Determination of the Radioactive Material and Holdup in Ducts and Piping in the 327 Building

D. L. Haggard L. W. Brackenbush

September 1995

Prepared for the U.S. Department of Energy under Contract DE-AC06-76RLO 1830

Pacific Northwest Laboratory Richland, Washington 99352

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Portions of this document may be illegible in electronic image products. Images are produced from the best available original document. Executive Summary

This report describes the measurements performed to determine the content and mass of plutonium in ducts, filters, and piping in the basement of the 327 Building at the U.S. Department of Energy Hanford Site in Washington State. This information is needed to characterize facility radiation levels, to verify compliance with criticality safety specifications, and to allow more accurate control using nondestructive assay (NDA) methods. Gamma assay techniques typically employed for NDA analysis were used to determine the gamma-emitting isotopes in the ducts, filters, and piping. Passive neutron counting was selected to estimate the plutonium content because high gamma levels from fission and activation products effectively mask any gamma emissions from plutonium. A high-purity gamma-ray detector was used to measure the mixed fission and activation . A slab neutron detector containing five 3He proportional counters was used to determine the neutron emission rates and estimate the mass of plutonium present. Both measurement systems followed the methods and procedures routinely used for nuclear waste assay and safeguards measurements.

The chronological order of events included a review of previously published documents on holdup assays at other sites and reviews of standards and regulatory guidelines relating to holdup and holdup measurement techniques. Reviews of facility historical radiation survey reports significantly helped in identifying potential holdup areas. A radiological survey mapping of the pipes and ductwork was the first subtask. During this subtask, collection zones were identified and permanently marked, and a unique naming protocol was established. This subtask was conducted in December 1993.

Subtask two was a preliminary assay of D and F cell ducts. The purpose was to determine if a significant holdup of was present at a "hot spot" reading 35 R/h between locations D5 and D6 of the D cell duct. Measurements were taken on March 23, 1994, and showed no significant quantities of special nuclear materials (SNM).

The fission and activation product inventories in the ducts and piping were measured using gamma- ray nondestructive assay techniques. A well shielded and collimated gamma detector was used to view specific locations on the ducts and piping. The intrinsic germanium detector was used to measure the characteristic gamma-ray spectra of the fission and activation products in the basement of the 327 Building. From the gamma-ray intensities measured and the decay scheme of identified radionuclides, it is possible to obtain reasonably accurate measurements of the activity of gamma-emitting nuclides. Almost all the gamma activity present originates from the fission products 137Cs, 134Cs, 154Eu, and the activation product ""Co. Radiations from other nuclides present were effectively masked by the intense gamma rays from the mixed fission and activation products. This was particularly true in attempting to directly measure for transuranic (TRU) materials. The small quantities of plutonium present did not produce sufficient gamma activity to be detected in the presence of so much fission and activation product activity. Transuranic activity was determined by measuring neutrons using a 3He neutron

iii detector. The fast neutrons emitted by plutonium are highly penetrating in metals and can easily be detected, even for gram quantities of plutonium surrounded by several inches of steel shielding.

Special data analysis techniques had to be used to estimate the plutonium content of the ducts since they were not in a standard counting geometry. The mass estimates may be too high if there are a significant number of neutrons produced by alpha-neutron interactions with low-atomic-number materials in the ducts and pipes, particularly alpha-emitters in intimate contact with fluorine, aluminum, boron, or beryllium. This is of concern if the high-efficiency particulate air (HEPA) filters contain finely divided plutonium oxide powder on borosilicate glass filter media. To be conservative, yet realistic, the neutron detector was calibrated using a known mass of weapons grade plutonium (6 weight percent 240Pu) in the form of an oxide. This source was well.characterized by previous NDA measurement to verify its mass.

The estimates of plutonium mass are highly dependent on the assumptions made in analyzing the data. Specifically, the estimates depend on the isotopic and chemical form of the plutonium, the distribution of plutonium, the amount of intervening shielding surrounding the plutonium, and the general neutron background in the basement and surrounding areas. Assumptions are that the plutonium is in the form of weapons grade plutonium oxide and at least 10 years old. This is a conservative assumption because much of the plutonium processed in the 327 Building is inthe form of high-burn-up material that emits more neutrons per gram (hence, the mass would be less for the same neutron emission rate).

Estimated plutonium mass in the basement ductwork and filters is 7.2 grams. The radioactive liquid waste system (RLWS) line has 4.2 grams and Special Environmental Radiometallurgy Facility (SERF) cell recirculating system contains 8.7 grams in the lower filter housing and associated piping. Total plutonium mass holdup estimates range from 20.1 grams, assuming that the plutonium is weapons-grade plutonium, to a best estimate of 11.0 grams plutonium, assuming 11 % 240Pu.

An estimated 0.3 curies of mixed fission and activation products are resident in the cell duct system. An additional 0.4 curies is in the lead shielded RLWS piping. Lastly, the SERF cell piping and duct work contain about 0.5 to 1.5 curies. A more definitive activity value for the complex SERF cell would require additional gamma measurements and equipment. This effort would not improve the TRU values already determined by neutron assay techniques.

In summary, the results of this survey indicate no significant levels of plutonium reside in the ductwork or RLWS lines in the 327 Building.

IV Acronyms

DOE U.S. Department of Energy

EBR-II Experimental Breeder Reactor

FFTF Fast Flux Test Facility

HPGe high-purity germanium

LLNL Lawrence Livermore National Laboratory

MCA multichannel analyzer

NDA nondestructive assay NIST National Institute of Standards and Technology

PNL Pacific Northwest Laboratory

RAM radioactive material RF radiofrequency RLWS radioactive liquid waste system

SERF Special Environmental Radiometallurgy Facility SNM special nuclear materials

TRU transuranic

V

Contents

Executive Summary iii

Acronyms v

1.0 Introduction 1.1

1.1 Facility Description 1.1 1.2 Radiological Survey • • • • 1-3

1.2.1 Radiological Survey Equipment 1.4 1.2.2 Delineation of Collection Zones and Assay Sites Procedure 1.4

1.3 Gamma Assay 1.5

1.3.1 Point Source Calibrations 1.8 1.3.2 Line Source Calibrations 1.8 1.3.3 Area Source Calibrations 1.8

1.4 Neutron Assay 1.9

2.0 Scope and Objectives 2.1

2.1 Definition of the Problem 2.1

2.2 Planning and Solution 2.2

3.0 Measurements 3.1

3.1 Measurement and Test Equipment 3.1 3.2 Calibration Standards . 3.1 3.3 Measurement and Quality Control 3.3 3.3.1 Preparations for Measurement and Test Equipment 3.3 3.3.2 Neutron Slab Detector System 3.4 3.3.3 Gamma Assay System 3.4 3.4 Pretest Verifications 3.5

3.4.1 Testing of Electronics 3.5 3.4.2 Angular Response Measurements 3.6 3.4.3 Calibration of Slab Detector with Plutonium Oxide 3.8

vii 3.5 Neutron Measurements in the 327 Building 3.8

3.5.1 Operational Check 3.8 3.5.2 Background Measurements 3.9

4.0 Analysis of Data 4.1

4.1 Analysis of Calibration and Verification Measurements 4.1 4.2 Analysis of 327 Building Measurements 4.3 4.2.1 Basic Assumptions Used for Calculations 4.3 4.2.2 Neutron Background 4.4 4.2.3 Calculated Mass of Plutonium 4.5

5.0 Results 5.1

5.1 Neutron Measurement Results 5.1 5.2 Gamma Measurement Results 5.4

6.0 References 6.1

Appendix A - Basic Holdup Assay Measurements Procedure MCA-510 A.l

Appendix B - Data Sheets for Radiological Survey B.l

Appendix C - Neutron Measurement Results C. 1

Appendix D - Gamma Measurement Results D.l

viii Figures

1.1 Point-Source Measurements by Collimated High-Purity Germanium Detector at Selected Distances from Source 1.5

1.2 Distribution of Gamma-Ray Activity Along the Horizontal Measurement Axis of the HPGe Detector 1.6

3.1 Neutron Energy Spectra from Plutonium Compounds 3.2

3.2 Energy Deposition Spectrum from the 3He Proportional Counters Exposed to a Neutron Source -. 3.6 3.3 Relative Angular Response of the Slab Neutron Detector as a Function of Polar Angle 3.7

3.4 Relative Angular Response of the Slab Neutron Detector as a Function of Azimuthal Angle 3.7

4.1 Example of Inverse Square Measurements with Slab Neutron Detector Exposed to a Plutonium Oxide Source Containing 108 Grams of Plutonium 4.2

IX Tables

3.1 Isotopic Composition of Plutonium Ash Reference Material, Weight Percent Isotope 3.3 3.2 Effect of Steel and Aluminum Shielding Materials Placed Between the Plutonium and the Slab Neutron Detector 3.9

5.1 Plutonium Holdup in Cell Ducts 5.1

5.2 Plutonium Holdup in SERF Cell Piping and Ducts 5.2

5.3 Plutonium Holdup in Radioactive Liquid Waste System Pipeline 5.3

5.4 Estimation of Plutonium in Exhaust Ducts at Accessible Locations Before Penetrations into the Basement 5.4 5.5 Cell Ductwork Curie Content By Direct Gamma Ray Measurement Using a Collimated High-Purity Germanium Detector 5.5

5.6 Gamma Dose Rates and Estimated Activity of Gamma-Emitters in the Ducts Leading from the Cells to High-Efficiency Particulate Air Filters 5.5

5.7 Diverter Box Curie Profile 5.8

5.8 Radioactive Liquid Waste System Pipeline Curie Profile 5.9

x 1.0 Introduction

This section provides a brief description of the facilities in the 327 Building to provide an overview of radionuclides that may be present in the various cells and ancillary ductwork. The equipment and methods used to provide estimates of the radionuclide inventories are discussed briefly.

1.1 Facility Description

The 327 Building Post Irradiation Testing Laboratory is operated by the Pacific Northwest Laboratory (PNL) for the U.S. Department of Energy (DOE).(a) The laboratory is used for temporary storage and for destructive and nondestructive examination of irradiated reactor fuels and structural materials. The facility contains 12 shielded hot cells, two water-filled basins, and dry storage. The 327 Building is a single story facility with a partial basement that provides access to the ducts, filters, and piping connected to the hot cells. The outside dimensions of the building are 65.6 m (215 ft) by 42.7 m (140 ft) by 9.8 m (32 ft).

The primary operating area is on the main floor and includes the canyon area, which contains the hot cells and the bays for auxiliary operations, including a machine shop, manipulator repair, and mock-up facilities. A metallographic facility is adjacent to the canyon and provides metallographic preparation and examination equipment for low-activity nonspent fuel specimens. The west end of the building contains a transfer and storage area that can be physically separated from the canyon by large doors. A Special Environmental Radiometallurgy Facility (SERF) cell is in a special bay on the north side of the canyon. The SERF cell provides an examination facility with a nitrogen atmosphere for specimens that may be affected by air. For example, plutonium metal specimens can be stored in the SERF cell without oxidation problems. The SERF facility is two levels with a storage area in the base• ment and an operating level on the first floor. During normal operations, nitrogen is continuously supplied to the cell and exhausted through the contaminated exhaust system in the basement adjacent to the lower SERF cell. Two airlocks provide entry into the SERF cell, and operations in the cell are performed remotely using manipulators. The SERF cell contains plutonium samples. One of the plutonium samples in the lower SERF cell was positioned next to the window to provide a performance check for the neutron counting equipment described later.

There are 11 hot cells containing an air atmosphere on the main floor in the canyon of the 327 Building. Two of the cells contain 6 in. of lead shielding; the other nine hot cells have 10.5 to 18 in. of high-density iron shielding. The two special cells exhaust into cells B and I. All operations in the hot cells are performed remotely using manipulators.

(a) The Pacific Northwest Laboratory is operated by the Battelle Memorial Institute for the U.S. Department of Energy under contract DE-AC06-76RLO 1830.

1.1 Work performed in the Post Irradiation Testing Laboratory includes measurement and testing of physical and mechanical properties, metrology, metallography, ceramography, autoradiography, fission product gas extraction and sampling, microdrill sampling, sampling for burnup and chemical analysis, encapsulation and disassembly of irradiated test modules, and encapsulation of irradiated structural materials for further irradiation testing. Irradiated fuels examined consist primarily of tested fuel pins containing plutonium oxides and oxides. Other fuels have been examined, including metal, carbides, and nitrides.

The individual hot cells are configured to support a wide variety of operations, and some are maintained free of alpha contamination. Typical work performed in each of the hot cells is described below. These activities may be used as a guide for estimating what types of nuclides and isotopic compositions may be present in the duct work and piping in the basement:

• A Cell is used for visual examination, sectioning and cutting, and packaging for disposal of irradiated fuel and high-level waste for transfers from the facility. A waste repackaging system is also installed in A Cell for disposal of transuranic (TRU) waste. High levels of neutrons may be present in the materials in A Cell.

• B Cell is used for structural materials postirradiation testing, including preliminary preparation for transmission electron microscopy and density measurements. Irradiated materials from experiments in the Fast Flux Test Facility (FFTF), the Experimental Breeder Reactor (EBR-II), N-Reactor, and light water reactors have been processed in B Cell.

• C Cell is used for metallography of irradiated fuels from experiments in die FFTF and EBR-II reactors. Fuel/cladding sections from light water reactor fuel rods and sections of N-Reactor fuel elements/process tubes have been examined in the cell.

• D Cell is used for material property testing of fuel pin cladding sections from EBR-II and FFTF experiments. Radio frequency (RF) heating and gas pressurization were used to simulate in-reactor accident conditions. Remote extensionmetry, in-cell welding, and macro- photography were also performed. The activities may have resulted in gaseous fission products, including cesium, being released into the ducts and filters.

• E Cell is identical to C Cell (see above).

• F Cell is used to recover test samples from test capsules, to recover fuel from cladding to prepare test specimens for density measurements or cladding tests in D Cell, and to "rough cut" sections for test specimens.

• G Cell is used for precision sample fabrication of material property test specimens (e.g., tensile, fracture toughness, etc.). The cell is maintained free of alpha contamination. One would not expect any measurable neutron emissions from the G Cell ducts and filters.

1.2 • H Cell is used for physical property tests of irradiated fuels and materials. Tests include thermal conductivity, helium leak testing, inerting cell interior for capsule disassembly, lithium reactions, gas sampling, and cesium reservoir spark tests of irradiated specimens from experiments (space power) in the EBR-II and the FFTF. The space power experiments may involve 238Pu with high neutron emission levels, making determination of fissile plutonium content difficult.

• I Cell is used for corrosion tests of irradiated fuel from waste repository studies. Fuel is exposed to controlled conditions of heat and moisture in long-term tests.

• SERF Cell is used for fission gas collection, precision sectioning of irradiated fuel pins, metallographic preparation and examination, and microhardness testing of fuel pins from materials irradiated in the EBR-H or FFTF reactors. An inert nitrogen atmosphere is maintained in the cell to exclude oxygen.

• Auxiliary Cell is used to carbon-coat specimens in preparation for electron microprobe examination.

1.2 Radiological Survey

Holdup measurements are required in all facilities where special nuclear materials (SNM) may exist in building systems. Potential criticality safety issues have been identified at other DOE facilities and have been addressed in PNL-MA-25, Criticality Safety, as action items.(a) The major objective of this project is to identify and quantify radioactive materials that may contain TRU holdup in various ducts, processes, and equipment in the 327 Building. The radiological mapping survey was used 1) to locate potential sources of radioactive materials/special nuclear materials (RAM/SNM) at specified locations; 2) to provide a baseline reference point for trend analysis on data from future measurements; 3) to augment ongoing radiological surveys; and 4) to provide supporting documentation regarding criticality safety and provide data for nuclear material accountability.

A radiological survey of all 327 Building cell ductwork, radioactive liquid waste system (RLWS) lines, and cells in the basement was conducted in December 1993 and January 1994. The survey map readings were taken by PNL health physics technicians. Guidelines for holdup assay are stated in U.S. Nuclear Regulatory Commission (NRC) Regulatory Guides 5.23 and 5.37 (1984, 1983) and the Los Alamos Nuclear Safeguards Training Program (LANL 1992).

(a) Pacific Northwest Laboratory (PNL). August 17, 1994. Criticality Safety. Controlled technical manual, PNL-MA-25. Pacific Northwest Laboratory, Richland, Washington.

1.3 1.2.1 Radiological Survey Equipment

Equipment used for the survey included extended probe instruments, such as the Teletector probe, an Eberline R07 having a range of 0-20 R/h, and an extended marking device for permanently labelling measurement locations. The use of extended probes and marking devices evolved from a hazards assessment of the areas to be surveyed. The potential hazards identified were possible contamination and high-dose-rate areas. Personnel safety issues were addressed and covered in the radiological work procedure specific to this work. The radiological survey procedure followed regulatory guidelines. Preliminary fadiation survey measurements of the facility were used to budget the measurement time, to emphasize high-holdup areas, to establish independent collection zones, and to determine detector positions within the zones.

1.2.2 Delineation of Collection Zones and Assay Sites Procedure

The procedure used to establish assay locations is given below.

1. At each collection zone, detector positions (assay sites) were chosen so that the material holdup could be measured from several vantage points around the zone.

Each cell ductwork constituted a collection zone, For example, G-cell with its ductwork from the ceiling to the secondary side of the high-efficiency particulate air (HEPA) bank was a collection zone.

2. Each assay site was marked with a permanent marker to ensure reproducible assay positions.

Dose rate measurements were taken at 120° points around the duct, beginning 6 in. down from the ceiling and at successive 2-ft intervals. Interval points were marked and identified using permanent ink or equivalent.

"Hot spots" between interval points were identified and marked. Dose rate information was recorded on a data sheet or schematic as provided by the subtask leader.

3. Each site was uniquely labeled to facilitate unambiguous reference to that site in the assay log. A labeling convention was established.

Labeling proceeded from top to bottom, beginning with cell letter designation followed by the interval identification number. For example, the location of the first interval measurement point in G cell was Gl. Measurements began on the cell ductwork having the lowest known dose rates and general area background and proceed to the higher-dose-rate ducts.

1.4 3000

-4-3 -2 -1 0,1 2 3 4 Distance (inches)

Figure 1.1. Point Source Measurements by Collimated High-Purity Germanium Detector at Selected Distances from Source

This enabled the radiation protection technicians (RPTs) and operators to become familiar with the sequence and operational procedure. Subsequent cell duct measurements were taken in the same manner.

1.3 Gamma Assay

Gamma-ray assay is based on the activity observed from specific energy peaks detected from the isotopes present. Equipment for gamma holdup measurements consisted of the following components:

• tungsten collimator

• high-purity germanium (HPGe) detector - efficiency 35 %

• angular response of the detector/collimator reference (see Figures 1.1 and 1.2)

• Canberra Genie-PC AIM Data Analysis and Acquisition System.

1.5 327 COLLIMATOR #1 12" FROM SOURCE

3000 T

<_> 1LU- LU Q C/3 I- o a

o -4 INCHES OFF-CENTER

CS137 661KEV ' C060 1173KEV = CO60 1332 KEV

Figure 1.2. Distribution of Gamma-Ray Activity Along Horizontal Measurement Axis of the HPGe Detector

The HPGe detector was shielded and collimated from other surrounding source material. A tungsten collimator was used with the gamma-ray detector. Additional shielding was also needed to eliminate background gamma rays originating behind the detector. The collimator provided sufficient shielding and directionality for the holdup measurement geometry. The detector's field of view was defined by the tungsten collimator. Calibration of the gamma system was performed with the same coilimation setup used during the holdup measurement campaign.

Initial calibrations for the holdup measurement system were performed for point, line, and area source geometries. These point, line, and area geometry calibration constants described the distribu• tion in the detector's field of view. The correction factors were specific to the detector and coilimation setup and were used to correct for the detector's radial response to the incident radiation of interest. These corrections were used in the correlating measurements to convert the results to activity or grams of material. This procedure used a single point source to determine these correction factors. Point source measurements were conducted at several horizontal positions, as indicated in Figure 1.1. Fig• ure 1.2 illustrates the gamma detector's response along the horizontal measurement axis. The gamma detector's relative efficiency for that geometry was determined. One reference standard centered in front of the detector at a fixed distance was measured to determine the detector's absolute response to the incident radiation.

1.6 Formulae required to determine calibration constants are as follows:

£c (1.1) L=2s±2—-; s

C0

where L = effective length of the detector field of view at distance r0 s = length of the horizontal increment G = net count rate at horizontal position i Cj = net count rate with the source centered in front of the detector.

ITS2 (1.2) ^-4 where a,, is the incremental area at horizontal position 0

2 2 ai=ir[(i + 0.5)s] -ir[(i-0.5)s] (1-3) where a; is the incremental area at horizontal position i

£(a-A) (1.4) A-.1"0

C0

where A is the effective area of the detector field of view at distance r0

KP= (1.5)

Kr (1.6) LC/o

mo (1.7) Ka AC0

1.7 where Kp = point source calibration constant (g-s/cts-cm2) K, = line source calibration constant (g-s/cts-cm2) Ka = area source calibration constant (g-s/cts-cm2) mo = mass of the reference source (g) CQ= net count rate of the reference standard centered in front of the detector face at distance

r0 (cts/s).

Once calibration constants are determined, quantitative estimates can be made for point sources, line sources, and area sources.

1.3.1 Point Source Calibrations

A point source is a small deposit centered in a large detector field-of-view. Its calibration is determined with Equation 1.8: *

m=KCr2 (1.8) p where m = mass of isotope of interest (g) K,, = point source calibration constant (g-s/cts-cm2) C = net count rate (cts/s) r = source-to-detector distance (cm).

1.3.2 Line Source Calibrations

A line source is a narrow uniform deposit, centered in a wide detector field-of-view, which spans the width of the detector field-of-view. Its calibration is found with Equation 1.9:

m=K1Crl (1.9) where m = mass of isotope of interest (g) K; = line source calibration constant (g-s/cts-cm2) C = net count rate (cts/s) r = source-to-detector distance (cm) 1 = length of a line source (cm).

1.3.3 Area Source Calibrations

An area source is a uniform deposit that fills the detector field-of-view. Its calibration is found with Equation 1.10:

1.8 m=KCa (1.10) a where m = mass of the isotope of interest (g) K„ = area source calibration constant (g-s/cts-cm2) C = net count rate (cts/s) a = area of an area source (cm2).

The equipment was moved to the 327 Building and checked for physical damage. A combination standard traceable to the National Institute for Standards and Technology (NIST) was measured periodically during the measurement campaign to demonstrate that the instrumentation was properly functioning and that the calibrations determined in the laboratory were still valid.

1.4 Neutron Assay

Accurate determination of the and radioisotope content in the ducts, filters, and piping of the 327 Building is a very difficult problem. Nondestructive assay techniques for fissile materials usually employ gamma spectroscopy to identify and quantify each gamma-emitting nuclide, such as plutonium. However, the massive amounts of fission products present completely obscure any gamma rays emitted by plutonium. Fortunately, the neutrons emitted by plutonium are highly penetrating and can easily pass through several inches of the steel or lead shielding used to reduce gamma doses, although the neutrons may be degraded in energy. With some degree of uncertainty, the amount of plutonium can be determined using appropriate calibration standards and neutron detectors. Passive neutron counting is a recognized technique for nondestructive assay of plutonium and is described in Chapter 14, "Principles of Neutron Counting" in Passive Nondestructive Assay of Nuclear Materials, NUREG/CR-5550 (Reilly et al. 1991).

Neutrons emitted by plutonium originate from two main sources:

• spontaneous fission of even-numbered plutonium isotopes

• alpha-neutron reactions with any low-atomic-number materials in intimate contact with the plutonium.

Most of the neutrons originate from ^Pu, 240Pu, and ^Am, which is the decay product of the short-lived nuclide M1Pu. Thus, it is important to know not only the isotopic composition of the plutonium, but also the time since chemical separation or the ^Am content to be able to estimate the neutron contribution from ^Am. In many practical situations, the number of neutrons produced by alpha-neutron reactions can exceed the spontaneous fission production from plutonium. Any alpha-emitter in contact with low-atomic-number nuclides will produce neutrons. For example, trans- uranics that are in intimate contact with beryllium, boron, aluminum, sodium, or even concrete will

1.9 produce neutrons. High neutron emission rates can result from finely divided plutonium oxide dust on borosilicate glass or aluminum spacers in HEP A filters. The neutron emission rates can be tens to hundreds of times higher than pure plutonium metal.

It is conceivable, although not likely, that the material may not contain any plutonium, but will still emit neutrons from alpha-neutron reactions from americium or other transuranic material with high alpha activity. The exact amount of plutonium present can be determined from the neutron measurements if we know the following:

• the exact isotopic composition of the plutonium

• the time since chemical separation to allow calculation of ^Am progeny

• chemical composition of the plutonium

• presence of low-atomic-number impurities in contact with alpha-emitters

• other transuranics present that may also emit neutrons, such as ^Cm.

We can make some simplifying assumptions that will limit the range of calculated values of the . amount of plutonium that may be present. For instance, we can assume that the plutonium will be low- exposure plutonium. Neutron emission rates of weapons grade (less than 6% 240Pu) or N-Reactor plutonium are quite low, typically about 150 neutrons per second per gram of plutonium. Higher- exposure plutonium would emit more neutrons and result in a smaller actual mass for a given measured neutron emission rate.

Plutonium metal is chemically unstable, and any small pieces would eventually convert into an oxide or oxide-hydroxide mixture in moist air. The oxide is the most chemically stable form of plutonium and would be the most likely form to be found. The neutron emission rates of oxylates, nitrates, and hydroxide mixtures will be somewhat similar to that from pure plutonium oxide. Assuming any plutonium present would be in the form of an oxide will result in conservative estimations. Any other chemical form, particularly fluorides, would emit more neutrons and result in a smaller actual mass for a given measured neutron emission rate. However, oxides are the most probable chemical form found in the reactor fuel.

1.10 2.0 Scope and Objectives

The objectives of this work are threefold:

• provide a radiological survey of the ducts and piping

• determine an upper limit of plutonium and estimate of the most probable amount of plutonium that may be present

• locate and quantify the amount of fission, activation, and TRU materials that are present in ducts, RLWS lines and other facility processes, as a baseline for future reference.

This work is limited in scope to the nondestructive assay of the contents of the ducts, filters, and piping in the 327 Building and low-level waste that resided in those locations at the time measurements were taken.

2.1 Definition of the Problem

Over many years of operation of the hot cells in the 327 Building, the internal ducts, filter boxes, and piping and drain lines from the hot cells have become contaminated with large amounts of fission and activation products and small amounts of plutonium and other transuranic nuclides. The fission and activation product inventory is large enough that it presents a radiation hazard, and large amounts of steel and lead shielding have been added to critical areas to reduce radiation doses. It is conceivable that enough plutonium and transuranic material has accumulated to be of concern for criticality safety and safeguards controls.

The purpose of this work is to attempt to quantify the amounts of radionuclides in the internal ducts, filter boxes, and pipes leading from the hot cells. Of particular concern are the liquid waste lines that may have accumulated fuel debris, with the potential build-up of quantities of plutonium. Because of its high activity, the radioactive cesium and cobalt are relatively easy to locate. However, the high gamma levels will mask the gamma emissions from plutonium and other transuranics. The problem is further exacerbated by the streaming of neutrons down ducts and openings from the hot cells. Inventories of plutonium, , arid other neutron-emitting nuclides will create variable backgrounds in the basement in the area around the ducts. It is extremely difficult to accurately assess plutonium inventories in the presence of variable neutron backgrounds. Passive neutron measurements are further complicated by the storage of over 70 grams of 238Pu oxide stored in a safe in the basement of the 327 Building. The neutron emission rate from the ^^ oxide is orders of magnitude higher than from other sources we are attempting to measure.

2.1 2.2 Planning and Solution

A review of the radiological conditions was made, including gamma surveys to determine hot spots in the ducts and piping. High contamination levels and excessive gamma dose rates from fission products preclude the use of gamma assay equipment routinely used to determine plutonium mass. The high dose rates from fission products would interfere with the plutonium isotopic analysis. Therefore, neutron assay techniques would be more reliable for locating and quantifying fissile material. A neutron slab detector based on a design from the Los Alamos National Laboratory (LLNL) Safeguards Assay Group was selected for measurements to determine the amount of plutonium that could be present. The amount of plutonium in the ducts and RLWS pipes can be estimated by:

• calibrating the neutron detector with appropriate calibration standards

• measuring the around the ducts, filter boxes, and piping

• carefully analyzing the data to relate the measured neutron fluxes to mass of plutonium.

The gamma-emitting nuclides were readily identified by standard gamma-assay techniques because in most instances the nuclides are not heavily shielded in the ducts and pipes. However, in areas with high concentrations of radiocesium and cobalt, up to 2 in. of lead shielding have been added. In most situations, the amount of shielding is well known and the radioactive material localized. A well collimated intrinsic germanium detector was used to quantify gamma-emitters. Where space permits, measurements are made at three different orientations and the results averaged to give an accurate estimate of the contents inside ducts and pipes.

2.2 3.0 Measurements

The measurements were conducted according to the following documents:

• Pacific Northwest Laboratory Quality Assurance Plan QAP FO-1, QA Plan for Safeguards Nondestructive Assay Measurements

• Pacific Northwest Laboratory Nondestructive Assay Measurement Procedure MCA-510, Holdup Assay Measurements (see Appendix A).

3.1 Measurement and Test Equipment

The neutron and gamma assay systems are "user to calibrate" equipment. The individual components are configured and tested as a system, and then the entire system is calibrated with appropriate calibration/verification standards. The neutron and gamma assay systems are calibrated with gamma check sources with calibrations traceable to the NIST.

3.2 Calibration Standards

Calibration standards are used in a measurement system to establish the relationship between the basic instrument response and the attributes of interest. The quality of the calibration is ensured by selecting the appropriate standard. The standard must be a physically and chemically stable item for which the attributes of interest are well characterized and for which other properties affecting the measurement are known. In this case, the count rate from neutrons, the amount of shielding, and the orientation and distance from the detector to the neutron source are related to the mass of weapons grade plutonium. The mass of other forms of plutonium can be inferred from the calculated neutron emission rates. The calibration of the gamma assay equipment is more straightforward. The counting efficiency of the gamma spectrometer with a collimator can be measured using calibration sources of known activity and photon emission rates. The viewing angle and efficiency can be determined directly from measurements, as shown in Figure 1.2.

Records indicate that the material in the hot cells was from a wide variety of sources, including both low-exposure and high-exposure reactor fuel samples. For these measurements, it was assumed that the plutonium would be low-exposure plutonium oxide with relatively low neutron emission rates (typically 150 neutrons/second/gram of plutonium). To give conservative results, any high neutron yields from alpha-neutron reactions were ignored. The neutron energy spectrum from plutonium dioxide is shown in Figure 3.1, taken from the document Neutron Spectra of Plutonium Compounds, BNW-1262 (Brackenbush and Faust 1970).

3.1 0 1 2 3 4 5 Neutron Energy (MeV) 39101085.9

Figure 3.1. Neutron Energy Spectra from Plutonium Compounds

Two neutron sources were selected for use in calibrating/verifying the neutron detector assay system. The first was a small 0.3-^g 252Cf neutron calibration source. This source's neutron dose rates have been determined by comparison with other neutron sources whose calibrations are directly traceable to NIST. This source has a neutron energy spectrum very similar to that of plutonium dioxide. (See mQ Shielding Guide, DP-1246 [Stoddard and Hootman 1971] or NBS Special Publication 633 [Schwartz and Eisenhauer 1982].) This source was used to determine the angular efficiency of the slab detector in the laboratory prior to the measurements at the 327 Building.

A sample of weapons-grade plutonium oxide was selected as the reference material standard for determining the basic instrument response for the attributes of interest. Specifically, the mass of plutonium in the reference material was related to the neutron count rate in the neutron detector. Weapons grade plutonium has a lower neutron emission rate than the higher exposure plutonium samples in the hot cells and should yield conservative (i.e., higher) plutonium mass estimates. The plutonium oxide sample (LLNL ash sample RA 146A) selected for the reference standard contains 108 grams of plutonium with the isotopic composition given in Table 3.1. The reference standard has been measured by several nondestructive assay (NDA) systems used for measuring safeguard mass

3.2 Table 3.1. Isotopic Composition of Plutonium Ash Reference Material, Weight Percent Isotope

m 238pu 239^ 240^ 241^ 24 2pu Am

Weight 0.13825 93.452 6.101 0.2793 0.019 0.220 Percent Isotope ±1-53% ±0.15% ± 2.21% ± 1.22% n/a ± 1.4%

Calculated age from ratio of^Am/241!^ is 12.3 ± 0.33 years. confirmatory standards. Data from this sample from high-level neutron coincidence counting and calorimetry with isotopic compositions obtained from gamma spectroscopy are in good agreement with the assigned book values assigned by LLNL when the material was shipped to PNL.

3.3 Measurement and Quality Control

The measurement personnel had the complete responsibility for monitoring and evaluating the quality of the data. Any indication that the system was out of control (i.e., significant gain shifts in the position of the thermal neutron peak or high gamma levels that produce pile-up in the neutron events) required the stoppage of work until the problem was resolved. Before NDA measurements were initi• ated, the overall system was calibrated. To verify the proper operation of the neutron assay system, the entire system was assembled in the laboratory and components checked for proper operation. Then, verification measurements were performed using a known mass of plutonium before and after the measurements in the 327 Building. All of the spectral data collected during the measurements were recorded on computer files and can be retrieved if questions should arise.

A measurement control check was performed daily before the equipment was used, and the results recorded. For the neutron assay system, the detector was positioned at a fixed location at 12 in. from the window of the SERF cell in which a plutonium sample was positioned near the window. The measurement control check was used to check the position of the thermal neutron peak and verify that the integral of the neutron events remained constant throughout the measurements. For the gamma assay system, the germanium detector was exposed to a calibration source containing a known activity of radioactive material. An analysis of photopeak areas and channel number is used to verify that the gain and sensitivity has remained constant. These procedures are recorded in Appendix A, and the data are archived on magnetic storage media.

3.3.1 Preparations for Measurement and Test Equipment

The approach to the problem of NDA assay was to select a very sensitive neutron detector to measure the neutrons emitted by any TRU (plutonium) contained in the ducts, filter boxes, and pipes.

3.3 The detector requirements included sufficient sensitivity to measure sub-gram quantities of plutonium at accessible measurement locations. A slab neutron detector was selected for the measurements because of its sensitivity and directionality.

3.3.2 Neutron Slab Detector System

The neutron detector system consists of

• a slab neutron detector containing several 3He proportional counters inside a polyethylene moderator

• ancillary NIMbin electronics (preamplifier, shaping amplifier, high voltage power supply)

• a multichannel analyzer (MCA) to record the gamma and neutron spectra from the detector.

The slab neutron detector consists of an array of five cylindrical 3He proportional counters, 1 in. in diameter by approximately 22 in. long, which are filled with four atmospheres of 3He to detect slow neutrons. The proportional counters are inside a polyethylene moderator at a depth of 2 in. from the front face and 4 in. from the back face. The tubes are carefully selected to have the same gain and are connected in parallel to a preamplifier and high voltage supply. The nominal operating voltage is +1400 volts. The polyethylene moderator is 6 in. deep, 16 in. wide, and 24 in. long. The detector is a standard design used for NDA safeguards assay and is described in Section 14.4.2 of NUREG/ CR-5550, Passive Nondestructive Assay of Nuclear Materials (Reilly et al. 1991).

The NIMbin electronics include a high-voltage power supply to provide a regulated voltage of +1400 volts DC to operate the proportional counters and a shaping amplifier to convert the signals from the preamplifier to pulses that can be processed by the MCA. An MCA was used to record the spectra, so that any possible gamma interference or instrument malfunction could be detected immediately. Data from the MCA (Canberra Series 35 Plus) were recorded in a lap top computer with a hard disk for permanent data storage. If there is any question about a measurement, the spectral data can be retrieved and examined later.

3.3.3 Gamma Assay System

Two HPGe detectors were used to determine radionuclide identification and quantities present at various measurement locations. One was a 4.2% efficient detector. It was used to measure the higher dose rate areas which could not be measured by the other 35% efficient detector due to high count rate and pulse pile-up. Some locations were measured with both detectors. The results of these measurements were compared and used to provide an error estimate at the measured location. The gamma assay system is composed of the latest innovative hardware and software commercially available.

3.4 3.4 Pretest Verifications

Pretest verifications consisted of testing the electronic equipment, making angular response measurements, and calibrating the slab detector with plutonium oxide.

3.4.1 Testing of Electronics

The neutron slab detector and supporting electronics were assembled in the laboratory, and all of the electronic components were checked to assure proper operation in the field. The electronics were placed in the NIMbin and the cables connected for the signals, high voltage, and preamplifier power. The methods used to verify the proper operation of the equipment in the laboratory generally followed the methodology given in Neutron Dosimetry at Commercial Nuclear Plants, NUREG/CR-3610 (Brackenbush et al. 1984), which describes the setup of neutron spectrometry equipment using 3He proportional counters. The output signals from the shaping amplifier was examined by exposing the detector to a ^Cf neutron source in the laboratory and observing the pulses with an oscilloscope to examine the pulse shape. Adjustments are made to correct for amplifier gain and pole zero to minimize pulse overshoot or undershoot, so that the pulse returns to the baseline in the shortest possible time to reduce pulse pile-up and allow operation in high count rates. Because this adjustment depends on the cable capacitance, the entire system was tested, including the long cables that were to be used. If any component failed, there were backup units that were also checked.

The equipment was calibrated as a complete system before the 327 Building measurements were initiated. Following safeguards measurement procedures, the system was calibrated at the system level. It is not necessary, or even desirable, to calibrate the individual components of the system, because proper operation of the system and accurate interpretation of the results cannot be obtained from individual component electronic calibration. System malfunctions usually can be easily identified from the spectra recorded during the measurements. Proper operation of the system was verified with measurements using a known mass of plutonium.

The calibration/verification measurements consisted of measuring the angular response and response to room-scattered neutrons using the 0.3-jtg ^^f neutron source in the ESB Building at PNL. The response of the detector would then be known at any angle and distance to a fission source. The mass of the plutonium was then correlated to the neutron count rate from the slab detector by exposing the detector to the 108-g plutonium oxide source in the 324 Building.

Figure 3.2 shows a typical spectrum obtained from the 3He proportional counters in the slab detector exposed to the neutron source. In this graph, the ordinate is the logarithm of the number of counts from the detector, and the abscissa is the energy deposited in the detector by gamma rays and neutrons. There is a clear separation between gamma events, shown on the left side of the plot, and neutron events. Gamma events can be eliminated by setting a region of interest and integrating only neutron events as shown in Figure 3.2.

3.5 1000 .Thermal neutron peak Gamma events

Start of neutron P. © events c • : c 100 • CO o w 0 a u c 10 3 o O

-»-*- • «'» • » » e st 200 400 600 800 1000 1200 Channel Number

Figure 3.2. Energy Deposition Spectrum from the 3He Proportional Counters Exposed to a Neutron Source

However, pulse pileup can occur in intense gamma fields, when several gamma pulses are counted within the resolving time of the electronics. This produces a gamma continuum that can extend into the neutron event region. The slab detector was exposed to a 137Cs gamma source in the ESB Building lab• oratory to determine the operating range in high radiation fields. There is no significant gamma pile- up that could interfere with the neutron detector at gamma exposure rates up to 1 R/h. The detector can function in gamma fields as high as 35 R/h, but there is serious gamma pile-up, and the region of interest must be adjusted. The spectral data were recorded, so that all of the measurements can be reviewed later, if necessary.

3.4.2 Angular Response Measurements

The first set of calibration/verification measurements were made in the ESB Building laboratory, using the 0.3-/tg ^Cf neutron source to set up the system electronics and determine the angular response of the slab detector. The response of the detector as a function of polar and azimuthal angles is given in Figures 3.3 and 3.4, respectively. Angles are measured in reference to an axis normal to the center of the front face of the slab detector. (The front face has the five 3He tubes located at 2 in. below the surface of the plastic moderator.) The azimuthal angles are measured in the plane passing through the longest axis of the rectangular slab; polar angles are measured in a plane at right angles to the longest axis of the rectangular slab. These measurements were made in the ESB Building labora• tory at a height of 4 ft above the floor. The angular responses were measured with the slab detector

3.6 1.00 r

0.90 o 0c9 •a 0.80 u 0.70

arm a 0.60 / 2 B 0.50 / » >

at i 0.40 OC 0.10

0.00 I I I J -180 -150 -120 -90 -60 -30 0 30 60 90 120 150 180 Polar Angle {degrees)

Figure 3.3. Relative Angular Response of the Slab Neutron Detector as a Function of Polar Angle

o.oo -180 -150 -120 -90 -60 -30 0 30 60 90 120 150 180 Azimuthal Angle (degress)

Figure 3.4. Relative Angular Response of the Slab Neutron Detector as a Function of Azimuthal Angle

3.7 positioned vertically on a turntable; the neutron source was at a distance of 6 ft (2 m) from the center of the 3He tubes. Details of the measurements are given in Appendix B. The graphs demonstrate that the angular counting efficiency does not change dramatically if the source is slightly off-axis, and that neutrons entering the back side of the detector are counted at about 50% of the efficiency of those entering the front.

3.4.3 Calibration of Slab Detector with Plutonium Oxide

The slab detector was also calibrated in the 324 Building by exposing the detector to a 108-g sam• ple of plutonium oxide (LLNL sample RA 146A) at a height of 4 ft above the floor and in a vertical configuration. The 108-g plutonium source has been described previously in Section 2.0. As will be explained in Section 4.0 on data analysis, it is possible to correct for the effects of room-scattered neu• trons from the series of measurements that relate distance to the measured neutron count rate. After these measurements were completed, it was possible to relate the mass of weapons-grade plutonium to the measured count rate, the distance from the detector, and the angle of incidence. As will be explained in the data analysis in Section 4.0, it is even possible to correct for the effects of neutrons scattered within the facility from inverse square measurements performed in the 324 Building.

Measurements were also made in the 324 Building using various thicknesses of steel and aluminum to estimate the effects of shielding. The shields were approximately 2-ft square and were positioned near the plutonium source to effectively shield it from the lab detector. The can of plutonium oxide containing the 108 grams of plutonium was positioned on its side on the floor inside a "cave" of steel and lead bricks. Measurements were then made with the bare source and with the source covered with large slabs of iron and aluminum positioned on top of the can containing the plutonium oxide. The steel and lead cave prevented neutrons from scattering around the edges of the intervening shield mate• rials. Details of the measurements are given in Appendix C, and the results of the shielding measure• ments are given in Table 3.2. Note that the neutrons were not absorbed like gamma rays in steel shields; the neutrons scattered within the shield and were lower in energy. These.data indicate that even 2 in. of steel or aluminum do not greatly influence the measured neutron flux.

3.5 Neutron Measurements in the 327 Building

Operational and background checks were made for neutron levels.

3.5.1 Operational Check

After the preliminary measurements were completed in the laboratory and the system was function• ing properly, it was moved to the 327 Building. Here, the neutron detector was connected to the MCA via 100-foot-long cables, so that the operator could remain in low-dose-rate areas while the detector was positioned near the ducts and piping. Before the measurements were started, a plutonium sample was positioned in the center of the lead glass window of the SERF cell. The slab neutron detector was

3.8 Table 3.2. Effect of Steel and Aluminum Shielding Materials Placed Between the Plutonium and the Slab Neutron Detector

Thickness Attenuation Shield Material (in.) Factor None 1.00 1 0.85 2 0.75 3 0.68 4 0.6 6 0.49 None 1 0.5 0.94 1.5 0.84 2 0.8 3.5 0.68 4 0.65 positioned at the center of the window at 12 in. from the surface of the SERF cell and the centerline of the 3He proportional counter tubes in the moderated detector. The neutron event spectrum, similar to that shown in Figure 3.2, was measured and recorded every day to verify that the detector and ancillary electronics were functioning properly. After the operational check indicated that the system was still functioning properly, it was ready for neutron measurements on the ducts and pipes.

3.5.2 Background Measurements

Background measurements were performed with an HPGe detector and a neutron counter.

High-Purity Germanium Detector

Background measurements were taken with an HPGe at each measurement location. A solid plug replaced the collimator and a background measurement was taken for the same counting time as the collimated measurement. The background contribution was subtracted from the collimated measurement with analysis conducted on the resulting background-corrected spectrum.

A small, low-power neon laser was used to align the gamma detector to the desired measurement location on the duct, RLWS line, and other measurement locations of interest. Detector-to-source distance was measured using a SONIN sonic measuring unit. Accuracy of the distance measurement was ±0.5 cm.

3.9 Neutron Measurement System

Other neutron sources in the building could contribute significantly to the neutron count rate, so careful background measurements were made before any measurements were made on the ducts or pipes. If the neutron count rate were too high, it would be necessary to modify the measurement plan. The first background measurements were made with the detector positioned in front of the door on the east side of the basement; the results indicated a background of some 38,000 counts in a 5-minute countmg period. This background was at least 3 orders of magnitude above that expected from nearby plutonium sources. It was found that a safe contained some 72 grams of ^'Pu oxide emitting about 1.15 x 106 neutrons/s. It was necessary to move the safe to the other end of the building before any measurements could be initiated.

The detector was designed to have a directional response, so that fission neutrons entering from the back side would be counted with only half the efficiency of those entering from the front of the slab detector. Background measurements were made on the periphery of the area around the ducts and filter boxes. The slab detector was positioned in a vertical orientation to measure the general background. At a given position, a measurement was taken; then, the slab detector was rotated 180° and the measurement repeated. This procedure gave pairs of neutron count rates with the detector facing the source and pointed away from the suspected neutron source. These data allowed the average room background to be determined, as explained in Section 4.0 on data analysis.

In some instances, however, a significant neutron background existed from neutrons scattering through openings in the ceiling from the hot cells above. In these cases, it was necessary to perform an inverse square measurement in order to estimate the contribution of scattered neutrons from the hot cells. Again, it was necessary to make measurements with the detector facing towards and away from the suspected source of neutrons in order to estimate general room backgrounds using the technique explained in Section 4.0 of mis report.

3.10 4.0 Analysis of Data

Data was analyzed for both the calibration/verification mesurements and for measurements taken in the 327 Building.

4.1 Analysis of Calibration and Verification Measurements

The calibration or verification measurements made before the 327 Building measurements allow one to correlate the mass of weapons-grade plutonium to the measured neutron count rate and account for scatter of neutrons from the walls, floor, and ceiling of the room. This method of correcting for scatter from inverse square measurements is presented in the National Bureau of Standards Special Publication No. 633 (Schwartz and Eisenhauer 1982). The response of any moderated neutron detector can be modeled as a response to neutrons coming directly from the source and a response to neutrons scattered from the floor, walls, and ceiling of the room. If the detector is more than 1 meter from the floor and walls, the scattered component can be assumed to be a constant. This can be expressed mathematically by Equation 4.1:

c = MB (4.1) r2 where C = the count rate (cts/s) M = the mass of plutonium in grams (directly proportional to the neutron source strength) B = the neutron background in the room due to scatter from the floor, walls, and ceiling r = the distance (in feet) from the plutonium source to the centerline of the proportional counters.

Multiplying by r2 gives Equation 4.2:

Cr2= MBr2 (4-2>

In the limit that r2 approaches zero, the term Br2 approaches zero while the term Cr2 approaches a constant. Plotting the product of the count rate times the distance squared versus the distance squared produces a straight line if the detector is not too close to the source or walls. An example of the inverse square plot from the 108-g plutonium oxide source is shown in Figure 4.1. A straight line has been fitted to the data points. The intercept on the ordinate axis is proportional to the neutron source strength or mass of plutonium, and the difference between the intercept and the value on the straight

4.1 5000 r

10 15 20 25 30 35 Distance Squared (sq. feet)

Figure 4.1. Example of Inverse Square Measurements with Slab Neutron Detector Exposed to a Plutonium Oxide Source Containing 108 Grams of Plutonium line gives the contribution to room scatter. In some small rooms, the scatter can be as high as 50% of the count rate at distances of 6 ft (2 m) from the neutron source. Thus, it is important to correct for the effects of room scatter.

This technique can be used to correlate the detector count rate to the mass of plutonium. Consider the case where the can of plutonium oxide was 4 ft above the floor and the slab neutron detector was positioned vertically. The spreadsheet in Appendix C contains a plot of the count rate times the square of the distance versus the square of the distance, which is a plot of the inverse square data for Equation 4.2. The least squares fit gives the values for the slope and y-intercept of a straight line fitted to the inverse square data points. Dividing the slope and intercept values by the mass of plutonium (108 g) gives the results normalized to 1 g. The normalized y-intercept is 0.507, and the normalized slope is 23.66. Putting these values into Equation 4.1 gives an expression for the count rate per gram of plutonium for the slab detector, oriented vertically at 4 ft above the floor. This is expressed mathematically by Equation 4.3:

23.7 cts/s/g 0.507 (4.3) where r is the horizontal distance in feet between the plutonium source and the centerline of the 3He proportional counters at 4 ft above the floor.

4.2 4.2 Analysis of 327 Building Measurements

The estimate of the amount of plutonium in the debris at the 327 Building depends on the assumptions made concerning:

• the composition, age, and location or distribution of the plutonium

• the position of shielding materials around the plutonium sources

• the neutron background in the area around the piutonium sources.

4.2.1 Basic Assumptions Used for Calculations

It is possible that there is very little plutonium in the ducts and piping. Many of the measured neutrons could originate from alpha-neutron reactions in low-atomic-number materials, such as borosilicate glass in HEPA filters, and from neutrons from plutonium and curium in the fuel samples in the hot cells that have been scattered into the ducts from above. It is more probable that the neutrons originate from plutonium and other transuranics (such as M1Am) in the ducts and piping, and great care must be taken in distinguishing between room background and plutonium source neutrons. For all the calculations, the following assumptions were used:

• The neutron emissions were from plutonium and its decay product, M1Am, in the form of an oxide, which is the most stable chemical form. Nitrates, oxylates, hydrated oxides, and hydroxides would produce similar neutron yields. Neutrons from other transuranics were ignored. This results in conservative estimates for the mass of plutonium.

• There is little interference from other plutonium compounds, such as fluorides, or from alpha-neutron reactions with low-atomic-number elements, such as beryllium, boron, aluminum, or sodium. It is difficult to judge whether there is a significant contribution from alpha-neutron reactions without more sophisticated measuring equipment that can measure two neutrons in coincidence from a single fission event. (Alpha-neutron events would produce only random coincidences.)

• Each plutonium source is treated separately and does not interact with other plutonium sources (i.e., there is no multiplication).

• Any multiplication effects were ignored, which is a good assumption for gram quantities of plutonium. There may have been some increase in the neutron emissions from the 108-g plutonium sample used as the verification standard, but any errors introduced will be small in comparison with other sources of error or uncertainty.

• Any plutonium holdup is low-exposure plutonium with very little ^"Pu. The weapons-grade plutonium from LLNL (ash sample RA 146A) is assumed to be representative of the plutonium

4.3 in the debris. This is a reasonable assumption and will produce conservative estimates. Any plutonium present with a higher 240Pu content will emit more neutrons per gram; consequently, the mass of plutonium will be overestimated.

4.2.2 Neutron Background

Before the measurements on the ducts and pipes were initiated, some measurements were made to determine the neutron backgrounds of the periphery of the area. When a source was located, two measurements were made: one measurement with the front face of the detector pointing towards the suspected plutonium source, and a second measurement with the slab detector pointing away from the source. Measurements were made with the slab neutron detector oriented at right angles to the suspected source, i.e., the long axis of the slab was perpendicular to the source. In general, the effective "field of view" was about 3 to 4 ft at a distance of 1 to 2 ft from the suspected plutonium source. The following discussion demonstrates how it is possible to estimate the effects of room background with the slab detector oriented at right angles to a suspected plutonium source in the ducts or piping.

Consider the example of the bottom prefilter box on the west side of the SERF cell in the 327 Building basement. The slab neutron detector was positioned directly to the side of the prefilter box at a distance of 12 in. from the surface of the iron shield placed around the box. With the front face of the slab detector pointed towards the prefilter box, the measured neutron count rate was 8.84 + 0.27 cts/s. When the detector was rotated 180° so that the front of the detector was pointed away from uie filter, the measured count rate was 4.48 ± 0.19 cts/s. We can be reasonably certain mat there were no neutron sources directly in front of the slab detector when it was pointed away from the filter box.

Fission neutrons entering from the back of the detector were counted with about 53 % of the efficiency of neutrons entering from the front face, as shown by the experimental data for angular efficiency discussed in Section 3.4.3. Let x represent die counts from the source and y represent the counts from room background. The count rate measured with the vertical slab detector pointed toward the suspected source is the sum of the counts from neutrons entering the front and 50% of the neutrons entering from the back:

x + 0.50y = 8.84 cts/s (4-4)

Likewise, the count rate measured with the vertical detector pointing away from me prefilter box and facing the room is the sum of the neutrons entering from the front (or room background neutrons) and 50% of the neutrons entering from the back side:

y + 0.50y = 4.48 cts/s (4-5>

4.4 Solving the two simultaneous equations for x and y gives:

y = 0.08 cts/s for neutrons from room background x = 8.80 cts/s for neutron emanating from the filter box.

Thus, we must correct the measured count rates by subtracting away 0.08 cts/s due to the neutron background in the room. This is a selected example in which the plutonium in the filter box is well isolated and the count rates are much higher than the general room background. In every other measured location, the count rate with the slab detector pointed towards the suspected plutonium source is similar to the count rate with the detector pointed away from the source. The background-corrected count rates are given in the data table in the spreadsheet calculations in Appendix C.

4.2.3 Calculated Mass of Plutonium

Once the contribution from room background is calculated (or estimated from measurement in the general area), it is possible to use Equation 4.3 to estimate the effects of room scatter and determine the mass of weapons-grade plutonium (6% 240Pu) in the source, if the distance to the center of the plutonium source is known. In the case of the prefilter box, discussed in the previous section, the dimensions of the box were 31.5 in. long by 31 in. high: the box was shielded by 2.5 in. of iron. The center of the slab neutron detector was 12 in. from the surface of the iron shield covering the filter box. If we assume that the plutonium is uniformly spread over the filter, the plutonium source can be modeled as a point source at a distance of 31.5/2 + 2.5 +12 = 30.25 in. (2.52 ft) from the center of the detector. Substituting a distance of 2.52 ft into Equation 4.3 gives:

0.507 + _f±_L = 4.24 cts/s/g (4.6) (2.52)2

Dividing the measured count rate of 8.80 cts/s by 4.24 cts/s/g of Plutonium gives a value of 2.1 g of plutonium in the filter box. This is the highest value measured anywhere in the ducts. It may be somewhat suspect because alpha-neutron reactions from finely divided alpha-emitters on borosilicate glass fiber can produce elevated neutron levels. Because the neutron emission rates from higher- exposure plutonium are higher than for weapons-grade plutonium, the value of 2.1 g of plutonium is an upper limit on the amount of plutonium that may be in the prefilter box.

4.5 5.0 Results

The results of this survey indicate no significant levels of plutonium reside in the ductwork or RLWS lines in the 327 Building.

5.1 Neutron Measurement Results

Cell duct neutron measurements were taken at equal distances along the cell duct using the neutron slab detector. Corrections for overlapping fields were applied to the data. Two sets of plutonium mass estimates were determined using an assumed weapons-grade isotopic composition for the maximum plutonium mass and a reactor fuel isotopic composition (11 % 240Pu) for the most likely mass. Table 5.1 details the results of the neutron measurements on the cell ducts. Table 5.2 details the plutonium mass estimates in the ducts and piping from the SERF cell. Table 5.3 lists the neutron measurement results for the RLWS line. Special attention was given to "hot spots" with high gamma activity, which were thought to be fuel particles in the RLWS pipe.

Table 5.1. Plutonium Holdup in Cell Ducts

Maximum Most Likely Location Plutonium (g)(a) Plutonium (g)(t A-Cell 0.79 0.43 B-Cell 0.24 0.13 C-Cell 0.45 0.25 D-Cell 0.64 0.35 E-Cell 1.22 0.67 F-Cell 0.65 0.35 G-Cell 0.08 0.04 H-Cell 1.44 0.79 I-Cell 1.71 0.93 Total in Ducts 7.22 3.94

(a) Maximum and most likely plutonium mass esti• mates are based on 6 and 11 weight percent 240Pu, respectively.

5.1 Table 5.2. Plutonium Holdup in SERF Cell Piping and Ducts

Maximum Most Likely Location Plutonium (g)(a) Plutonium (g)

SERF Cell Pipe Locations SCE1 0.16 0.09 SCE2 0.24 0.13 SCE3 0.49 0.27 SCE4 0.19 0.10 SCE5 0.24 0.13 SCE6 0.2 0.11 SCE7 0.47 0.26 SCE8 0.16 0.09 S2-3 0.36 0.20 S4-5 0.24 0.13 S6 0.15 0.08 S7 0.19 0.10 S8 0.36 0.20 S9 0.28 0.15 S10 0.3 0.16 S25 0.11 0.06 S26 0.34 0.19 S27 0.21 0.11 in pipes 4.69 2.56

SERF Cell Duct Locations SD1 0.38 0.21 SD2 0.11 0.06 SD3 3.31 1.81 SD4 0.23 0.13 Total in ducts 4.03 2.20 Total in SERF cell 8.72 4.76

(a) Maximum and most likely plutonium mass estimates are based on 6 and 11 weight percent 240Pu, respectively.

5.2 Table 53. Plutonium Holdup in Radioactive Liquid Waste System Pipeline

Maximum Most Likely Location Plutonium (g)w Plutonium (g)w Diverter 0.13 0.07 1 0.05 0.03 2 0.05 0.03 3 0.03 0.02 4 0.02 0.01 5 . 0.03 0.02 6 0.01 0.01 7 0.00 0.00 8 0.05 0.05 9 0.06 0.03 10 0.08 0.03 11 0.08 0.04 12 0.12 0.07 13 0.11 0.06 14 0.08 0.04 15 0.08 0.04 16 0.16 0.09 17 0.19 0.10 18 0.24 0.13 19 0.21 0.11 20 0.18 0.10 21 0.16 0.09 22 0.14 0.08 23 0.16 0.09 24 0.15 0.08 25 0.08 0.04 26 0.05 0.03 27 0.09 0.05 28 0.15 0.08 29 0.17 0.09 30 0.11 0.06 31 0.02 0.01 32 0.02 0.01 33 0.01 0.01 34 o.oo 0.00 35 0.10 0.05 36 0.15 0.08 37 0.10 0.05 38 0.12 0.07 39 0.36 0.20 Total in RLWS Line 4.18 2.28

(a) Maximum and most likely plutonium mass estimates are based on 6 and 11 weight percent 240Pu, respectively.

5.3 Totals for all neutron measurement locations are

Plutonium maximum 20.12 g Plutonium most likely 10.97 g

Other neutron measurement locations not included in the overall total are listed in Table 5.4. These measurements were taken upstairs directly under the hot cells and are suspect because of neutron streaming directly from the hot cells through the openings for the ducts.

Table 5.4. Estimation of Plutonium in Exhaust Ducts at Accessible Locations Before Penetrations into the Basement

Maximum Most Likely In-cell Location Plutonium (g)(a) Plutonium (g)(<0 Plutonium (g) B-Cell 2.0 1.1 44.9 C-Cell 34.4 18.8 187.0 D-Cell 67.1 36.6 43.0 E-Cell 45.0 24.5 40.0 Totals 148.5 81.0 314.9

(a) Maximum and most likely plutonium are based on 6 and 11 weight percent 240Pu, respectively.

5.2 Gamma Measurement Results

A summary of curie content in the various ducts is presented in Table 5.5. These data show the activity levels of each duct by curie content and percent of total. F-cell duct is the "hottest" of all the cell ducts.

Individual cell ductwork activity profiles are presented in the following tables. The appendix contains figures that show the position of the measurements along the ducts leading from each cell to the HEPA filter banks. The gamma dose rates were measured with cutie-pie survey instruments at 120° intervals around the circumference of each duct at 2-ft (0.6-m) intervals, as shown in Table 5.6.

The curie contents of the diverter box and the RLWS lines are shown in Tables 5.7 and 5.8, respectively.

5.4 Table 5.5. Cell Ductwork Curie Content By Direct Gamma Ray Measurement Using a Collimated High-Purity Germanium Detector

Activity in Curies Percent Cell 137Cs "'Co mEu 134Cs Total Total A 1.23E-02 O.OOE+00 O.OOE+00 0.00E+00 1.23E-02 4.9 B 3.23E-04 5.89E-04 O.OOE+00 O.OOE+00 9.13E-04 0.4 C 2.23E-02 O.OOE+00 0.00E+00 3.49E-04 2.27E-02 9.0 D 4.71E-02 O.OOE+00 O.OOE+00 1.71E-03 4.88E-02 19.4 E 3.34E-03 3.58E-05 1.50E-04 4.20E-05 3.57E-03 1.4 F 1.33E-01 2.23E-03 1.45E-03 1.44E-03 1.38E-01 55.0 G 1.23E-04 8.86E-05 0.00E+00 O.OOE+00 2.12E-04 0-1 H 2.39E-03 O.OOE+00 5.96E-05 9.28E-04 3.38E-03 1.3 I 2.09E-02 9.20E-06 2.23E-05 4.12E-05 2.10E-02 8.4 Total 2.42E-01 2.95E-03 1.68E-03 4.51E-03 2.51E-01 Percent 96.4 1.2 0.7 1.8 1000.0

Table 5.6. Gamma Dose Rates and Estimated Activity of Gamma-Emitters in the Ducts Leading from the Cells to High-Efficiency Particulate Air Filters

Dose Rate (mrem/h) Average Distance Position Every 120° Dose Rate from top Estimated Curie Content

154 Location A B C (mrem/h) (ft) ' 137Cs ""Co Eu 134Cs A-cell A-1 500 650 700 620 0.5 1280 0.0 0.0 0.0 A-2 . 250 650 2700 1200 2.5 4990 0.0 0.0 0.0 A-3 250 1500 500 750 4.5 3120 0.0 0.0 0.0 A-4 300 250 400 320 6.5 1320 0.0 0.0 0.0 A-5 100 100 100 100 8.5 415 0.0 0.0 0.0 A-6 70 70 70 70 10.5 291 0.0 0.0 0.0 A-7 150 150 150 150 12.5 624 0.0 0.0 0.0 A-8 100 20 20 47 13.5 97 0.0 0.0 0.0 A-9 45 45 45 45 14.5 94 0.0 0.0 0.0 A-10 25 25 35 28 .16.5 118 0.0 0.0 0.0 179 346 472 330 12300 0.0 0.0 0.0 3320

5.5 Table 5.6. (contd)

Dose Rate (mrem/h) Average Distance Position Every 120° Dose Rate from top Estimated Curie Content Location A B C (mrem/h) (ft) 137Cs "to 154Eu ,34Cs B-cell B-1 12 10 10 11 0.5 10 18 0.0 0.0 B-2 10 8 10 9 2.5 17 32 0.0 0.0 B-3 12 ' 10 10 11 4.5 20 36 0.0 0.0 B-4 15 15 15 15 6.5 28 51 0.0 0.0 B-5 30 20 25 25 8.5 47 85 0.0 0.0 B-6 30 40 30 33 10.5 62 113 0.0 0.0 B-7 30 30 30 30 12.5 56 102 0.0 0.0 B-8 20 15 20 18 13.5 17 31 0.0 0.0 B-9 20 20 20 20 14.5 19 34 0.0 0.0 B-10 30 20 25 25 16.5 47 85 0.0 0.0 21 19 19 20 323 590 0.0 0.0 197 C-cell C-l 35 30 40 35 0.5 461 0.0 0.0 7 C-2 40 35 50 42 2.5 1100 0.0 0.0 17 C-3 50 50 50 50 4.5 1318 0.0 0.0 21 C-4 60 90 60 70 6.5 1845 0.0 0.0 29 C-5 70 70 85 75 8.5 1977 0.0 0.0 31 C-6 55 70 60 62 10.5 1625 0.0 0.0 25 C-7 50 55 70 58 12.5 1537 0.0 0.0 24 C-8 70 90 65 75 14.5 1977 0.0 0.0 31 C-9 100 200 150 150 16.5 3954 0.0 0.0 62 C-10 90 150 150 130 18.5 3426 0.0 0.0 54 C-11 40 40 40 40 20.5 1054 0.0 0.0 16 C-12 50 40 50 47 21.5 615 0.0 0.0 10 C-13 30 30 30 30 22.5 395 0.0 0.0 6 C-14 40 40 40 40 24.5 1054 0.0 0.0 16 56 71 67 64 22338 0.0 0.0 349 903 D-cell D-l 75 75 100 83 0.5 250 0.0 0.0 9 D-2 120 150 150 140 2.5 842 0.0 0.0 31 D-3 300 1000 500 600 4.5 3607 0.0 0.0 131 D-4 250 700 350 433 6.5 2605 0.0 0.0 95 D-5 500 900 600 667 8.5 4008 0.0 o:o 146 D-6 3000 2500 3000 2833 10.5 17033 0.0 0.0 619 D-7 3000 2500 3000 2833 12.5 17033 0.0 0.0 619 D-8 75 75 75 75 14.5 451 0.0 0.0 16 D-9 150 100 100 117 15.5 351 0.0 0.0 13 D-10 100 100 100 100 17.5 601 0.0 0.0 22 D-ll 60 50 50 53 19.5 321 0.0 0.0 12 694 741 729 721 47102 0.0 0.0 1711 7935

5.6 Table 5.6. (contd)

Dose Rate (mrem/h) Average Distance Position Every 120° Dose Rate from top Estimated Curie Content Location A B C (mrem/h) (ft) 137Cs "to 154Eu 134Cs E-ceU E-l 15 20 15 17 0.5 35 0 2 0 E-2 25 20 25 23 2.5 98 1 4 1 E-3 100 70 40 70 4.5 295 3 13 4 E-4 90 95 50 78 6.5 330 3 15 4 E-5 70 35 35 47 8.5 197 2 9 2 E-6 120 700 200 340 10.5 1433 15 . 64 18 E-7 150 150 150 150 12.5 632 7 28 8 E-8 5 5 5 5 14.5 21 0 1 0 E-9 20 15 15 17 15.5 35 0 2 0 E-10 10 10 10 10 17.5 42 0 2 0 E-ll 60 50 50 53 19.5 225 2 10 3 60 106 54 74 3344 36 150 42 810 F-Cell F-1 1600I 1000 1000 1200 0.5 20239 339 226 224 F-2 700 500 1000 733 2.5 24736 414 277 274 F-3 500 700 500 567 4.5 19114 320 213 212 F-4 500 300 1000 600 6.5 20239 339 226 224 F-5 500 2000 1000 1167 8.5 39353 659 439 436 F-6 50 50 50 50 10.5 1687 28 19 19 F-7 120 100 120 113 11.5 3823 64 21 21 F-8 80 80 80 80 12.5 2698 45 15 15 F-9 40 20 40 33 14.5 1124 19 13 12 454 528 532 505 133015 2227 1449 1437 4543 G-Cell G-l 10 10 7 9 0.5 6 4 0.0 0.0 G-2 8 9 8 8 2.5 10 7 0.0 0.0 G-3 9 9 9 9 4.5 11 8 0.0 0.0 G-4 8 9 9 9 6.5 11 8 0.0 0.0 G-5 7 9 9 8 8.5 10 7 0.0 0.0 G-6 30 30 30 30 10.5 37 27 0.0 0.0 G-7 15 15 15 15 11.5 10 ' 7 0.0 0.0 G-8 15 15 15 15 12.5 10 7 0.0 0.0 G-9 15 15 15 15 14.5 18 13 0.0 0.0 13 13 13 13 123 89 0.0 0.0 118

5.7 Table 5.6. (contd)

Dose Rate (mrem/h) Average Distance Position Every 120° Dose Rate from top Estimated Curie Content Location A B C (mrem/h) (ft) 137Cs ""Co 154Eu 134Cs H-cell H-1 8 4 6 6 0.5 51 0.0 1 20 H-2 8 12 6 9 2.5 148 0.0 4 58 H-3 4 6 6 5 4.5 91 0.0 2 35 H-4 4 7 5 5 6.5 91 0.0 2 35 H-5 6 10 7 8 8.5 131 0.0 3 51 H-6 8 15 12 12 10.5 200 0.0 5 78 H-7 15 10 15 13 12.5 228 0.0 6 89 H-8 20 15 20 18 14.5 314 0.0 8 122 H-9 60 60 60 60 15.5 514 0.0 13 200 H-10 15 15 20 17 18.5 428 0.0 11 166 H-11 6 6 6 6 20.5 103 0.0 3 40 H-12 5 5 5 5 22.5 86 0.0 2 33 13 14 14 14 2388 0.0 60 928 164 I-Cell 1-1 20 35 20 25 0.5 368 0 0 1 1-2 30 25 25 27 2.5 785 0 1 . 1 1-3 60 45 45 50 4.5 1472 1 1 3 1-4 75 65 65 68 6.5 2011 1 2 4 1-5 300 300 350 317 8.5 9320 4 10 18 1-6 85 350 90 175 10.5 5151 2 5 10 1-7 45 45 45 45 12.5 1324 1 1 3 1-8 15 20 15 17 13.5 245 0 0 0 1-9 8 8 8 8 14.5 118 0 0 0 1-10 4 4 4 4 16.5 118 0 0 0 64 90 67 73 20912 9 22 41 735

Table 5.7. Diverter Box Curie Profile (Ci)

137,C s 60C, o 1541E u 134€, s 4.81E-2 2.08E-3 4.08E-3 2.38E-3

5.8 Table 5.8. Radioactive Liquid Waste System Pipeline Curie Profile

RLWS Dose Radionuclide Activity (Ci) Line Rate Location (mR/h) 137Cs "Co mCs 154Eu Diverter 450 5.78E-02 7.20E-04 4.20E-04 1.08E-03 1 40 5.14E-03 6.40E-05 3.73E-05 9.60E-05 2 40 5.14E-03 6.40E-05 3.73E-05 9.60E-05 3 20 2.57E-03 3.20E-05 1.87E-05 4.80E-05 4 15 1.93E-03 2.40E-05 1.40E-05 3.60E-05 5 10 1.28E-03 1.60E-05 9.33E-06 2.40E-05 6 10 1.28E-03 1.60E-05 9.33E-06 2.40E-05 7 8 1.03E-03 1.28E-05 7.46E-06 1.92E-05 8 8 1.03E-03 1.28E-05 7.46E-06 1.92E-05 9 10 1.28E-03 1.60E-05 9.33E-06 2.40E-05 10 20 2.57E-03 3.20E-05 1.87E-05 4.80E-05 11 40 5.14E-03 6.40E-05 3.73E-05 9.60E-05 12 80 1.03E-02 1.28E-04 7.46E-05 1.92E-04 13 100 1.28E-02 1.60E-04 9.33E-05 2.40E-04 14 80 1.03E-02 1.28E-04 7.46E-05 1.92E-04 15 40 5.14E-03 6.40E-05 3.73E-05 9.60E-05 16 40 5.14E-03 6.40E-05 3.73E-05 9.60E-05 17 40 5.14E-03 6.40E-05 3.73E-05 9.60E-05 18 80 1.03E-02 1.28E-04 7.46E-05 1.92E-04 19 40 5.14E-03 6.40E-05 3.73E-05 9.60E-05 20 20 2.57E-03 3.20E-05 1.87E-05 4.80E-05 21 20 2.57E-03 3.20E-05 1.87E-05 4.80E-05 22 30 3.85E-03 4.80E-05 2.80E-05 7.20E-05 23 30 3.85E-03 4.80E-05 2.80E-05 7.20E-05 24 30 3.85E-03 4.80E-05 2.80E-05 7.20E-05 25 30 3.85E-03 4.80E-05 2.80E-05 7.20E-05 26 35 4.50E-03 5.60E-05 3.27E-05 8.40E-05 27 30 3.85E-03 4.80E-05 2.80E-05 7.20E-05 28 25 3.21E-03 4.00E-05 2.33E-05 6.00E-05 29 25 3.21E-03 4.00E-05 2.33E-05 6.00E-05 30 25 3.21E-03 4.00E-05 2.33E-05 6.00E-05 31 25 3.21E-03 4.00E-05 2.33E-05 6.00E-05 32 20 2.57E-03 3.20E-05 1.87E-05 4.80E-05

5.9 Table 5.8. (contd)

RLWS Dose Radionuclide Activity (Ci) Line Rate Location (mR/h) 137Cs '"Co 134Cs 154Eu 33 15 1.93E-03 2.40E-05 1.40E-05 3.60E-05 34 10 1.28E-03 1.60E-05 9.33E-06 2.40E-05 35 10 1.28E-03 1.60E-05 9.33E-06 2.40E-05 36 10 1.28E-03 1.60E-05 9.33E-06 2.40E-05 37 10 1.28E-03 1.60E-05 9.33E-06 2.40E-05 38 20 2.57E-03 3.20E-05 1.87E-05 4.80E-05 39 20 2.57E-03 3.20E-05 1.87E-05 4.80E-05 40 20 2.57E-03 3.20E-05 1.87E-05 4.80E-05 Total 2.10E-01 2.61E-03 1.52E-03 3.91E-03

5.10 6.0 References

Brackenbush, L. W., and L. G. Faust. 1970. Neutron Spectra of Plutonium Compounds. BNW-1262, Battelle Northwest, Richland, Washington.

Brackenbush, L. W., W. D. Reece, and J. E. Tanner. 1984. Neutron Dosimetry at Commercial Nuclear Plants. NUREG/CR-3610, U.S. Nuclear Regulatory Commission, Washington, D. C.

Los Alamos National Laboratory (LANL). 1992. Nuclear Safeguards Training Program: "Non• destructive Assay of Special Nuclear Materials Holdup" (presented May 4-8, 1992) and "Non• destructive Assay Techniques for Safeguards Practitioners" (presented August 10-14, 1992). Los Alamos National Laboratory, Los Alamos, New Mexico.

Reiliy, D., N. Ensslin, and H. Smith, Jr. 1991. Passive Nondestructive Assay of Nuclear Materials. NUREG/CR, LA-UR-732, U.S. Nuclear Regulatory Commission, Washington, D.C.

Schwartz, R. B., and C. M. Eisenhauer. 1982. Procedures for Calibrating Neutron Personnel Dosimeters. NBS Special Publication 633, National Bureau of Standards, Washington, D.C.

Stoddard, D. H., and H. E. Hootman. 1971. ^Cf Shielding Guide. E. I. DuPont de Nemours, Aiken, South Carolina.

U.S. Nuclear Regulatory Commission (NRC). 1983. In Situ Assay of Residual Holdup. U.S. Nuclear Regulatory Guide 5.37, U.S. Nuclear Regulatory Commission, Washington, D.C.

U.S. Nuclear Regulatory Commission (NRC). 1984. In Situ Assay of Plutonium Residual Holdup. U.S. Nuclear Regulatory Guide 5.23, U.S. Nuclear Regulatory Commission, Washington, D.C.

6.1 Appendix A

Basic Holdup Assay Measurements Procedure MCA-510 Appendix A

Basic Holdup Assay Measurements Procedure MCA-510

1. The assay system components are set up in the laboratory and pretest verifications are performed.

2. Perform energy calibration and measurement control procedure, completing Exhibit A (see Figure A.l).

3. Perform calibrations for the assay scenario.

4. Move instrumentation to location and verify that no damage has occurred. Perform an energy calibration and measurement control procedure, completing Exhibit A.

5. Determine ambient background^) at the counting location and evaluate counting technique(s).

6. Perform measurements with intermittent control checks, record data and location, and assign non• destructive assay (ND A) log numbers.

7. Transfer data and analyze, using software acquisition and analyses codes and computer spread• sheets. Retain the original data for archive, logging the file name on the Exhibit B, Pacific Northwest Laboratory (PNL) NDA Analysis Log/Report (see Figure A.2).

8. Evaluate the data results and determine the appropriate correction factors, including the detector filters. Initial basic assumptions are made of the source material, one being that the sample self- attenuation is negligible. Actual attenuation correction factors can be applied after measurements are made and more reasonable assumptions are determined as to the material matrix and configuration. All counts measured are corrected by the attenuation correction factors (CF).

Generally, a density value in grams per cubic centimeter (g/cm3) is derived for the item based on gross weight of the container minus the container weight and/or shielding and/or packaging materials. This matrix density value is used for the attenuation corrections for the gamma energies measured.

9. Perform the analysis, review the results, and issue a report.

A.l PNL NONDESTRUCTIVE ASSAY MEASUREMENT PROCEDURE

Procedure No.: NDA-510 Revision: 1 Effective Date: July 26, 1994 Page 31 of 34

EXHIBIT A

M & TE SET-UP AND MCA ENERGY CALIBRATTON CHECKLIST

Date Location Work Pkg #. Operator. .Time.

List Assayed Items (s) and NDA Log #: / / ./. / ./. / /

SET-UP CHECKLIST

(1) Review RWP, CSP 0 (2) Gamma Detector Filled With LN, Stabilized () (3) Filters on Detector Y / N List () W Gamma Detector # () (5) HV Supply Settings; Pos Neg _ (6) Neutron Detector Mod/Ser. CO HV Supply Settings; Pos Neg

ENERGY CALIBRATION (7) Spectroscopy Amplifier Settings: (10) ADC Settings Model No. Model/Serial No. Course Gain 30 Gain 4096 Threshold AUTO Range 4096 Polarity Input NEG Digital Off Set NONE Polarity Output POS Coincidence Gate H NONE Shaping Time 2/isec Fine Gain (11) MCA Identification Shaping Multiplier Model/Serial No. (8) Energy of Sources & Peak Settings (12) Neutron System °Co, 1332^ kev .CH2665 Model/Serial No.. 662.0 kev .CH1324 Counter Model/Serial No. 152•Eu, , 344.0 kev .CH 688 Threshold Setting IS*Eu , 265.0 kev .CH 530 ^Eu, 1275.0 kev .CH2550 133 Ba> 81.0 kev . CH 162 (13) Gamma Measurement Control 133Ba , 356.0 kev . CH 712 Initial PreTest Count Time (9) Systems Adjustments and Noise Peak Energy Net Area With Oscilloscope at 10 jtisec/div ( )

Check Pole Zero () Final Post Test Count Time Check Peak Shape () Peak Energy Net Area

SPECIALIST REVIEW/DATE.

Figure A.l. Calibration Checklist

A.2 PNL NONDESTRUCTIVE ASSAY MEASUREMENT PROCEDURE

Procedure No.: NDA-510 Revision: 1 Effective Date: July 26, 1994 Page 32 of 34

EXHIBIT B

PNL NDA ANALYSIS ASSAY LOG/REPORT

DATE:. NDA LOG*.

TRU U MFP Activation SAMPLE* .

WEIGHT #/G. MEASUREMENT SYSTEMS/METHODS:

Gamma Energy Analysis GEA Calorimeter _^ Gamma Isotopics Passive Neutron Coincidence Weil Counting _ Active Neutron Well Counting Passive Neutron Counting / Spectrometry Segmented Gamma Assay (Transmission Corrected) Combined Neutron Coincidence/Segmented Gamma Assay (Transmission Corrected) Far Field Assay Techniques Barrels Process Hold-up, HEPA Filters Waste Boxes __ Project Specific Other

Measurement information Dose Rate

ASSAY RESULTS: Report attached Y/N Issue Date.

ISOVER Report Y/N Issue Date.

DATA ARCHIVED: Disk(s) Directory / Prefix

File Names _"

Specialist Review _

Figure A.2. Assay Log/Report Form

A.3 Appendix B

Data Sheets for Radiological Survey IUUU) tlB (JBilV 301S 3StniA3t) 33S

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Date Time Purpose of Survey: (I Routine r ^Demand C£U- Oxer &x(uzy Room(s) Building Area RWP Number fy\5Ert&jr~ 321 3O0 3Z1-2D Map or Sketch of Location ALL Ref\&-»i

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EXPOSURE RATE - mR/hr (Thousands)

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B.28 IP ..Battelle OHAWH uy •i-. YOUNG

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LOCATION FROM CEILING - FEET D A + B o C A AVERAGE Radiation Protection Record Survey Report Number Battelle RADIOLOGICAL SURVEY REPORT Pacific Northwest Laboratories Oate Time Purpose of Survey: [ ] Routine 3-17-fi ^Demand C&0- Oact SvJ2.\)Z Room(s) Building Area 'X- RWP Number gAS£fllE>)> 327 3CO 327- ZO Map or Sketch of Location kL 0MJWAS LoO.TtO-4 c OS! (3) OS 2 ^L (3) (*) Ds3 (*L &) f*> • OS1/ (n) OSS ® (s) <*> OS <» 6s) D5-7 6s) f

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RPT Signature k ^ | Reviewed By RP Supervisor

PNL Routine Survey (llrii)

B.31 Radia'.ion Pio'.etiion Reee!d Survey Report Number Battelle RADIOLOGICAL SURVEY REPORT ISOMW^

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PNL Routine Survey (11/22)

B.32 Radiation Protection Record Survey Report Number Battelle Pacific Northwest Laboratories RADIOLOGICAL SURVEY REPORT /S04V8 Date Time Purpose of Survey: 7-7- «.;«* Sli- 2o ">*/»« ^&> SIH - 10 i"»/rf«. SIS-- 25(**/»«•. S20 - 110 /MA /*£ Sit — 8o /**•/»&•

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    B.33 327 BUILDING HOLDUP H-CELL

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    LOCATION FROM CEILING - FEET DA + B O C A AVERAGE 327 BUILDING HOLDUP I-CELL tuu 350 KA 300 x \ iK \ 250 -

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    LOCATION FROM CEILING - FEET D A + B O C A AVERAGE 327 BUILDING HOLDUP RLWS - Gamma Measurements

    X CO \ OJ DC ON fc

    Location D mR/hr Appendix C

    Neutron Measurement Results Appendix C

    Neutron Measurement Results

    This appendix contains the spreadsheet analysis of the neutron counting data using the methods outlined in Section 4.2. Basically, the count rate is determined, then corrected for neutron back• ground using the method explained in Section 4.2.2. Then, a correction is made to compensate for room scatter and distance from the source, using Equation 4.6. Finally, the background-corrected count rate is divided by the counts/second/gram of plutonium to derive the mass of plutonium.

    C.l Plutonium Inventory in 327 Building Ducts and Pipes II 1 1 Estimates of neutron background based on front/back ratios Let | 1 I | . x = count rate from neutrons entering detector from front y = count rate from neutrons entering detector from bade side F = count rate with front face towards source 8 = count rate with back face towards source 1 1 1 Count rate with front of detector facing source F = x + 0.5y I I Count rate with back of detector towards the source B=0.5x*y 1 Solving for x and y gives: Example at position 9: X = (2/3)(2F-8) Source = 0.977778 y = (2/3)(2B-F) 8kgnd= 0.777778 1 Measurements made with detector facing RLWS line and pointed away from line 1 1 1 1 1 • 1 ROI = channel 287 to 2047— Counting efficiency = 80% of neutron events Therefore, increase count rate by factor of 1/0.8= 1.25 1.25X Back• Distance Cts/sec/ Grams J RLWS Detector Time Cnt. Rate Source ground (feet) gramPu ofPu Position Orientatjo Counts ' (sec) (c/s) (c/s) (c/s)

    • 1 divbox 248 100 2.48 1.60 0.88 1.417 12.31 0.13 2 182 100 1.82 0.94 0.88 1.167 17.91 0.05 3 177 100 1.77 0.89 0.88 1.167 17.91 0.05 4 146 100 1.46 0.58 0.88 1.167 17.91 0.03 5 125 100 1.25 0.37 0.88 1.167 17.91 0.02 6 135 100 1.35 0.47 0.88 1.167 17.91 0.03 7 108 100 1.08 0.20 0.88 1.167 17.91 0.01 7bkg 1753 2000 0.8765 8

    9 9F 164 150 1.37 0.98 1.167 17.91 0.05 98 152 150 1.27 0.78 10 10F 177 150 1.43 1.04 1.167 17.91 0.06 10B 166 150 1.38 0.86 11 11F 170 150 1.42 1.42 1.167 17.91 0.08 11B 68 150 0.57 -0.19 12 12F 294 150 Z4S Z13 1.167 17.91 0.12 12S 204 150 1.70 0.63 13 13F 279 150 2.33 1.95 1.167 17.91 0.11 13B 207 150 1.73 0.75 14 14F 201 150 1.68 1.41 1.167 17.91 0.08 14B 148 150 1.23 0.53 15 15F 201 150 1.68 1.38 1.167 17.91 0.08 158 153 150 1.23 0.58

    16 16up 1128 300 4.70 16RT 883 300 3.68 £93 1.167 17.91 0.16 16RA 711 300 Z96 1.50 16LT 951 300 3.96 Z59 16LA 971 300 4.05 2.75 17 478 150 3.98 3.33 1.30 1.167 17.91 0.19

    18 18F 507 150 4.23 4.23 1.167 17.91 0.24 188 - 236 150 1.97 -0.19

    19 449 150 3.74 3.76 0.13 1.167 17.91 0.21

    20 20F 386 150 3.22 3.15 1.167 17.91 0.18 20B 205 150 1.71 0.13

    C.2 21 401 150 3.34 287 0.47 1.167 17.91 22 409 150 3.41 Z47 0.94 1.167 17.91 0.14

    23 23F 416 150 3.47 2.69 1.167 17.91 0.15 23B 347 150 2.83 1.54 24 24F 492 150 4.10 277 1.167 17.91 0.15 i 248 485 150 4.04 2.66 25 2SF 397 150 3.31 1.44 1.167 17.91 0.08 25B - 535 150 4.46 3.74

    26 419 150 3.49 0.86 2.63 1.167 17.91 0.05 27 372 150 3.10 1.58 1.52 1.167 17.91 0.09

    28 28F 349 150 2.91 2.70 1.167 17.91 0.15 28B 212 150 1.77 0.42 29 29F 388 150 3.23 3.12 1.167 17.91 0.17 298 215 150 1.79 0.23 30 30F 265 150 221 1.88 1.167 17.91 0.11 308 5100 4000 1.S9 0.65

    31 267 150 223 0.41 1.81 1.167 17.91 0.02 32 254 150 2.12 0.30 1.81 1.167 17.91 0.02 bkg-wall 2903 2000 1.81 33 222 150 1.85 0.04 1.81 2417 4.56 0.01 34 214 150 1.78 -0.03 1.81 1.375 13.04 0.00

    35 35F 186 150 1.55 1.31 1.375 13.04 0.10 35B 137 150 1.14 0.49

    36 222 150 1.85 0.94 0.91 2 6.43 0.15 37 241 150 2.01 0.68 1.33 2 6.43 0.10 38 300 150 2.50 0.74 1.76 2 6.43 0.12

    39 39F 406 150 3.38 229 2 6.43 0.36 398 399 150 3.33 218

    total grams P\i- 3.83

    C.3 Corrected Count Rate (c/s)

    RLWS Source Backgnd Position

    1 1.6 0.88 - 2 0.94 0.88 3 0.89 0.88 4 0.58 0.88 Towards-Away Count Contributions 5 0.37 0.88 6 0.47 0.88 4.5 I 7 0.2 0.88 4 • 9 0.98 ' 0.78 10 1.04 0.86 11 1.42 0 3.5 12 Z13 0.63 13 1.95 0.75 14 1.41 0.53 15 1.38 0.58 ^zs J Z76 a 16 Z13 o 17 3.33 1.3 S 2- 18 4.23 0 19 3.76 0.13 O £ 1.5 • 20 3.15 0.13 o 21 2.87 0.47 u 22 2.47 0.94 1 • 23 Z69 1.54 24 Z77 Z66 0.5 • 25 1.44 3.74 i 26 0.86 Z63 27 1.58 1.53 — 28 Z7 0.42 0 1U

    - RLWS Grams Position Plutonium

    1 0.13 2 0.05 3 0.05 Grams of Plutonium in RLWS Line 4 0.03 5 0.02 • 0.40 r 6 0.03 7 0.01 0.35 9 0.05 10 0.06 0.30 11 0.08 12 0.12 S 13 0.11 3 • J 14 0.08 "5 0.25 15 0.08 1 16 0.16 20 17 0.19 £ °- 18 0.24 a 19 0.21 • rvi A 1 (0 20 0.18 15 21 0.16 i °- 22 0.14 2 23 0.15 (9 24 0.15 25 0.03 26 0.05 0.10 k/ Mj 27 0.09 0 5 10 15 20 25 30 35 40 28 0.15 0.05 23 0.17 Position along Line 30 0.11 31 0.02 0.00 32 0.02 33 0.01 34 0.00 35 0.10 36 0.15 37 0.10 38 0.12 39 0.36

    C.5 Grams of Plutonium in RLWS Line

    0.40

    0.35 -

    0.30 • E 3 *C 0.25 n 2 as 2 °- 0.20 (• o « sE 0.15 0.10

    0.05

    0.00 ^ 10 15 20 25 30 35 40 Position along Line

    Appendix D

    Gamma Measurement Results Appendix D

    Gamma Measurement Results

    This appendix contains the results of gamma measurements along the radioactive liquid waste sys• tem (RLWS) piping that runs under the hot cells in the basement along the length of the 327 Building. The locations numbers correspond to marking every 2 to 4 feet along the pipe where radiological sur• vey measurements were performed. See page B.3 of Appendix B for the approximate location of the RLWS piping. Note that the zero position corresponds to the diverter box in the south-east corner of the basement.

    The figures show the curie content of the RLWS line given in Table 5.8. Almost all of the gamma activity is associated with 137Cs, as shown in the first figure. The activity associated with other radionuclides is a factor of 100 times less, as shown in the second figure.

    D.l 0 5 10 15 20 25 30 35 40 Location

    o Co60 * Cs134 x Eu154

    Figure D.l. 327 Building Holdup in Radioactive Liquid Waste System Piping, Gamma Measurements for ^Co, 134Cs, and 154Eu 100 T

    80 Cs-137

    60 p I 40 .--

    20

    HI—•—• ,

    10 15 20 25 30 35 40 RLWS Location

    Figure D.2. 327 Building Holdup in Radioactive Liquid Waste System Piping, Gamma Measurements for 137Cs PNL-10787 UC-610 Distribution

    No. of No. of Copies Copies

    OFFSITE 17 Pacific Northwest Laboratory

    • 12 DOE/Office of Scientific and M. J. Bagaglio P7-75 Technical Information L. W. Brackenbush K3-70 D. J. DesChange ONSITE D. L. Haggard (5) K3-70 J. F. Henderson P8-24 DOE Richland Operations Office W. A. Hoober P8-24 J. D. Jensen P8-24 J. E. Trevino P7-35 S. B. Johnston P8-24 G. L. Ketner P8-24 D. E. Knowlton P7-35 J. M. Seay P8-24 R. L. Spinks P8-24 R. W. Stevens P8-24

    Distr.l