Evaluation of Holdup Measurement Models by Use of
Facility Simulation
Laniece Miller1 (Physics), Alan Rossner2 (Biology)
Radioactive or nuclear material is highly regulated, including the residue material left behind in piping
and process equipment. This residue material is termed “holdup” and must be accounted for when an
inventory of radioactive material is conducted per the NRC and DOE. Taking a holdup inventory of
equipment and piping can be very time consuming and costly to building owners. Many surveys for
holdup required weeks of work by a team of trained individuals.
The holdup residue can result in the build up of radiation emitting material that will expose workers in
that area. Radiation can emit four main types of radiation. Most nuclides will emit either alpha or beta
particles where are small particles with mass. Along with alpha or beta emissions, often a specific energy
gamma particle, or photon, is emitted. Each radioactive nuclide emits had a unique energy signature
which can be used to identify the individual nuclide. Neutrons are also emitted in some circumstances,
most notably during chain reaction fission events. To protect against emitted radiation, time, distance,
shielding are used. The issue of fissile material collecting to where it reaches a critical mass is also an
issue. Criticality happens when a sustained nuclear reaction occurs, and has happened throughout history
by accident on a number of occasions. [1] Such events must be prevented by not allowing unnecessarily
large amounts of material to gather.
The US Department of Energy (DOE) controls several sites and regulates numerous others which deal
with various amounts of different nuclear species. DOE Order 5480 uses the nuclear hazard categories
for a government facility set forth in the STD 1027-92[2] to regulate safeguards for radiation workers and
the surrounding environment. To determine the hazard categorization of a nuclear or radiological facility,
the complete inventory of radioactive material must be taken. The radioactive holdup deposited in
process equipment, ductwork, and drainage systems must be counted as part of the building inventory.
The accumulated holdup can found in all nuclear process equipment, ductwork, drains, pumps, sumps,
etc. An accurate quantification of the material is necessary to meet applicable regulations and minimize
occupational exposure.
1
2
Senior Physics Major, Honor’s Student, Oral Presentation
Mentor, Biology Department, Environmental Health Science Program
Holdup measurements become difficult to take for several reasons including but not exclusively:
unknown isotopic distribution, uncertain holdup locations, uncertain measurements, equipment location,
hazard area regulations, and facility conditions like temperature. The isotopic distribution within facility
structure and process equipment can not be fully known because different materials have different
properties and will deposit in different ways. These differences lead to uncertainties which are hard to
define and quantify. They also lead to an uncertainty in the location and size of the holdup deposits in the
various equipment and piping. Again this uncertainty is not easily quantifiable but is very real and differs
from facility to facility and must be reanalyzed for each building. Depending on the chosen detector the
uncertainty in the count rate of each nuclide differs though is significant regardless. While this is a
quantifiable uncertainty, it must still be taken into account between differing methods.
Accessing equipment and piping in corners, on ceilings, and behind other immovable obstacles becomes a
challenge with such heavy detectors. These obstacles must be over come as much as possible to acquire
the best data possible and reduce uncertainty. Some of these obstacles come in the form of regulations of
entering certain sections of the facility like hazard areas. Entering these area can be a challenge as well as
dangerous for those not familiar with the safety regulations in place. Some areas personal protective
equipment (PPE) is required. Wearing PPE can cause problems because of its interference with
measurements and personnel working ability. Anti-contamination PPE is traps heat and in warm areas
personnel have been known to overheat, thus interfering with reliable measurements.
The current methods of taking measurements are slow, expensive, and result large uncertainties (2550%).[3] This research is attempting to develop a model to improve holdup estimates. A simple, sample
facility has been created complete with labs, piping, and process wastes. Examining the uncertainty in the
most widely used method will provide a starting point for improvement on the methods. The suggested
improvement is to ratio the amount of material found in 1 cm of horizontal pipe to elbows, tees, and other
geometries found in typical nuclear laboratory facilities.
Methodology
The amount of each gamma emitting nuclide found from the scenario data will be used, along
with the waste profile to calculate estimated amounts of the other, non-detected nuclides, both
hidden gamma emitters and non-gamma emitters. The likelihood of computing differing
amounts of non-detected nuclides at each scan location is high from the different detected
nuclides. This discrepancy could arise from the fact that different species will remain with
varying frequency. Multiple, unrelated process could be utilizing the same waste system which
makes the use of the waste profile for just one of the process incomplete and therefore
inaccurate. A viable estimated amount each nuclide must be determined from the available
information.
The uncertainty calculations will be calculated by use of Monte Carlo Simulations. All non-accepted
values are entered as distributions instead of point-estimates with uncertainties. This allows the software
to determine the probability of a certain outcome. The software then runs a series of trials where the
spreadsheet is recalculated with random values following the input distributions. The resulting output is a
set of values with correlating statistics giving mean and standard deviation giving a probability density for
the result. This will allow the uncertainty in the measurements to estimated.
Changes in the scenarios are allowing analysis of the new method for taking measurements. Changes
include changes in amount found in locations. This will allow the study of how differing amount of
material change then total amount of material and the uncertainty in that total.
The changes to the facility are predominantly the number of each type of geometry. These differences
will allow for the sensitivity on the number of each type of geometry and determine if the majority of the
holdup comes from the horizontal pipes or from the tees.
Calculations
There are two different set of calculations going into this project. One is the straightforward calculation
of how much material is present. The other main calculations are the more complex calculations dealing
with how much uncertainty is present in the calculations.
The basic calculation starts with determining the isotopes and amounts present from the scans. With this
all the numbers must be corrected for attenuation, including air, the piping, and any additional shielding
placed to protect workers. Once the amount of material in each scan is corrected, species not detected by
gamma signatures can be calculated by determined ratios between them and detected isotopes. Once all
amounts of all isotopes are calculated, the sums through the building can be summed and total amount
determined.
Results:
The output from the Monte Carlo simulation is in the form of a histogram with sample statistics including
mean and standard deviation. For an initial nuclide, U-235, holdup in the building has been calculated to
be 2.1 mCi of material and four orders of magnitude less than the maximum amount of U-235 in a
radiological facility. [2] The standard deviation for this measurement is about 10% of the calculated
amount of material.
These results and conclusions must be verified using field studies which are outside the scope of this
work. These numbers are strictly theoretical and must be verified for accuracy.
Conclusions
The uncertainty in the various scenarios are being compared using Monte Carlo simulations. The
standard deviations of the final distributions can be compared offering comparisons of the uncertainties in
the measurements and calculations. The uncertainties found for the simulations will also be compared to
the current accepted uncertainty of 25-50% for holdup measurements.
This research is proposing a model for taking holdup measurements which are slightly more accurate and
less time consuming than current methods. From initial results this appears to have less uncertainty
which results in less regulatory issues for facility operators.
The proposed method should allow facility operators to not only more easily determine the holdup within
the facility but assess exposure in areas which are not normally assessed.
References
[1] “A Flash, and an Uncontrolled Chain Reaction” NY Times. October 1, 1999
[2] DOE-STD-1027-92. Department of Energy. 1997
[3]Zucker, M.S., “An Overview of Holdup Measurement Technology”, Journal of Nuclear Materials
Management. V.26(4), 1988. p.21.