73765 31.15 Average 3073.5 3.07 0.0307 1.30 l
| |
| Standard Deviation 8801.9 8.80 0.0880 3.72
| |
| ) 0.015 0.00015 0.00633 Minimum 15 i Maximum 34850 34.85 0.3485 14.72 No. Individuals 24 24 24 24 Geometric Mean 259.58 0.260 0.00260 0.110 1
| |
| Geometric Standard Deviation 7.87 7.87 7.87 7.87 i
| |
| 1 F-59
| |
| | |
| 1 ,
| |
| i
| |
| :d
| |
| : i l
| |
| i j - Table Task-6-A3. Summary information for 40 accidents involving sealed sources and 231 individuals with known or estimated do5es.
| |
| Whole
| |
| ,,, Sedy Tkne-
| |
| . O E and-f 5 i Pmemny Gamma Feder g (SWh Gamma *Act. (heurs at a Code >- Ima2fBen f9Wh a im) metern ALG78 78 Ir-192 IndRad 25 9.25E+11 1.00E-13 0.148 87.837838 ALG78 78 Ir 192 fndRad 25 9.25E+1# .OE-13 0.148 88.827027 ALG78 78 Ir-192 IndRad 25 9.2SE+11 1.00E-13 0.144 81.001081 ALG78 78 Ir-192 IndRad 25 9.25E+11 1.00E-13 0.148 74.324324 ALG78 78 Ir-192 IndRad 25 9.25E+11 1.00E-13 0.148 270.27027 ALG78 78 k-192 IndRad 25 9.25E+11 1.00E-13 0.148 6.7587568 ALG78 78 Ir-192 IndRad 25 9.25E+11 1.60E 0.148 6.7567568 AUS70 70 Ir 192 IndRad 22 8.14E+11 1.6E-13 0.13024 1.23 AUS70 ' 70 tr-192 indRad 22 8.14E+11 1.6E-13 0.13024 3.53 BAN 85 85 k-192 indRed 50 1.85E+12 1.6E 13 0.296 8.45 CA79 79 fr-192 indRad 28 1.04E+12 1.6E-13 0.16576 0.0120656 CA79 79 fr-192 indRad 28 1.04E+12 1.6E-13 0.16576 0.0241313 CA79 79 fr 192 IndRad 28 1.04E+12 1.6E 13 0.16576 0.0422297 CA79 79 Ir-192 indRad 28 1.04E*12 1.6E 13 0.16576 0.0844506 CA79 79It192 lodRad 28 1.04E+12 1.6E-13 0.16576 0.1025579 CA79 79 Ir-192 IndRad 28 1.04E+12 1.6E-13 0.16576 0.11764 CA79 79 it-102 IndRad 28 1.04E+12 1.6E-13 0.16576 0.1870174 CA79 79 Ir-102 IndRad 28 1.04E+12 1.8E-13 0.18576 0.3819001 CA79 79 Ir-102 indRad 28 1.04E+12 1.6E 13 0.10678 0.3021332 CA79 79 Ir-192 indRad 28 1.04E+12 1.eE-13 0.18678 1.3875483 CA79 79 Ir-192 Instad 28 1.04E+ f 2 1.eE-13 0.10678- 5.2787182 CZE86 08 l-131 Meecal 2.25 8."3E+';0 7.63E-14 C.008351975 0.002 CZEe8 as L131 Medical 2.25 8.33E+10 7.63E-14 0.008361975 0.004 CZE86 GB l-131 Medical 2.25 6.33E+10 7.83E-14 0.008361075 0.008 CZE86 08 L131 Mescal 2.25 8.33E+10 7.63E-14 0.008361975 0.008 CZEOS 08 l-131 Me4cel 2.25 8.33E+10 7.83E 14 0.008361975 ' O.01 CZE06 08 L131 Medical 2.25 8.33E+10 7.63E 14 0.008361975 0.012 CZE86 80 5131 Medcol 2.25 8.33E+10 7.83E 14 0.008361975 0.014 CZESS es 8-131 Medcol 2.25 4.33E+10 7.83E-14 0.008351975 0.016 CZEe8 08 l-131 Medcol 2.25 8.33E*10 7.83E 14 0.008361975 0.018 CZE86 08 L131 Mescal 2.25 8.33E+1J 7.83E-14 0.008361975 0.02 CZE08 08 L131 Medcol 2.25 8.33E+10 7.83E 14 0.008361975 0.022 CZEe8 08 8131 Messel 2.25 8.33E+10 7.83E-14 0.008361975 0.024 CZESS OS l-131 Medical 2.25 8.33E+10 7.83E-14 0.008381975 0.028 CZE86 88 l-131 Meecal 2.25 8.33E+10 7.83E-14 0.008361975 0.029 CZES8 68 l-131 Medical 2.25 8.33E+10 7.63E-14 0.00836i'75 0.03 CZE06 86 6-131 Medical 2.25 8.33E+10 7.63E 14 0.008361975 0.032 CZE73 73 C+40 Tale 2973 1.10E+14 3.7E-13 40.7 0.0025 CZE73 73 Co 80 Tsie 2973 1.10E+14 3.7E-13 40.7 0.0344 FRG84 64 k-192 IndRad 7.8 2.00E+11 1.eE-13 0.048176 32.4 i FRG72 72 Ir-182 IndRed 29.73 1.10E+12 1.6E-13 0.178 1.7 FRG01 81 Ce40 Tale 2505 9.00E+13 3.7E-13 35.52 0.0113 FRG41 diC+40 Tale 2505 9.00E+13 3.7E-13 ' 36.52 0.0086 IN068 64 k-192 IndRad 1.4 5.18E+10 1.sE 13 0.008288 158.05 ISR90 90 Ce40 Sierg 340541 1.26E+16 3.7E-13 4882 0.0032 f
| |
| e F-60
| |
| | |
| 4-J i
| |
| i i
| |
| \
| |
| i Table Task-6-A3 Continued. Summary information for 40 accidents involving sealed sources j and 231 individuals with known or estimated doses.
| |
| i 1
| |
| ! I i Whole Sody Time-3 S E *ad-
| |
| < < Proximdy i
| |
| j f. g- Gamme Factor Cada y g l e
| |
| (SWh Garnme*Act. (haurs at a M Im*2 sad f 9Wh a im) maior)
| |
| ; ITA75 75 Co.80 Store 38000 1.33E*15 3.7E 13 402.M 0.0284 JOHet 50Ce40 Instad 1.75 8.40E+10 3.7E-13 0.0230675 1.04 3 JPN71 71 tr 192 IndRad 5.26 1.96E+11 1.eE-13 0.0311302 3.2113083 j 4- JPN71 71 Ir-182 IndRad 5.20 1.96E*11 1.8E-13 0.0311302 4.1748022 j i JPN71 71 Ir-192 andRad 5.26 1.06E+11 1.0E-13 0.0311382 4.8170794 i JPN71 71 Ir-102 IndRad 5.26 1.96E+11 1.eE-13 0.0311302 8.02Me57 l l
| |
| * JPN71 71Ir192 indRad 5.26 1.95E+11 1.8E-13 0.0311302 16.060931 i
| |
| ; JPN71 71 Ir 102 IndRad 5.26 1.95E+11 1.6E-13 0.0311392 42.711438 i KOR90 90 it-192 indRad 4 1.48E+11 1.00E-13 0.02360 0.0063345 1 KOR90 90 fr-192 IndRad 4 1.48E+ 11 1.60E-13 0.02368 0.00M459 '
| |
| l KOR90 90 Ir 192 IndRad 4 1.48E+11 1.00E-13 0.02368 0.0064459 I KOR90 90 fr-102 IndRad 4 1.48E+11 1.60E-13 0.02368 0.0126689 l KOR90 90 Ir-102 indRad 4 1.48E+11 1.60E-13 0.02368 0.0168919 4 KOR90 90 tr-192 IndRad 4 1.48E+11 1.60E-13 0.02368 0.0160919 KOR90 90 Ir-192 IndRad 4 1.48E+11 1.00E-13 0.02368 0.0211149
| |
| ! KOkt0 90 It 192 IndRad 4 1.40E+11 1.00E-13 0.02308 0.0295608 i KOR90 90 Ir 102 indRad 4 1.40E+11 1.00E-13 0.02388 0.0033446 KOR00 90 Ir 192 IndRad 4 1.40E+11 1.00E 13 0.02388 0.0844505 KORe0 00 Ir-192 lnar.ed 4 1.40E+11 1.00E-13 0.02300 0.0820064 KORSO 80 fr-192 indRad 4 1.40E+11 1.00E 13 0.02300 0.0071284 KOR90 to Br-182 IndRad 4 1.4eE+11 1.00EA 3 0.02388 0.08712M KORSO to Ir-te2 ladRad 4 1.40E+11 1.60E-13 ".02300 0.1383681 KOR90 90 tr-102 indRad 4 1.4eE+11 1.00E-13 0.02300 0.1383581 KORSO 90 it-192 indRad 4 1.4eE*11 1.80E-13 0.02308 0.1880100 KORe0 90 Ir-192 IndRad 4 1.40E+11 1.00E-13 0.023ee 0.10e4797 KOR90 90 ar-192 indtad 4 1.40E+11 1.80E 13 0.02308 0.1084797 KOR90 90 Ir-192 IndRad 4 1.4eE+11 1.00E-13 0 02388 0 3322036 KOR90 90 fr-192 indled 4 1.40E+11 1.80E-13 0.02300 0.3420000 KOR00 to tr-192 instad 4 1.4eE+11 1.00E-13 0.02308 0.4000011 ,
| |
| KORSO 90 tr-182 IndRad 4 1.40E+11 1.00E-13 0.02300 2.3848849 i KOR90 to tr-192 instad 4 1.40E+11 1.00E-13 0.0230s 11.830514 l KOR90 90 Ir-102 indtad 4 1.40E+11 1.00E-13 0.02300 14.717081 i KY78 70 tr-192 instad 78 2.00E+12 1.80E-13 0.48176 2.06 !
| |
| LA78 74 Ir-182 IndRad 100 1.70E+12 1.00E-13 0.002 0.0045 MEXS2 82 Co40 indRad 5 1.8eE+11 3.70E-13 0.00M5 175.31045 MENS 2 82 Ce40 IndRad 5 1. gee +11 3.70E-13 0.00045 808.83258 MEXB2 82 Ce40 indRad 5 1.00E+11 3.70E-13 0.00045 511.32213 MEXe2 82 Co40 IndRad 5 1.86E+11 3.70E-13 0.00045 418.20415 ;
| |
| MEXe2 82 C+40 instad 5 1.85E+11 3.70E-13 0.08845 438.27011 MORS4 M tr.192 IndRad 18.2 0.00E+11 1.00E-13 0.000 10.418887 MORM M Ir-102 IndRad 18.2 8.00E+11 1.00E 0.008 20.041087 MOR$4 M tr 192 indRad 16.2 6.00E+11 1.00E 13 0.006 87.700333 MORM M tr-192 IndRad 18.2 0.00E+11 1.00E-13 0.006 83.88 MORM M tr-192 IndRad 18.2 S.00E+11 1.00E-13 0.000 10e5138 MORS4 84 tr-192 andtad ,16.2 S.00E+11 1.00E-13 0.006 123.3823 i MORM M 1r-192 ladRad 16.2 0.00E+11 1.00E-13 0.006 138.457 l 0.006 158.8383 l MORe4 44 Ir-192 IndRad 16.2 8.00E+11 1.00E 13 1
| |
| F-61 !
| |
| | |
| i Table Task-6-A3 continued. Summary information for 40 accidents involving sealed sources and 231 individuals with known or estimated dose.;.
| |
| l l
| |
| I whole c- 9 Body h
| |
| . o e and-
| |
| [e 2 2 Prommity Gamma Factor g g (SWh Gamme*Act. (hours at a e
| |
| Code >. Ima2/ bed (SWh a imi meter) 1 MOR44 84 Ir-192 enefted 16.2 6.00E+11 1.00E-13 0.096 17T.0292 MORM M lt-192 indRed 16.2 6.00E+11 1.60E-13 0.096 206.0017 MORM 84 Ir-192 IndRad 16.2 6.00E+11 1.60E-13 0.006 260.42 MfT4 74 Co40 Siert 120000 4.44E+15 3.70E 13 1642.8 0.0025 KTT7 77 Co40 8 tort 500000 1.85E+16 3.70E-13 6845 0.0003088 NOR82 82 Co 60 Ster 9 65720 2.43E+15 3.70E-13 899.7068 0.0445 NYS3 83 Co40 indRed 25 9.25E+11 3.70E-13 0.34225 0.0015 NY83 83 Co40 indRed 25 9.25E+11 3.70E-13 0.34225 0.002 NY83 83 Co-60 indRad 25 9.25E+11 3.70E-13 0.34225 0.0025 NY83 83 Co-60 indRad 25 9.2SE+11 3.70E-13 0.34225 0.003 NY83 83 Co-60 indRad 25 9.25E+11 3.70E-13 0.34225 0.0035 NY83 83 Co40 indRad 25 9.25E+11 3.70E-13 0.34225 0.004 NY83 83 Co-60 IndRad 25 9.2SE+11 3.70E-13 0.34225 0.0045 NY83 83 Co40 indRad 25 9.2SE+11 3.70E-13 0.34225 0.005 NY83 83 Co40 indRad 25 9.25E+11 3.70E 13 0.34225 0.0055 NY83 83 Co40 IndRad 25 9.25E+11 3.70E-13 0.34225 0.0058 PA92 92 Ir-192 Brachy 3.7 1.37E+11 1.00E-13 0.021904 0.0002575 PA92 92 Ir-192 Brachy 3.7 1.37E+11 1.00E-13 0.021904 0.0002575 PA92 92 Ir-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 0.0006757 PA92 92 Ir-192 Brachy 3.7 1.37E+11 1.00E-13 0.021904 0.0008757 PA92 92 Ir-102 Brachy 3.7 1.37E+11 1.00E-13 0.021904 0.0006757 PA92 92 tr-192 Brachy 3.7 1.37E+11 1.00E-13 0.021904 0.0005757 PAE2 92 Ir-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 0.0008757 PA92 92 fr-192 Brachy 3.7 1.37E+11 1.60E-1.1 0.C21904 0.0008757 PA92 92 It-192 Brachy 3.7 1.37E+11 1.00E-13 a.021904 0.0008757 PA92 92Ir192 Brechy 3.7 1.37E+11 1.60E 13 0.021904 0.0008757 PA92 92Ir182 Brachy 3.7 1.37E+11 1.80E-13 0.021904 0.0008757 PA32 92 Ir 192 Brachy 3.7 1.37E+11 1.90E-13 0.021904 0.0008757 PA92 92 lt-192 Bretty 3.7 1.37E+11 1.90E 13 0.021904 0.0005757 PA92 92 Ir-192 Brachy 3.7 1.37E+11 1.00E-13 0.021904 0.0029875 PA92 92 It-192 Brachy 3.7 1.37E*11 1.00E-13 0.021904 0.0029875 PA92 92 Ir-192 Brachy 3.7 1.37E+11 1.00E-13 0.021904 0.0029875 PA32 92 Ir-182 Brachy 3.7 1.37E+11 1.00E 13 0.021904 0.0043371 PA92 92 Ir-192 Brachy 3.7 1.37E+11 1.00E-13 0.021904 0.0043371 PA92 92 Ir-192 Brachy 3.7 1.37E*11 1.00E-13 0.021904 0.0043371 PA92 92 Ir-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 0.0043371 j PAB2 92 Ir-192 Brachy 3.7 1.37E+11 1.00E 13 0.021904 0.0043371 PA92 92 Ir-192 Brachy 3.7 1.37E+11 1.60E 13 0.021904 v.0043371 PA92 92 Ir-192 Brachy 3.7 1.37E*11 1.80E-13 0.021904 0.0043371 PA92 92 It-192 Brachy 3.7 1.37E+11 1.00E-13 0.021904 0.0070763 PA92 92 Ir 192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 0.015294 PA02 92 Ir-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 0.015294 PA92 92 Ir-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 0.0159798 PA92 92 Ir-192 Brachy 3.7 1.37E+11 ' 1.00E-13 ' O.021904 0.0264792 ,
| |
| PA92 92 tr-192 Brachy 3.7 1.37E+11 1.60E 13 0.021904 0.0602191 !
| |
| PA92 97Ir192 Brachy 3.7 1.37E+11 1.8CE-13 0.021904 0.0547845 F-62
| |
| | |
| 4 i
| |
| Table Task-6-A3 continued. Summary information for 40 accidents involving sealed sources I and 231 individuals with known or estimated doses.
| |
| 1 whol.
| |
| ._, - Body Tune- I
| |
| * E. d ami-y <
| |
| g
| |
| < Proximity
| |
| . . a Gamma Fedor code gf
| |
| >. f M E
| |
| h J
| |
| h (SWh Gamma *Act. (hours at a
| |
| # im 2/a Ban fswh a imi molari PA92 92 tr-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 0.0639153 4
| |
| PAB2 92 Ir-192 Brachy 3.7 1.37E+11 1.00E-13 0.0219C4 0.0776114 PA92 92 tr-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 0.1529401 PA92 92 tr-182 Brachy 3.7 1.37E+11 1.60E-13 0.021904 0.1529401 PA92 92 Ir-102 Brachy 3.7 1.37E+11 1.00E-13 0.021904 0.175767 PA92 92 fr-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 0.2061267 PA92 92 Ir 192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 0.2077246 j 3.7 1.37E+11 PA92 92 fr-192 Brachy 1.60E 13 0.021904 0.2077246 PA92 92 Ir-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 0.2259861 j PA92 92 Ir-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 0.2396822 j PA92 92 Ir-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 0.2876187 PA92 92 fr-192 Brachy 3 7 1.37E+11 1.60E-13 0.021904 0.3013148 PA92 92 fr-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 0.3743608 PA92 92 fr 192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 0.3971877 PA92 92 fr-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 0.607195 PA92 92 Ir 192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 0.6117604 PA92 92 Ir-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 0.84916 PA92 92 fr-192 Brachy 3.7 1.37E+11 1.00E 13 0.021904 0.8834003
| |
| . PA92 92 It-192 Brachy 3.7 1.37E+11 1.00E-13 0.021904 0.0002484 PA92 92 tr 192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 0.9062272 PA92 92 Ir-192 Brachy 3.7 1.37E*11 1.00E-13 0.r'21904 0.9153579 l PA92 92 Ir-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 0.0*99233 l
| |
| ! PA92 92 tr-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 1.0642768 ,
| |
| PA92 92 tr-192 Brachy 3.7 1.37E+11 1.00E-13 0.021904 1.2189554 PA92 92 Ir 192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 1.2189554 PA92 92 Ir-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 1.3011322 PA92 92Ir192 Brachy 3.7 1.37E+11 1.60E-13 0.421904 1.3513514 PA92 92 It-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 1.3696129 PA92 92 Ir-192 Brachy 3.7 1.37E+11 1.80E-13 0.021904 1.4677664 i PA92 92 Ir-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 1.529401 l PA92 92 Ir-192 Brachy 3.7 1.37E+11 1.00E-13 0.021904 1.6207005 j PA92 92 Ir-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 1.6252730 l I PA92 92 fr-192 Brachy 3.7 1.37E+11 1.80E-13 0.021904 1.6823411 l PA92 92 it 192 Braey 3.7 1.37E+11 ,1.80E-13 0.021904 2.1754018 PA92 92 Ir 192 Brachy 3.7 1.37E+11 1.00E-13 0.021904 2.306515
| |
| , PA92 92 Ir*92 Brachy 3.7 1.37E+11 1.00E-13 0.021904 2.305515 PA92 92 Ir-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 2.4447589 PA92 92 Ir 192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 1 6547063 PA92 92 It 192 Brachy 3.7 1.37E*11 1.60E-13 0.021904 2.8259679 PA92 92 tr 192 Brady 3.7 1.37E*11 1.00E-13 0.021904 1 876187 PA92 92 tr-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 1 876187 PA92 92 tr-192 Brachy 3.7 1.37E+11 1.00E-13 0.021904 2.876167 PA92 92 tr-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 2.876187 PA92 92 tr 192 Brachy 3.7 1.37E+11 1.60E 13 0.021904 3.264244 PA92 92tr192 Brady 3.7 1.37E+11 1.00E-13 0.021904 3.264244 PA92 92 tr-192 Brachy 3.7 1.37E+11 1.00E-13 0.021904 3.3829438 5 PA92 92 tr-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 3.7801315 i i 4
| |
| F-63 ,
| |
| | |
| i l l
| |
| 1 1
| |
| l Table Task-6-A3 continued. Summary information for 40 accidents involving scaled sources
| |
| ; and 231 individuals with known or estirnW doses.
| |
| 8 Whole
| |
| - Body Time- l J g S k # I p < Proiemity
| |
| .I g Gamma Factor g -
| |
| (SWh Gamme*Ad (hours at e A52 92 Ir-192 Brachy 3.7 1.37E*11 i PAA2 92 Ir-192 Brachy 3.7 1.37E+11- 1.60E-13 0.021904 5.1132213
| |
| ; PAR 2 32 Ir-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 5.1546751
| |
| + PA92 92 Ir 102 trachy 3.7. 1.37E+11 1.60E-13 0.021904 5.7296471 PAA2 92 Ir-192 Brachy 3.7 1.37E+11 1.00E-13 0.021004 5.7296471 PAA2 92 Ir 192 Brachy 3.7 1.37E+11 1.00E 13 0.021904 5.9665332 ;
| |
| PA02 92 Ir-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 6.5284879 'l' PA92 92 Ir-192 Brachy 3.7 1.37E+11 1.60E-13 0.021904 8.628561 i PRC63 63 Co.80 indRad
| |
| * 10 3.70E+1) 3.70E-13 0.1369 14.000204 PRC63 63 Co 60 IndRad 10 3.70E+11 3.70E-13 0.1369 29.218400 l PRC63 63 Co40 IndRad 10 3.70E+11 3.70E-13 0.1369 43.827611 i PRC63 63 Co 60 IndRad 10 3.70E+11 3.70E-13 0.1369 58.436815 ;
| |
| PRC63 63 Co40 wfRad 10 3.70E+11 3.70E-13 0.1369 292.18400-PRC63 63 Co40 IndRad to 3.70E+11 3.70E-13 0.1369 584.36815 PRC80 80Co40 Stent' 53000 1.90E+15 3.70E-t3 725.57 1.89 PRCSS 85 Cs-137 IndRad? 10 3.70E+11 9.25E-14 0.034225 3.35 j PRC86 asCo40 (1) 6888 156E+14 3.70E-13 94.29672 3.83 j PRC87 87 Ce40 8terd - 89000 3 29E+15 3.70E 13 1218.41 7.3 PRCO2 92 Co40 Empt 12 4.44E+11 3.70E-13 0.1642s 8.143138 PRC92 92 o.40 Eapt 12 4.44E+11 3.70E-13 0.18428 .8.877649 PRC92 92 Co40 Empt 12 4.44E411 3.70E-13 0.16428 9.118739 ,
| |
| PRC92 > 92 Co40 Empt 12 4.44E+1' 3.70E 13 0.1642s 9.521521 l I
| |
| PRC92 92 Co40 Empt 12 4.44E+11 3.70E-13 0.16428 9.912457 PRC92 92 Co40 Empt 12 4.44E+11 3.70E-13 0.14428 10.31026 j PRC02 82 Ce40 Empt 12 4.44E+1. 3.70E-13 0.16428 10.73358 PRC82 92 Co40 Empt 12 4.44E+11 170E-13 0.16428 11.20769 PRC02 92 Co40 Empt 12 4.44E+11 3.70E-13 0.16428, 11.77738 PRC02 92 Ce40 Empt 12 4.44E+11 3.7.4-13 0.16428 12.55044 PRC02 92 Ce40 Empt 12 4.44E+11 3.70E 13 0.16428 14 PRCO2 82 Ce40 Esqst 12 4.44E+11 3.70E-13 0.1842s 80.87168 ;
| |
| PRCE2 92 Ce40 Eaget 12 4.44E+11 3.70E-13 0.16428 91.30752 PRC02 92 Co40 East 12 4.44E+11 3.70E 13 0.16428 121.7434 alAF77 77 Ir-192 IndRed 8.78 2.50E*11 1.00E-13 0.04 15 '
| |
| -SAF77 77 Ir-IS2 Insted 8.78 150E+11 1.00E-13 0.04 4.25 SAF77 77 Ir-192 ladRed 6.76 2.50E+11 1.00E-13 0.04 29 10000 8.esE+14 3.70E-13 0.013 i SALAG esCo40 Store 248.42 SALas seCode start 10000 8.esE+14 3.70E-13 248.42 0.010 ,
| |
| 1 SALSD atCe40 Stord 10000 8.68E+14 3.70E 13 248.42 0.034 SCO90 GB Ir-192 IndRed 25 9.25E+11 1.00E-13 0.140 4.05 i TN71 71 Ce40 Empt 7700 2.86E+14 3.70E-13 105.413 0.0247 TRK78 78Co40 Tale 2200 0.3eE+13 3.70E-13 30.0304 0.0001118 l TX72 72 Co-137 indRed 4 1.40F+11 9.25E-14 0.01389 730 UKTT 77 Ir-192 IndRad 21.6 0.00E+11 1.00E-13 0.128 0.78 UK81 81 Co 137 Brachy 0.12 4.44E*00 9.2SE-14 0.0004107 0.407 WA76 76 Am-241 Dhse 343 1.27E+13 7.57E-15 0.008106142 0.062 Wl61 61 Co40 Empt 200 7.40E+12 3.70E-13 2.738 0.913 F-64
| |
| | |
| 1 Table Task-6-A3 continued. Summary information for 40 accidents involving sealed sources and 231 individuals with known or estimated doses.
| |
| wnw.
| |
| - - Body Time-t E 5 "
| |
| g < < Proemdy g g 4 Gamma Factor Code g
| |
| B S
| |
| j 4
| |
| f a (Se Gamma *Act. (hours at a
| |
| < Im 2AleD (SvAi a im) meter)
| |
| Muumum 0.0001118 Manmum 730 Avere0e 30 std Dev 100 FW 0.37 GSD 42.53 Median 0.46 Mode 6.76E-04 Number 231 i
| |
| i i
| |
| i l
| |
| i l
| |
| 4 l
| |
| l l
| |
| 4 F-65
| |
| | |
| APPENDIX G: PRELIMINARY RISK ANALYSIS FOR SELECTED SOURCES A ft" im,lementation of the risk analysis describ :d in Sect:ca . vf the Task 7 Final Report is beyond the scope of work of the current project. However, a sample risk analysis is given below.
| |
| G1.0 RADIOLOGICAL PROPERTIES OF RADIONUCLIDES REGISTERED BETWEEN 1987 AND 1992 Data provided to PNL by S. L. Baggett (Baggett 1993) included 24 valid radionuclide names.
| |
| Table G-1 lists those radionuclides as row labels in column 1, in order of increasing atomic !
| |
| pumher. The specific dose equivalent rate constant! for each nuclide (or chain, in the case of 226 ,
| |
| Ra) is given in the next column in units of rems per hour at I m from 1 Ci (Unger and !
| |
| Tmbey,1981). 'Ihe next two columns list the most restrictive ALI values for inhalation and I ingestion, respectively. There is no ALI for s5Kr. The half-life for each nuclide is given in the center column.
| |
| Since distributions of source st engths and numbers of sources in service with activities below 20 millicuries (the limit of the ORAU report) are not available, a source strength of 20 mci !
| |
| was arbitnrily chosen as a reference value in Table G-1. Also, for simplicity, the " worst plausible case" reference value for the time-and-proximity factors, F,, of 1000 hours at one meter was chosen (730 was the highest observed) for use in Table G-1. Similarly, a " worst i plausible case" reference value of the fraction-taken-in, F,, of 10d was chosen for use in Table G-1.
| |
| Column 7 shows the dose equivalent in rems for extemal exposure for 1000 hours at 1 meter away from an unshielded 20 mci source. Columns 8 and 9 show the committed effective 1 dose equivalent in rems for intakes of 104 of a 20 mci source, i.e.,2 Ci, for inhalation and ingestion, respectively. Note that these columru can be interpreted as doses in millirems for 1 hour at a meter and 104 fraction-taken-in, both more central values from the PNL analysis presented in Appaadir F, Task 6. :
| |
| 'Ibe sixth column of Table G-1 gives the ratio of the value in column 7 divided by the value i in column 9. 'Ibe column 6 values are indapaadant of source strength. Column 6 values are l a ratio of external exposure hazard (dose) to inimire hazard (dose) under the 1000 hours at a meter and 104 scenarios. Figure G-1 is a bar plot of Column 6 values. For s5 Kr, there is no ALI even though it poses an external hazard; for 63Ni, 55pe, 3H,210Bi, I"Ru,14C, and
| |
| .I Although it would be desirable to use effective dose equivalent rate constants, none have been published in the peer-reviewed literature. Calculations of such constants, based on the methods of the ICRP and ICRU, as shown in Appendix A of the Task 7 Final Report, lead to unrealistically high numbers for low-energy photon emitters if " bare" sources are assumed (e.g., 241Am).
| |
| Further research is naaded in this area. Thus, the work of Unger and Trubey (1981), based on ANSI /ANS-6.1.1-1977, has been used for dose eouivalent rate consants.
| |
| G-1
| |
| | |
| Table G-1. Summary of radiological data needed for risk analysis for radionuclides listed the NMSS General License Database System from 1987 through 1992.
| |
| Dose from Dose Dose
| |
| ' Dose 1000 h at 1 from 1000 from from Most m / Dose hours at 1 inhalation Ingestion Most from m from 20 of 10^-4 of 10^-4 Restridive Restrictive ingestion of rem /h at 1 inhalation ingestion Half-Life mci of 20 mci of 20 mci 1V-4 of Source Source Nucfde m from 1 Cl Source H-3 ALI(uct) ALI(uci) (years) Source (rems) 0 3000 3000 (rems) (rems) 12.4 0 C-f4 0 90 0.0033 0.0033 90 5,730 0 Sc-46 1.17 200 0.11 0.11 900 0.230 2101 71- 4 4 0.145 23 0.05 0.011 6 300 47.3 Fe-55 86.8 2.9 1.7 0 2000 0.033 9000 2.7n 0 Co-60 ' 1.37 200 0.0050 0.0011 500 5.27 1370 27 Ni-63 0 0.05 0.020
| |
| '2000 9000 96.0 0 Kr-85 0.00157 - -
| |
| 0.0050 0.0011 10.7 -
| |
| 0.031 St-90 0 4 30 29.1 0 Ru-f06 0 10 2.5 0.33 200 1.01 0 Cd-fo9 0.184 40 1.0 0.050 300 1.30 111 3.7
| |
| - /,-129 0.126 0.25 0.033 9 5 1.57E+7 Cs-137 1.26 2.5 1.1 0.382 200 100 2.0 30.0 76.4 7.6 Ba-f33 0.455 700 0.05 0.10 2000 10.7 1822 Pm-147 9.1 0.014 0.0050 2.676E-06 100 4000 2.62 0.0214 5.35E-05 Eu-152 0.744 20 0.10 0.0025 800 13.3 1191 15 77-204 0.00112 2000 0.50 0.013 2000 3.78 4.46 Po-2fo 5.269E-06 0.022 0.0050 0.0050 0.6 3 0.379 81-210 0 3.16E-5 1.055-04 17 3.3 200 800 0.0137 0 Ra-226 1.43 0.05 0.013' 0.6 2 1.600 Pu-238 5.71 29 17 50 0.0790 0.007 0.9 87.7 0.142 1.6 1,429 Am-241 0.314 0.006 0.8 11 432 0.502 6.3 1,667 Cm-244 0.0644 0.01 13 1 18.1 0.129 1.3 1,000 Cf-252 0.0418 0.002 10 2 2.64 0.167 0.84 5.000 5._0 "Sr the specific dose equivalent rate constant is 0, even though, in the 90Sr case
| |
| #Y, of there is an external irradiation hazard. For sources with Column 6 ratios greater than 1, external exposure dami=*~ risk; for sources whose values are less than 1, intake dominates risk. Note that these are based on ingestion ALIs; values ba.W on inhalation ALIs, in general, are less than or equal to the Column 6 values. .
| |
| Figure G-2 shows Column 9 values plotted as a function of Column 7 values, illustrating the .
| |
| very wide range of doses associated with 20 mci sources under the exposure scenarios described above.
| |
| G-2 i
| |
| | |
| So-46 Ba-133 -MMM Co-Go MMMM I
| |
| Eu-152 MM M
| |
| ; cd.tos t Dose from 1000 h at 1 m / Dose from Ingestion mm 3 i
| |
| T1-44 of 10^-4 of Source men cs-137 m-Ra-228 j T1-204 m
| |
| ! M2s a Extemal exposure j ,
| |
| dominates risk I
| |
| cr-252 i
| |
| Pu-238 l Intake dominates risk l m m
| |
| ; Cm-244 Pm-147 m
| |
| mm Po-210 Kr-as MMMMm NI-83 r.-ss H-3 81-21o Ru-106 c-14
| |
| * St-90 f
| |
| i 0.000001 0.00001 0.0001 0.001 0.01 0.1 1 10 100 1000 10000 Figure G-1. Ratios of (Dose Equivalent from 1000 h at 1 m) divided by (Committed Effective Dose Equivalent from intake of 10-4 of a source. These dimensionless ratios are independent of source strength.
| |
| l G-3 I
| |
| _ _, _ _ . _ . , _ _ _ . _ . _ . _ - . _ _ . _ . _ . . _ . . . _ . _ _ . . _ , _ . , , _ _ . . ~ _ , . , . _ . . . , _ , _ . , _ _ , . , , _
| |
| | |
| 7 E 100 5
| |
| E 10
| |
| * f i 4 7
| |
| o 1
| |
| Co i, 5 0.1 :
| |
| I '
| |
| s,
| |
| .E .
| |
| 0.01 " *^
| |
| E
| |
| = "
| |
| ! 0.001 "
| |
| U 1E-05 0.0001 0.001 0.01 0.1 1 10 100 Dose from 1000 h @1m from 20 mci (rems)
| |
| Figure G-2. Committed effective dose equivalent from ingestion of 10 d of a 20 mci source (vertical axis) as a function dose equivalent from external exposure for 1000 hours at I meter from an unshielded 20 mci source. The relative doses are seen to be widely variable.
| |
| G2.0 NUMBER OF DEVICES ovf Numbers of sources are given in Table Task-3-1B for by Device Code. For this analysis,
| |
| /
| |
| Device Code D, Gamma Gauges, is chosen for radionuclide 137Cs, which includes 2493 sources with average activity 883 mci. Table G-2 shows a possible (hy;Weil) distribution of numbers of sources and activities to give the correct total and average. It would be best to have exact numbers from a full database.
| |
| S G-4
| |
| | |
| 3
| |
| ; i i Table G-2. Possible (hypothetical) distribution of numbers of sources 137 Cs and acti gamma gauges for a total of 2493 sources and an average of 883 mci per source, in Table Task-3-1B.
| |
| 2 Activity (mci) Number 4
| |
| 1 500
| |
| )
| |
| 10 500 t
| |
| 100 559 1000 800 i
| |
| 10,000 134 TOTAL 2493 G3.0 RATES OF DEVICE INVOLVEMENT G3.1 RATES OF INCIDENTS Table Task-3-8 shows that there were 6 incidents in ORAU Category B (Regula 10.10 Device Code D) over a 10 year period, 5 incidents 137 involving Cs. Assnming that one out of two incidents is reported, this gives a rate of (6 incidents reported) x (2 incide c4 per incident reported) + (10 years) = 1.2 incidents per year fo'r all Device Code ,
| |
| D devices. Using the total number of Device Code D sources in Table Task-3-2, i and correcting for the fraction of Device Code D sources that are 137 Cs, one calculates (1.2 incidents per year) x (2493 / 24,679) = 0.121 incident per year D devices.
| |
| for 137 Cs Device Code '
| |
| One could argue from Table Task-3-8 that the rate for cesium sources is predomin is, mium sources account for S/6 of all incidents. Using this logic would yield a rate of (1.2 incidents per year) x (5/6) = 1.0 incident per year for 137 This analysis uses the latter value. Cs Device Code D devices.
| |
| G3.2 NUMBERS OF DEVICES PER INCIDENT Table Task-3-7 shows 9 devices were involved in the 6 incidents listed in Table Ta for a rate of 1.5 device per incident.
| |
| G3.3 RATE OF DEVICE INVOLVEMENT IN INCIDENTS The rate of device involvement in incidents is the product of the number of inciden year and the number of devices per incident, (1.0 incident per year) x (1.5 devices per G-5
| |
| | |
| - . - - - - - . - . - _ . - ...-.. - - - - .--=_- -.-. -. - - _- _ ..-~
| |
| i
| |
| 'I
| |
| } incident) == 1.5 devices involved in incidents per year.
| |
| G4.0 SEVERITY OF INCIDENTS 3
| |
| .'Ihere are several aspects to severity.
| |
| ; G4.1 FRACTION OF INCIDENTS RESULTING IN LEAKING SOURCES In Table Task-3-6,19 out of 114 incidents resulted in leaking sourtes although no ,
| |
| :,. these involved Device Code D (ORAU Category B) gamma gauge sou,rce that such sources are made to withstand leaking. Ahernatively, Table Task-3-7 sh 75 of 300 devices involved in incidente leaked, primarily static elimiantors and ORAU e
| |
| 4 Category L sources. For this value, 0.125 of incidents result in a leaking source G4.2 1 NUMBER OF PERSONS INVOLVED IN INCIDENT
| |
| {
| |
| Most incidents involve 1 person, but more may be involved. For intakes, Table 3.1.1 1
| |
| Task 6 showed the distribution of numbers of persons involved in accidents, as sum in Table G-3. 'Ihere are 4364 persons involved in the 60 incidents, using 77 per
| |
| ; of the value of 20 in Table 3.3.1) for Goiania.
| |
| i Due to the difficulty of incorporating different numbers of persons with different fraction i taken-in and different time-and-proximity factors, the accompanying analysis co 1-person incidents.
| |
| i i
| |
| l 4
| |
| }
| |
| l 1
| |
| l f
| |
| f e
| |
| e i
| |
| 4 4
| |
| l
| |
| , ~
| |
| 4 G-6
| |
| | |
| i i
| |
| i
| |
| ! 4 4
| |
| Table G-3. The distribution of numbers of persons involved in incidents.
| |
| : Number of Persons Frequency of l Involved in Incident Occurrwce j
| |
| . 1 34 :
| |
| i \
| |
| 2 9 l
| |
| j 3 3 4
| |
| 4 1 4
| |
| ! 5 3 ;
| |
| 6 2 11 1 16 1 ,
| |
| 22 1 24 1
| |
| . 28 1 77 1 94 1 4000 1 G4.3 DISTRIBUTION OF FRACTIONS-TAKEN-IM These data are also taken from Table 3.3.1 of the Task 6 Report. The factions-taken-in are 4355 values, *==W from Table 3.3.1, with the omission of the Swiss accident of 1985 (which involved an Insealed 3H source and a, very high fraction-taken-in, namely,0.02).
| |
| 'Ibe average fraction-taken-in was 1.34 x 10-7, with a standard deviation of 4.17 x 104 .
| |
| 'Ibe latter two persmis were used to define a loena-at distribution.
| |
| G4.4 DISTRIBUTION OF TIME-AND-PROXIMITY FACTORS
| |
| 'Ibese factors were modeled using a lognormal distribution with mean of 30 hours at 1 m and a senadard deviation of 100 hours at 1 m, based on the results presented in Table Task-6-A3.
| |
| G4.5 DISTRIBUTION OF SOURCE REMOVAL FROM SHIELD A scenario of the source removal from the shield was ammad to occur in 0.5 of incidents.
| |
| This is an arbitrary figure that is probably an overestimate.
| |
| G-7
| |
| | |
| t 4
| |
| ; G5.0 PRELIMINARY PROBABILISTIC RISK ANALYSIS RESULTS f ils output of a 1000-trial Crystal Ball simulation is attached at the end of th 1
| |
| i External dose equivalent per year was modeled as (Incilient Rate [ devices
| |
| } Activity) x (Probability ofRemova!from Shield) x (Time-and-Proximity Fact i Dose Equivalent Rate Constant), where the factors in italics were sampled f 4
| |
| these sources have activities of 10,000 millicunes, ;
| |
| j maximum of 1056 rems. Had the source activities been limited to 20 mci, the hi would have been 20/883 as large, or 24 rems. i i
| |
| i i
| |
| Committed effective dose equivalent (CEDE) from ingestion intakes was modeled a (Incident Rate [deviceslycat}) x (Source Activity) x (Probability ofLeakage) x (Fra l Taken-In) x (5 rems CEDE /ALI) x (1 ALI/0.1 mci), where the factors in italics were sampled from distributions described above. The mean dose was 0.4 mrem with a!
| |
| of 154 mrem. Had the source activities been limited to 20 mci, the highest dose woI been 20/883 as large, or 3.5 millirems.
| |
| i G6.0 DISCUSSION This analysis is preliminary and was made to demonstrate the proof-of-principle. Ti mathsh should be reviewed and refined, and calculations for extremity doses non stoc effects, and collective doses should be made for each nuclide ana cach Device Cod! ,
| |
| Table Task-3-1. It should be detennined whether there is a need to res source strengths less than or equal to 20 mci as was done m the scope of work for the ORAU Report. There is a need for detailed data on numbers of devices by sourc Regulatory Guide 10.10 Device Code is not adequate), isotope (s), dves placed in activities, and design. These data should be available for each of the 500,000 or so sour noW in use.
| |
| It was discovered that incorporating realistic scenarios (e.g., different fractions-taken-in a different time-and-proximity factors for each individual in multiple-person incidents) is difficult than expected, the remaining funds did allow us to complete this extra task.
| |
| G7.0 ADDITIONAL REFERENCES FOR APPENDIX G Baggett, S. L. 1993. Letter to D. J. Strom of Pacific Northwest laboratory dated November 2,1993. Washington, DC: Scaled Source Safety Section, U.S. Nuclear Regulatory Comminion.
| |
| Unger, L. M., and D. K. Trubey 1981. Specific Gamma-Ray Dose Constantsfor Nuclides important to Dosimetry and Radi l o ogical Assessment. ORNL RSIC-45. United States G-8
| |
| | |
| . . . .- -. - . . - . _. -. .. -~
| |
| G8.0 SAMPLE PROBABILISTIC RISK ANALYSIS: CRYSTAL BALL REPORT Forecast: Extemal Dose i
| |
| Cell: C13
| |
| . Summary:
| |
| Display Range is from 0.00 to 150.00 roms Entire Range is from 0.00 to 1,056.46 rems After 1,000 Trials, the Std. Error of the Mean is 1.53 Statistics:
| |
| Value Trials Mean 1000 6.81 E + 00
| |
| , Median U.00E + 00 Mode 0.00E + 00 Standard Deviation 4.84E + 01 Variance 2.34E + 03 Skewness 1.44E + 01 Kurtosis 2.62E + 02 Coeff. of Variability 1 7.10E + 00
| |
| . Range Minimum 0.00E + 00 Range Maximum 1.06E + 03 Range WNith 1.06E + 03 Mean Std. Error 1.53E + 00 Forecast Collective Extemal Dose Cell C13 Frequency Chart
| |
| .sst . M1 Triais Shown 843
| |
| .ssa s32 N 3 5 .42s 421 .Em
| |
| .E E
| |
| [ .213 210 k
| |
| .000 7 o 0.00 37.50 75.00 112.50 150.00 rems Page G2-1
| |
| | |
| Appendix G J
| |
| j Forecast: Collective Extemal Dose (cont'd)
| |
| Cell: C13 i
| |
| Fercanuies:
| |
| 3, Percentde Inmi
| |
| ^
| |
| 0.0%
| |
| 0.00
| |
| { 2.5 %
| |
| 0.00 5.0%
| |
| ; 0.00 50.0 %
| |
| l 0.00-95.0 %
| |
| 21.48
| |
| ! 97.5 %
| |
| 50.46 100.0 % 1,056.46 i End of Forecast I
| |
| i Page G2-2
| |
| | |
| .- =- - _ _ . .. . . . ._ - .. .. . .-. -
| |
| l l
| |
| l Append;x G l l
| |
| Forecast: Intake Dose Cell: C14 Summary: j Display Range is from 0.OOE +0 to 1.50E-2 rems Entire Range is from 0.OOE+0 to 1.54E-1 rems After 1,000 Trials, the Std. Error of the Mean is 1.74E-4 Statistics: Value Trials 1000 I Mean 3.99E-04 Median O.OOE+00 Mode O.OOE + 00 Standard Deviation 5.51 E-03 )
| |
| Verlance 3.04E-05 Skewness 2.33E + 01 Kurtosis 6.22E + O2 Coeff. of Variability 1.38E + 01 Range Minimum 0.OOE + 00 Range Maximum 1.54E-01 Range Width 1.54E-01 Mean Std. Bror 1.74E-04 Forecast: Collective intake Dose Cell C14 Frequency Chart M3 Trials Shown
| |
| .965< 958
| |
| .724 710 I 5 .4a2 479 S
| |
| .E E
| |
| .241 230 i
| |
| .000 0
| |
| > 1 0.00E + 0 3.75E 3 7.50E-3 1.13E-2 1.50E-2 rems I j
| |
| Page G2-3 i . . - .
| |
| | |
| Y Appendix G i
| |
| Forecast: Collective intake Dose (cont'd) Cell: C14 '
| |
| Percentiles:
| |
| i Percantile rerns 0.0% 0.00E + 00 2.5 % 0.00E + 00 5.0% 0.00E + 00 >
| |
| 50.0 % 0.00E + 00 f 95.0 % 4.95E-05 97.5 % 6.15E-04 100.0 % 1.54E-01 4
| |
| \
| |
| End of Forecast l
| |
| l I
| |
| l i
| |
| l l
| |
| l l
| |
| l Page G2-4
| |
| ~
| |
| | |
| Appendix G Assumotions 1
| |
| Assumption: Time-and-Proximity Cell: C9 Lognormal distribution with parameters: "--
| |
| ; Mean 30.00 Standard Dev. 100.00 3
| |
| Selected range is from 0.00 to +1nfinity k-Mean value in simulation was 33.57 " " ' ' **" '"
| |
| Assumption: Fraction-Taken-in Cell: C8 Lognormal distribution with parameters:
| |
| Mean 4 -19.26 (log space)
| |
| Standard Dev. 2.62 (log space) 4 i . Selected range is from -Infinity to +1nfinity Mean value in simulation was 0.00 1
| |
| j
| |
| : l. '
| |
| mm
| |
| )
| |
| Assumption: Probability of Leakage 4 Cell: C7
| |
| ' Custom distribution with parameters:
| |
| Single point Relative Prob.
| |
| 0.00 Single point 0.875000 I 1.00 Total Relative Probability 0.125000 1.000000 Mean value in simulation was 0.14 1
| |
| l 4
| |
| A i
| |
| Page G2-5
| |
| | |
| f Appendix G t-Assumption: Probability of Leakage (cont'd) a Cell: C7 9
| |
| ===. .,w m
| |
| , m
| |
| = I
| |
| .f l-l Assumption: Device Activity Cell: C6 I
| |
| Custam distribution with parameters:
| |
| Relative Prob.
| |
| i Single point 1.00 i
| |
| Single point 0.200562 10.00 0.200562 Single point 100.00 Single point 0.224228 1,000.00
| |
| ; Single point 0.320899 10,000.00 0.053751
| |
| } Total Relative Probability 4
| |
| 1.000000 Mean value in simulation was 764.28 w,
| |
| s.
| |
| m m I A- -
| |
| ,Ca.: ProbabWty of Removal from Shield Cell: C11 Custom distribution with parameters:
| |
| R=lative Prob.
| |
| Single point 0.00 Single point 0.500000 1.00 0.500000 Total Relative Probability 1.000000 Mean value in simulation was 0.47 I
| |
| l li Page G2-6 i
| |
| _ 1
| |
| | |
| App:;ndix G j Assumption: Probability of Removal from Shield (cont'd) Cell: C11 e nmey.e w w sa w
| |
| .soo ,
| |
| J7S <
| |
| .no .
| |
| . 26 .
| |
| me $
| |
| ... .,m .. ,,. ...
| |
| l l
| |
| Assumption: Number of Persons . Coll: C10 Custom distribution with parameters: Relative Prob.
| |
| Single point 1.00 1.000000 Total Relative Probability 1.000000 Mean value in simulation wa's 1.00
| |
| .96
| |
| ~
| |
| I
| |
| .. .. .u.
| |
| t End of Assumptions Page G2-7
| |
| ,}}
| |