Review of Information for 

Spent Nuclear Fuel 

Burnup Confirmation 

Office of Nuclear Regulatory Research 

NUREG/CR-6998 
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Review of Information for 

Spent Nuclear Fuel 

Burnup Confirmation 

Manuscript Completed: June 2009 

Date Published: December 2009 

Prepared by 

B.B. Bevard, J.C. Wagner and C.V. Parks (ORNL) 

M. Aissa (NRC) 

Oak Ridge National Laboratory 

Managed by UT-Battelle, LLC 

Oak Ridge, TN 37831-6170 

M. Aissa, NC Project Manager 

NRC Job Code Y6517 

Office of Nuclear Regulatory Research 

NUREG/CR-6998 

iii 

ABSTRACT 

The Interim Staff Guidance on burnup credit (ISG-8, revision 2) for pressurized-water-reactor spent 

nuclear fuel in storage and transport casks, issued in 2002 by the U.S. Nuclear Regulatory Commission’s 

Spent Fuel Project Office, recommends an out-of-core burnup measurement to confirm the reactor record 

and compliance with the assembly burnup value used for cask loading acceptance. This recommendation 

is intended to prevent unauthorized loading (i.e., misloading) of assemblies due to inaccuracies in reactor 

burnup records and/or improper assembly identification, thereby ensuring that the appropriate subcritical 

margin is maintained. The purpose of this report is to detail information and issues relevant to 

preshipment burnup measurements when using burnup credit in pressurized-water-reactor spent nuclear 

fuel transport and storage casks. In particular, this report reviews the role of burnup measurements in the 

regulatory guidance for demonstrating compliance with burnup loading criteria, burnup measurement 

capabilities and experience, generation and accuracy of utility burnup records, fuel movement and 

misloading experience, and the consequences of misloading assemblies in casks designed for burnup 

credit. 
iv 
v 

CONTENTS 

Page 

ABSTRACT.................................................................................................................................................iii

LIST OF FIGURES ....................................................................................................................................vii

LIST OF TABLES.......................................................................................................................................ix

EXECUTIVE SUMMARY .........................................................................................................................xi

ABBREVIATIONS ....................................................................................................................................xv

1 INTRODUCTION..................................................................................................................................1

2 ROLE OF BURNUP MEASUREMENTS IN DEMONSTRATING COMPLIANCE WITH 

BURNUP LOADING CRITERIA .........................................................................................................3

3 OUT-OF-CORE BURNUP MEASUREMENTS—EXPERIENCE AND CAPABILITIES.................5

3.1 MEASUREMENT TECHNIQUES ..............................................................................................5

3.2 PRACTICAL AND OPERATIONAL CONSIDERATIONS.......................................................6

3.3 ISSUES AFFECTING MEASUREMENT EQUIPMENT ACCURACY ....................................6

3.4 MEASUREMENT SYSTEMS USED AT COMMERCIAL NUCLEAR PLANTS....................8

3.4.1 Fork Detector Systems........................................................................................................9

3.4.2 High-Resolution Gamma-Ray Spectroscopy....................................................................14

3.4.3 Summary of U.S. Out-of-Core Measurement Results ......................................................15

3.5 CERENKOV RADIATION MONITORING DEVICES ...........................................................17

3.6 COMMENTS ON INTERNATIONAL PRACTICES AND REGULATIONS .........................19

3.6.1 Summary of Current Practices and Regulatory Requirements .........................................20

3.6.2 Summary...........................................................................................................................23

4 REACTOR BURNUP RECORDS: GENERATION AND ACCURACY ..........................................25

4.1 BACKGROUND.........................................................................................................................25

4.2 THERMAL POWER MEASUREMENT ...................................................................................26

4.3 IN-CORE POWER DISTRIBUTION MEASUREMENT .........................................................27

4.4 IN-CORE MEASUREMENTS PROCESSING..........................................................................29

4.4.1 Flux Map Processors.........................................................................................................29

4.4.2 Reactor Analysis Codes ....................................................................................................31

4.4.3 Regular Validation of Reactor Analysis Codes ................................................................32

4.5 IN-CORE BURNUP MEASUREMENT ACCURACY.............................................................34

4.6 SUMMARY ................................................................................................................................38
vi 

CONTENTS (CONTINUED) 

Page 

5 CONSEQUENCES OF FUEL ASSEMBLY MISLOADING.............................................................39

6 FUEL MOVEMENT AND MISLOADING EXPERIENCE...............................................................41

7 SUMMARY DISCUSSIONS...............................................................................................................47

7.1 OUT-OF-CORE MEASUREMENTS.........................................................................................47

7.2 REACTOR BURNUP RECORDS..............................................................................................48

7.3 MISLOADS IN A PWR BURNUP CREDIT CASK..................................................................48

7.4 MISLOADING EVENTS ...........................................................................................................49

7.5 ISSUES RELATED TO OLDER SPENT NUCLEAR FUEL....................................................50

7.6 METHODS FOR CALCULATING BURNUP ..........................................................................50

8 OBSERVATIONS................................................................................................................................53

9 REFERENCES.....................................................................................................................................57
vii 

LIST OF FIGURES 

Page 

Fig. 1.1. Illustrative loading curve superimposed on the U.S. discharged 

pressurized-water-reactor (PWR) fuel inventory ......................................................................2

Fig. 3.1. Neutron/gamma-ray detector head, referred to as the “fork” ....................................................9

Fig. 3.2. Schematic showing the method for making measurements in a spent fuel 

pool using the fork detector system.........................................................................................11

Fig. 3.3. Log-Log best-fit plot of neutron signal vs reactor record burnup at Arkansas 

Nuclear One Unit-1 .................................................................................................................12

Fig. 4.1. In-core instrumented locations in a three-loop Westinghouse pressurized 

water reactor............................................................................................................................28

Fig. 4.2. Typical Westinghouse pressurized water reactor in-core flux map system. ...........................29

Fig. 4.3. Comparison of measured vs predicted axial power distribution at an instrumented 

core location ............................................................................................................................31

Fig. 4.4. Measured vs RW-859 reported assembly burnup values ........................................................36

Fig. 6.1. Drawing of a boiling water reactor fuel assembly with assembly identification number. ......42

Fig. 6.2. Picture of the top of a pressurized water reactor fuel assembly in a spent fuel pool 

showing the unique assembly identification number. .............................................................42
viii 
ix 

LIST OF TABLES 

Page 

Table 3.1. Advantages and disadvantages of various burnup characterization techniques..........................8

Table 3.2. Burnup verification systems comparison..................................................................................17

Table 3.3. Intensity measurement of six fuel assemblies...........................................................................19
x 
xi 

EXECUTIVE SUMMARY 

This report is only applicable to pressurized-water-reactor (PWR) spent nuclear fuel (SNF). No 

information was evaluated for boiling water reactors. 

Criticality safety analyses for commercial PWR spent fuel storage and transport have historically made 

the conservative assumption that the SNF is fresh (unirradiated), with uniform isotopic composition 

corresponding to the maximum allowable enrichment. Hence, the criticality-safety-related loading 

criteria are based on as-manufactured fuel design specifications. While this “fresh fuel” assumption 

provides a conservative, well-defined approach to the criticality safety analysis that is not dependent on 

the fuel operating history, it ignores the decrease in reactivity that occurs as a result of irradiation. Taking 

credit for the decrease in reactivity due to irradiation is termed burnup credit. In contrast to safety 

analyses that use the fresh fuel assumption, burnup credit analyses necessitate consideration of the fuel 

operating history, additional validation of calculational methods (due to prediction and use of SNF 

nuclide compositions), consideration of new conditions or configurations for the licensing basis, and 

additional measures to ensure proper cask loading. A “burnup credit cask” refers to a cask where burnup 

credit is implemented in the design and safety analysis. A burnup credit cask can enable an extended 

range of allowable contents and/or increased cask capacities but requires knowledge of the SNF 

characteristics commensurate with the criteria (e.g., minimum discharge burnup) for approved contents as 

established in the safety analysis. 

The current Nuclear Regulatory Commission (NRC) guidance for use of burnup credit in transport and 

storage casks is limited to PWR SNF and was issued in 2002 as interim staff guidance by the NRC 

Division of Spent Fuel Storage and Transportation (formerly the Spent Fuel Project Office). This interim 

staff guidance (referred to as ISG-8, rev. 2) includes a recommendation for a measurement to confirm 

knowledge of a key PWR SNF characteristic related to reactivity (e.g., discharge burnup) and help 

prevent loading of a more reactive assembly than that permitted for the approved burnup credit cask 

contents. A burnup measurement is recommended, in part, to verify the accuracy and completeness of 

utility reactor burnup records for all SNF and allay concerns associated with the potential for 

calculational, procedural, and other human errors in selection of assemblies for loading. 

The purpose of this report is to detail information and issues relevant to preshipment burnup 

measurements when using burnup credit in PWR SNF storage and transport casks. In particular, this 

report reviews the role of burnup measurements in the regulatory guidance for demonstrating compliance 

with burnup loading criteria, burnup measurement capabilities and experience, generation and accuracy of 

utility burnup records, fuel movement and misloading experience, and the consequences of misloading 

assemblies in casks designed for burnup credit. 

Over the past 20 years, there have been multiple SNF out-of-core examinations conducted at U.S. utilities 

to compare utility records for assembly burnup and cooling time with out-of-core measurement data. The 

measurement equipment used [neutron/gamma detection systems (e.g. the “Fork Detector” system from 

Los Alamos National Laboratory) and high-resolution gamma-ray spectrometry (HRGS) systems 

available commercially from several vendors] was developed to detect fissile material diversion and was 

not specifically designed for ease of use in reactor spent fuel pools or optimized to accurately verify or 

quantify SNF burnup independent of utility reactor records. However, the data gathered from these 

examination campaigns have been fairly uniform and consistent and are useful for evaluating the 

effectiveness of available measurement techniques. 

The following observations, also provided in Section 8, highlight key points from the information 

reviewed for this report. 
xii 

Reactor Burnup Records

• Utility records for fuel burnup are based on either (1) the measured core thermal output, 

with burnup distributed to individual assemblies using validated computer codes, or (2) a 

combination of information provided by in-core detectors, measured core thermal output, 

and validated computer codes. 

• A significant amount of data from 1980 to the present is available to support a finding 

that utility records for fuel burnup are accurate for individual spent fuel assemblies to at 

least 5% of “true” assembly burnup. These data primarily originate from previous ex-core 

burnup measurement programs, comparisons between calculated burnup values (on 

which reactor record values are based) and burnup values inferred from in-core 

measurements, and retrospective evaluations based on comparisons between existing 

reactor record values and calculated values. 

• Utilities do not all use the same methods to calculate and verify the assembly burnup 

values that are recorded in their reactor records. Moreover, the computational methods 

used by many utilities have evolved over time such that burnup values for older and 

newer fuel are based on different methods. Hence, for SNF assemblies to be transported 

in burnup credit casks, utilities need to demonstrate how the burnup values in their 

reactor records were developed and recorded and document information as to the 

accuracy of their recorded burnup values. Examples of this type of activity are the 

reactor record verification programs described herein by Duke Energy and the Tennessee 

Valley Authority (TVA). 

• In at least one case, burnup values in a utility’s reactor records for some assemblies are 

based on the “batch average” burnup, which is definitely not an accurate representation of 

the individual burnup for each of the assemblies. The number of utilities that may have 

used batch average data is unknown. However, based on a review of various data sources 

for discharged fuel, it is expected that the use of batch averaged values is limited in 

number and would apply to older fuel records. Contemporary reactor records use 

individual, assembly-specific burnup calculations and values. 

• Unless a quality assurance program is implemented to ensure the accuracy of the reactor 

records, reactor records may become less accurate over time due to transcription errors as 

record media degrade or are changed to newer media or as software is updated. 

Burnup Verification Measurements

• Out-of-core measurement systems can adequately verify reactor record burnup 

information. However, as is also the case for in-core measurement systems, out-of-core 

measurement systems cannot measure fuel burnup directly. Instead, these out-of-core 

measurement systems measure gamma-ray and/or neutron emissions from the assemblies, 

which are then compared to a calibration curve to develop an estimated fissile content 

and corresponding assembly burnup. 

• The burnup measurement programs reviewed for this report concluded that out-of-core 

measurement systems provided somewhat less accurate burnup values, as compared to 

reactor record burnup values developed using reactor in-core monitoring systems and 

design codes. The modern in-core systems, which use core measurements and some data 
xiii 

from design codes, produce burnup values that are generally within approximately 2% of 

the burnup values predicted by design codes, whereas the out-of-core measurement 

systems produce average assembly burnup results that are within the expected fork 

detector accuracy range of 2.2–5% of “true” burnup. However, there were some fork 

detector examinations with maximum assembly deviations as high as 9.1%. Regardless, 

these examination programs demonstrated that out-of-core measurements could identify 

substantially underburned fuel assemblies and hence could be used to prevent them from 

being accidentally loaded into spent fuel transport casks. 

• Careful calibration against an assembly of known burnup, known cooling time, and 

identical geometry is required to achieve the reported accuracies with the fork detector 

used in the U.S. studies. Thus, in practice, these out-of-core measurements are dependent 

on and calibrated against reactor burnup records. Consequently, the burnup values 

inferred from these measurements tend to have a higher uncertainty than reactor record 

assembly burnup values determined using in-core measurements. 

• Fuel assembly axial-burnup profiles have a significant impact on reactivity and are 

therefore an important component in determining “average assembly burnup.” For the 

fork detector examination programs evaluated in this report, personnel used 

measurements at the assembly centerline (midplane) and assumed that the axial profile of 

the reference assembly could be used to estimate the assembly average burnup (i.e., it 

was assumed that the axial profiles were the same for the reference and measured 

assemblies). This approach could give erroneous burnup measurement results for 

assemblies that have different axial burnup characteristics. Hence, if out-of-core 

measurements are used, care should be taken to ensure that correct average burnup 

information is collected, commensurate with the measurement accuracy goals and criteria 

for approved contents. 

• The costs and risks associated with performing out-of-core burnup measurements should 

be balanced with the risks and potential costs of not performing the measurements. Outof-core measurement campaigns require utility resources for planning and execution, 

increase the dose to personnel, increase the risk of damage to assemblies and potential 

fuel mishandling events due to the increased assembly movements, and have associated 

financial cost to the utility. The risks and potential costs associated with loading a 

significantly underburned assembly into a transport or storage configuration have not 

been explored in this report. 

• The neutron-counting measurement systems appear to be more accurate than the HRGS 

systems and require less skilled operators for handling the SNF assemblies. 

• Considerable efforts, primarily motivated by interests related to nuclear material 

safeguards, are ongoing to develop better and more accurate measurement systems. 

Consequences of Fuel Assembly Misloading 

• The consequences to keff of loading assemblies that have slightly reduced burnup (e.g., 

5%) due to uncertainties in the burnup verification process), as compared with the 

required burnup, are fairly small (≤ 1%). On the other hand, loading one or more highly 

enriched (i.e., >4 wt %) fresh fuel assemblies has a significant consequence on criticality 

safety. These findings suggest that while it may not be overly important to precisely 
xiv 

verify the burnup value, it is important to ensure that fresh or very-low-burnup (nearly 

fresh) fuel assemblies are not misloaded into a burnup credit cask. 

Fuel Movement and Operational Considerations

• Although utilities’ record of reliability in selecting and moving assemblies during fuelhandling processes may be characterized as “good” from the standpoint that no 

inadvertent criticalities have occurred as a result of fuel misloading, there are a number of 

documented examples of fuel misloading. Considering the large number of fuel 

assembly movements that have been executed, relatively few mishandling events 

involving movement of an incorrectly identified fuel assembly have been reported. 

Procedural violations related to SNF movements that resulted in violations of plant 

Technical Specifications have occurred with a fairly low frequency. Most of these events 

were not a result of incorrect burnup values assigned to the SNF but were instead the 

result of personnel error in selecting assemblies for movement. 

• Fresh fuel is visibly different from SNF due to the oxidation, crud buildup, and 

bending/twisting of the latter. Visual inspection should easily differentiate new 

assemblies from SNF. There is some uncertainty about the appearance of an assembly 

with very limited burnup (e.g., removed promptly after start-up due to leaking fuel rods 

or some other problem) after it has resided in a spent fuel pool for a number of years. 

• Visual inspections and/or simple field measurements (e.g., gross radiation measurements, 

Cerenkov radiation detector) could be performed during cask loading to detect and 

prevent accidental loading of fresh or nearly fresh fuel. A basic detector system could be 

devised to ensure that an assembly has some minimum activity level. One examination 

tool that may suffice is the digital Cerenkov viewing device (DCVD). Any measurement 

program would have to be evaluated to determine its limitations to detect SNF. For 

example, the DCVDs may not be able to differentiate between very old, moderately 

burned SNF and very-lightly-burned (i.e., nearly fresh) SNF. Also, any examination 

limitations, such as equipment alignment for Cerenkov examinations, must be evaluated 

to ensure that adequate examination results are achievable for all SNF to be evaluated. 
xv 

ABBREVIATIONS 

ALARA as low as reasonably achievable

ANO Arkansas Nuclear One

ANS American Nuclear Society

ANSI American National Standards Institute

ARMP Advanced Recycle Methodology Program

B&W Babcock & Wilcox

BWR boiling water reactor

CoC Certificate of Compliance

CVD Cerenkov viewing device

DCVD digital Cerenkov viewing device

EPRI Electric Power Research Institute

GWd/MTU gigawatt-days per metric ton of uranium initially loaded into an assembly

HRGS high-resolution gamma-ray spectroscopy

IAEA International Atomic Energy Agency

ID identification (number)

keff effective neutron multiplication factor

LER Licensee Event Report

LWR light water reactor

NCTL Nuclear Component Transfer List

NRC Nuclear Regulatory Commission

pcm percent mille, which means thousandths of a percent

pcm = 1 × 10-5 Δk/k

PWR

QA

pressurized water reactor

quality assurance

RE residual enrichment 
xvi 

SCSS Sequence Coding and Search System (NRC)

SDM

SNF

shutdown margin, as in reactor shutdown margin

spent nuclear fuel

SNM special nuclear material

TMI Three Mile Island

TVA Tennessee Valley Authority

235U uranium isotope with 143 neutrons

UV ultraviolet

ZPPT zero power physics testing
1

1 INTRODUCTION 

This report is only applicable to pressurized-water-reactor (PWR) spent nuclear fuel (SNF). No 

information was evaluated for boiling water reactors (BWRs). 

Criticality safety analyses for commercial PWR spent fuel storage and transport have historically made 

the conservative assumption that the SNF is fresh (unirradiated), with uniform isotopic composition 

corresponding to the maximum allowable enrichment.1

 Hence, the criticality-safety-related loading 

criteria are based on as-manufactured fuel design specifications. While this “fresh fuel” assumption 

provides a conservative, well-defined approach to the criticality safety analysis that is not dependent on 

the fuel operating history, it ignores the decrease in reactivity that occurs as a result of irradiation. Taking 

credit for the decrease in reactivity due to irradiation is termed burnup credit. In contrast to safety 

analyses that use the fresh fuel assumption, burnup credit analyses necessitate consideration of the fuel 

operating history, additional validation of calculational methods (due to prediction and use of SNF 

nuclide compositions), consideration of new conditions or configurations for the licensing basis, and 

additional measures to ensure proper cask loading. A “burnup credit cask” refers to a cask where burnup 

credit is implemented in the design and safety analysis. A burnup credit cask can enable an extended 

range of allowable contents and/or increased cask capacities but requires knowledge of the SNF 

characteristics commensurate with the criteria (e.g., minimum discharge burnup) for approved contents as 

established in the safety analysis. 

The reduction in SNF reactivity due to irradiation is addressed in the U.S. Nuclear Regulatory 

Commission (NRC) Standard Review Plan for Transportation Packages for Spent Nuclear Fuel, 

NUREG-16172

 and Interim Staff Guidance–8, revision 2, Burnup Credit in the Criticality Safety Analyses 

of PWR Spent Fuel in Transport and Storage Casks (ISG-8, rev. 2),3

 which provides recommendations for 

the acceptance of a burnup credit approach in the criticality safety analysis of PWR spent fuel casks. The 

products of a burnup credit safety evaluation typically include loading curves, which specify cask-loading 

criteria in terms of the minimum required burnup as a function of initial assembly enrichment. A loading 

curve represents combinations of burnup and initial enrichment that correspond to a limiting value of the 

effective neutron multiplication factor (keff) for a given configuration (e.g., a cask). Figure 1.1 is an 

example of a loading curve. Assemblies with insufficient burnup, as compared with the loading curve, 

are not acceptable for loading. Misloading of an underburned (i.e., relative to the loading curve) fuel 

assembly into a cask will cause an increase in total reactivity. The extent of the reactivity increase is 

dependent on several factors but is dominated by the amount by which the actual assembly burnup is less 

than the minimum burnup value for loading acceptance and the position of the assembly in the cask. It is 

worth noting at this point that the number or percentage of assemblies that do not meet the burnup 

requirement for loading, and hence have the potential to be misloaded, is dependent on the location of the 

loading curve relative to the discharged SNF inventory, ranging from a fairly large percentage for loading 

curves corresponding to actinide-only burnup credit and high-capacity casks (see for example Fig. 1.1) to 

a fairly small percentage or even zero for loading curves corresponding to full (i.e., actinides and fission 

products) burnup credit and/or low capacity casks. 

Recognizing the importance of assembly burnup as a loading criterion with burnup credit, NUREG-1617 

and ISG-8 both include a section to address the issue titled “Assigned Burnup Loading Value.” The 

following excerpt from ISG-8, rev. 2, which was released in September 2002, provides the most recent 

guidance available. 

Administrative procedures should be established to ensure that the cask will be loaded with fuel 

that is within the specifications of the approved contents. The administrative procedures should 

include a measurement that confirms the reactor record for each assembly. Procedures that 

confirm the reactor records using measurement of a sampling of the fuel assemblies will be 
2

considered if a database of measured data is provided to justify the adequacy of the procedure in 

comparison to procedures that measures each assembly. 

The measurement technique may be calibrated to the reactor records for a representative set of 

assemblies. For confirmation of assembly reactor burnup record(s), the measurement should 

provide agreement within a 95% confidence interval based on the measurement uncertainty. The 

assembly burnup value to be used for loading acceptance (termed the assigned burnup loading 

value) should be the confirmed reactor record value as adjusted by reducing the record value by a 

combination of the uncertainties in the record value and the measurement.3

The recommendation for a measurement (generally accepted to be an out-of-core measurement) is 

included to ensure that a burnup credit cask is not incorrectly loaded with a more reactive assembly 

(including a fresh fuel assembly) than that permitted by the loading criteria. 

The purpose of this report is to detail relevant information and issues related to preshipment burnup 

measurements when using burnup credit in PWR SNF transport and storage casks. In particular, this 

report reviews the role of burnup measurements in the regulatory guidance for demonstrating compliance 

with burnup loading criteria, burnup measurement capabilities and experience, generation and accuracy of 

utility burnup records, the consequences of misloading assemblies in casks designed for burnup credit, 

and fuel movement and misloading experience. 

Fig. 1.1. Illustrative loading curve superimposed on the U.S. discharged 

pressurized-water-reactor (PWR) fuel inventory. 
3

2 ROLE OF BURNUP MEASUREMENTS IN 

DEMONSTRATING COMPLIANCE WITH 

BURNUP LOADING CRITERIA 

A Certificate of Compliance (CoC) for a spent fuel transportation cask contains technical requirements 

and operating conditions (fuel specifications, cask leak testing, surveillance, and other requirements) for 

the cask design and specifies the authorized contents for the cask system. Fuel specifications for 

authorized contents that are influenced by criticality safety considerations include fuel assembly design 

and initial fuel enrichment. Uncertainties and/or tolerances in the assembly design parameters and initial 

fuel enrichment are considered and, where appropriate, accounted for in the criticality safety evaluation. 

The specifications listed in the CoC are all verified before assembly loading, typically via comparison 

with utility records for the assemblies. With burnup credit, fuel burnup (and cooling time) becomes a 

parameter that must be included in the CoC, typically in the form of a loading curve (see Fig. 1.1), and 

verified prior to assembly loading. As with any parameter that affects safety (e.g., initial fuel 

enrichment), burnup must be verified to ensure compliance with the safety criteria. However, unlike 

initial fuel enrichment, which has a well-defined, as-manufactured value with a known uncertainty, 

burnup is a measure of the power production of a fuel assembly during its residence in a reactor core. 

Hence, the issues for burnup verification are as follows: (1) the assembly burnup value and (2) the 

uncertainty in the assembly burnup value. 

Two means for verifying assembly burnup include (1) comparison with reactor record data, which include 

assembly burnup from core-follow calculations supplemented by in-core flux measurements (to varying 

extents) and (2) out-of-core, post-irradiation burnup measurements. Out-of-core burnup measurements 

provide a “defense-in-depth” backup or verification of the reactor record data to prevent unauthorized 

loading (i.e., misloading) of assemblies due to inaccuracies in the reactor burnup records and/or improper 

assembly identification, thereby ensuring that the appropriate subcritical margin is maintained. The 

burnup measurements are recommended in the current regulatory guidance to verify that SNF assemblies 

meet the cask loading criteria and confirm the proper fuel assembly selection before loading. Although 

available out-of-core measurement techniques must be calibrated against reactor burnup records, they can 

provide a means to quantify the uncertainty in the reactor record data that can subsequently be used to 

determine an assembly burnup value to be used for loading acceptance (termed the “assigned burnup 

loading value” in ISG-8). The assigned burnup loading value is the confirmed reactor record value as 

adjusted by reducing the record value by a combination of the uncertainties in the record value and the 

measurement. Hence, large uncertainties in the measured and/or reactor burnup records will result in 

higher values of burnup being required for loading (for a given initial enrichment) and, subsequently, a 

reduced loading acceptance (as compared with the unadjusted burnup value; see Fig. 1.1). 
4
5

3 OUT-OF-CORE BURNUP MEASUREMENTS—

EXPERIENCE AND CAPABILITIES 

3.1 MEASUREMENT TECHNIQUES 

As it is not possible to measure assembly fissile content or burnup directly, existing burnup measurement 

equipment measures neutron and gamma radiation emitted from an assembly and uses this information, 

along with calibration data, to infer the assembly burnup. For example, some of the transuranic elements 

formed in nuclear fuel from successive neutron absorptions have relatively short spontaneous fission 

half-lives and thus emit relatively large numbers of neutrons (n) and/or decay by alpha (α) emission at a 

sufficient rate to produce significant neutrons from the α-n reaction. The rates of such neutron emissions 

have a very strong dependence upon burnup. A complicating aspect of such measurements is that, after 

significant exposure in a reactor, the radiation signatures of interest may be masked by radiation from 

fission products, activated structural components, and transuranic elements that build up as a result of the 

fission process. 

“Active” neutron detection systems measure the enhanced neutron flux from induced fission events that 

occur in the residual fissile material. Some active measurement systems use pulsed sources; others use 

randomly pulsed sources or steady state sources. The sources are typically neutron sources but could also 

be high-energy gamma-ray sources for gamma-fission reactions. The detectors are used to detect prompt 

and delayed neutrons and/or gamma rays. The count rate is proportional to the fissile material in some 

cases, but often the time dependence of the count rate is used to quantify the amount of fissile material. 

Such measurements require calibration standards. In general, active neutron counting systems are not 

typically used in spent fuel pools because the background neutrons produced by the 244Cm in the SNF 

result in relatively long measurement times for convergence of time-dependent detector count rates. The 

convergence of the time-dependent data is directly related to the source-induced fission rate relative to 

that resulting from the inherent fission process as a result of 244Cm and other spontaneous fission isotopes 

that are present in SNF. In research reactors, the “background” neutrons produced by the 244Cm are low 

enough that several active measurement techniques can be used to determine fissile content in the SNF. 

For gamma-ray detection, the events being measured are counts in specific, narrow energy bands that are 

characteristic of selected individual fission products or groups of fission products. The isotopes selected 

for measurement ideally should have a high fission yield, a long half-life, and an energetic gamma whose 

energy falls in a band not shared by other fission products. Because of the short half-life of some of the 

primary gamma-ray emitters, the fission product, 137Cs, which has a half-life of 30.07 years and a 

gamma-ray energy of 661.7 keV, is the major gamma-ray emitter measured after 5 years of SNF cooling. 

Cesium-137 also has the favorable characteristics that its neutron absorption cross sections are negligible 

and its yields from both 235U and 239Pu are approximately the same. Because the gamma rays of interest 

are typically only a small fraction of the total incoming gamma rays, the problems of background and 

background subtraction are significant, requiring longer counting times to obtain adequate statistical 

accuracy within the very narrow energy bands needed for good resolution. 

“Passive” neutron detection systems measure spontaneous fission neutrons from higher actinides, mainly 

244Cm. Passive neutron examination methods have a strong count rate dependency on burnup, cooling 

time, and initial enrichment. Curium-244 has a half-life of 18.1 years and is formed by successive 

neutron captures, beginning with 238U. After 10 years of cooling, the 244Cm provides ~95% of the source 

neutrons and by the 20th year of cooling, still provides nearly 92% of the source neutrons.4

 The neutron 

emission of 244Cm is very sensitive to variations in assembly burnup. Additionally, measurement of the 

244Cm neutrons is very sensitive to many physical factors in the measurement setup (such as equipment 
6

positioning errors) that could affect the neutron count. Factors such as boron concentration of the spent 

fuel pool water must also be known and replicated if the equipment calibrations are not performed in 

conjunction with the SNF measurements. Additionally, because of the relatively short half-life of 244Cm, 

the cooling time of the fuel must be known to correctly interpret the total neutron output. Operatordeclared values for the cooling time must be used. These cooling times can be verified by several 

measuring techniques, such as measuring the total gamma-ray activity of the assembly.5

3.2 PRACTICAL AND OPERATIONAL CONSIDERATIONS 

Relevant issues associated with the use of out-of-core measurements for burnup verification include the 

following: accuracy and reproducibility of the out-of-core measurements and related calculations; 

accuracy of the reactor records used for equipment calibration; accuracy of the burnup measurement 

relative to reactor records; potential increase in probability for fuel-handling accidents due to the increase 

in fuel-handling operations; any additional radiation exposure to personnel; cost of the equipment 

installation and use; and the impact of the measurement activity on other ongoing utility operations, 

including both the duration and degree of interference with other operations (almost all out-of-core 

measurement devices require the use of the operator’s bridge crane). The issues of accuracy and 

reproducibility of the measurements and related calculations, accuracy of reactor records used for 

measurement calibration, accuracy of the burnup measurement relative to reactor records, and potential 

increase in probability for fuel-handling accidents due to the increase in fuel-handling operations are 

discussed in separate sections of this report. 

The impacts of the out-of-core measurement activity on other ongoing utility operations and any 

additional radiation exposure to personnel are important topics and are briefly discussed in this report, but 

adequate data to develop conclusions were not available. Cost information in the form of potential 

cost-benefit analyses is available, but the cost information is highly dependent on assumptions that may 

be easily biased, thus making the analyses very judgmental. For that reason, cost information was not 

included in this report. 

3.3 ISSUES AFFECTING MEASUREMENT EQUIPMENT 

ACCURACY 

The accuracy of any out-of-core burnup measurement device is primarily a function of the equipment’s 

ability to discriminate between the radiation of interest and the background radiation field, the quality of 

the equipment calibration and subsequent equipment operation, and the ability of the operator to properly 

use the equipment and interpret the data in terms of burnup. The radiation data must further be 

interpreted as to what fraction of a fuel assembly it represents and what this fraction implies with respect 

to the whole assembly.4

 When using radiation-monitoring equipment to verify assembly burnup, certain 

parameters must be known or assumed. For example, in the ideal case, a measured count rate would be 

related (via knowledge of detector efficiency) to the total decay rate of the isotope in that part of the 

assembly and thus (via a known isotope half-life) to the total inventory of the fission product isotope. 

Then, via knowledge of the percent fission yield of the isotope, the total number of fissions would be 

known. Via knowledge of the energy yield per fission, the total energy output, and hence the burnup of 

that part of the assembly, would be determined. The ability to achieve high accuracy in the determination 

of the burnup from count rate in the ideal situation depends upon knowledge of four physical parameters 

and three favorable constraints. The physical parameters are detector efficiency, isotope half-life, 

percentage fission yield, and fission energy yield. The constraints are direct fission yield of the isotope, 

very low neutron cross section (avoiding in-core depletion of the fission product isotope), and long halflife (avoiding significant decay of the isotope before measurement). In the absence of accurate 

knowledge of the absolute detector efficiency, a calibration of the counting system can be accomplished if 
7

a geometrically similar fuel assembly of well-known burnup is used as a primary standard. In summary, 

the interpretation of the counting data can be accomplished in the ideal case either through accurate 

knowledge of absolute detector efficiency and three other physical parameters or by calibration of the 

system with a primary standard fuel assembly of known burnup and of nearly identical characteristics 

(except for burnup). The result of the interpretation in this simple ideal case is that the burnup of the 

measured part of the assembly can be determined directly from the count rate when this value is 

compared with the count rate and the known burnup of the standard assembly.4

Considering the situation that is encountered in actual practice, the three favorable constraints of the ideal 

case are not encountered, and the four physical parameters of the ideal case are not all known with 

precision. The actual situation is as follows. 

1. The fission product of interest does not have a significantly long half-life. The principal 

consequence of this is that a single measurement or measurement ratio cannot yield both the 

burnup and the age; two measurements or measurement ratios of isotopes with different time 

behavior patterns are necessary if both assembly burnup and age of a spent fuel assembly are to 

be determined. A second consequence of the time-varying behavior of measured isotopes is that 

a portion of these isotopes will decay before fuel discharge and the decay fractions will be 

different for different core residence times and burnup values. Because the cooling age is 

measured from discharge, corrections for different predischarge decay should be made. The 

practical effect of such corrections is that postdischarge count rates are no longer proportional to 

burnup. 

2. The fission yield of the isotope is not direct. In fact, many of the isotopes of interest may be first 

or second daughters of direct-yield fission products or may not even result from fission—they 

may be the product of one or more serial captures of neutrons. Furthermore, fission yields 

depend on the fissioning material. 

3. Some isotopes of interest may not have small neutron cross sections, and hence, neutron capture 

will transmute that isotope into other isotopes. 

4. Absolute detector efficiency (fraction of total decays that are measured) is very difficult to 

calculate accurately. Some of the other four physical parameters may be known quite accurately. 

Many decay half-lives are known with sufficient accuracy that they do not contribute 

significantly to overall uncertainty. The energy yield per fission (about 200 MeV/fission) is 

somewhat dependent upon the fissioning element but is generally known quite accurately. 

Additionally, any calibration of the counting system will use a geometrically identical fuel assembly with 

an inexact burnup that is taken from utility reactor records. Thus, the preceding discussion conveys some 

of the complexity of the processes influencing burnup measurement that occur when fuel is irradiated 

over extended periods of time (note that total fuel “age” includes irradiation time in the reactor and time 

after assembly discharge from the reactor). Simple models of measurable events cannot, in most cases, 

adequately represent the complex and interdependent processes that are taking place. The measurement 

approach that has evolved is a close coupling of direct calibration with the use of experimentally based 

analytical models of fuel behavior to interpolate between calibrations or, if necessary, extrapolate beyond 

calibrations. 

Table 3.1 summarizes the advantages and disadvantages of various nondestructive measurement 

techniques such as passive neutron counting or gamma-ray spectroscopy.6
8

Table 3.1. Advantages and disadvantages of various burnup characterization techniques 

Technique Advantages Disadvantages

Absolute count rate of the 662-keV 

gamma ray from 137Cs

• Simple linear relationship 

between 137Cs and burnup

• Half-life of 30 years

• Insensitive to variations in 

reactor power rating and 

dwell time

• Absolute measurement 

requires a well-defined and 

reproducible geometry 

between the detectors and the 

fuel assembly

The nuclide activity ratio: 

134Cs/137Cs

• The ratio method makes it 

insensitive to geometry

• Half-life of 134Cs (2.2 years) 

requires significant decay 

correction and can be applied 

only to fuel with cooling time 

< 20 years

• Burnup correlation is 

dependent on initial 

enrichment and power rating

The nuclide activity ratio:

106Ru × 137Cs/(134Cs)2

• Insensitive to geometry

• Independent of enrichment 

and power rating

• Useful only for fuel with 

cooling time < 9 years (106Ru 

has a 372-day half-life)

Passive neutron measurement 

(predominantly from 244Cm)

• The neutron signals are 

received uniformly from all 

pins in the assembly 

(gamma-ray measurements 

are sensitive only to the 

outer pins)

• Good for safeguards 

applications as it is sensitive 

to missing or removed fuel 

pins

• Curium-244 inventory is a 

strong function of initial 

enrichment 

• Neutron assay is very 

geometry sensitive and can 

also be affected by 

multiplication and neutron 

poisons in the pool or within 

the assembly

3.4 MEASUREMENT SYSTEMS USED AT COMMERCIAL 

NUCLEAR PLANTS 

Several spent fuel measurement devices are available to support SNF burnup verification. These devices 

generally use high-resolution gamma-ray spectrometry (HRGS) and passive neutron counting to measure 

both neutron and gamma-ray emissions from the spent fuel assemblies. When used with depletion 

computer codes, these measurement devices can determine the SNF assembly burnup and radionuclide 

inventories.7,8 The “fork” detector system, developed by Los Alamos National Laboratory to support the 

detection of potential fissile material diversion from SNF (Fig. 3.1), is designed to determine the extent of 

the variation among assembly burnup values and to identify any anomalous values.9 The fork system 

uses gross gamma-ray ion chambers in addition to the neutron detectors and has been used at several 

utility spent fuel pools for the independent verification of declared fuel burnup values in connection with 

International Atomic Energy Agency (IAEA) safeguards inspections. With proper calibration and 

corrections, the fork detector can determine the burnup of individual fuel assemblies to an average 

accuracy of about 5% of plant records.5

 The observed data must be corrected for both the variation in 

cooling times among the assemblies and initial assembly enrichment. The cooling time correction is 

accomplished by extrapolating the neutron data (after background subtraction) back to the date of 
9

discharge of each assembly using an exponential factor with a half-life of 18.1 years, the half-life of the 

principle neutron emitter, 244Cm. The initial enrichment correction is required because 244Cm, which 

produces the neutrons, is produced by activation of 238U and is determined by the reactor flux rather than 

the fission rate. This correction (described in detail in Reference 10) is accomplished by using a factor to 

adjust the observed count rates for the variation in initial enrichment among the assemblies, using the 

information in reactor records for initial enrichment.9

 Another similar system, developed by 

Westinghouse, was demonstrated on PWR fuel at the Surry nuclear power station.4

 

Fig. 3.1. Neutron/gamma-ray detector head, referred to as the “fork.” (Source: Reference 8.)

3.4.1 Fork Detector Systems 

3.4.1.1 Calibration Methodology 

Central to the measurement methodology of the out-of-core measurement equipment is use of a reference 

assembly of known burnup and cooling time and of similar geometry to the assemblies being examined. 

The use of this reference assembly is essential because the analysis requires the detection and counting of 

specific events in a fuel assembly of known reference burnup, Br, and known reference age, tr. A 

reference net count rate, Cr, can then be obtained. Subsequently, an assembly of interest can be measured 

that has nominally identical geometry and initial enrichment but an undefined burnup, B, and age t. The 

corresponding net count rate is C. An analytical model is then used to characterize the burnup-dependent 

and time-dependent behavior of the isotope being counted. The basic form of the general-purpose 

characterization methodology for any specific isotope would then be 

p m

r rr

C Bt

C Bt

−

⎛ ⎞ ⎛ ⎞⎛ ⎞ ⎜ ⎟ ⎜ ⎟⎜ ⎟ =

⎝ ⎠ ⎝ ⎠⎝ ⎠

 , (1)
10

where the p and m coefficients are determined directly from the analytical results for any particular 

isotope or group of isotopes over the range between the reference burnup and age and the approximate 

burnup and age of the assembly being measured.4

 Specific values of these coefficients for neutrons, 

gamma-ray, and heat have been published as part of Appendix 1C of DOE/RW-0184-R1 (Ref. 11). 

3.4.1.2 Examinations at ANO and Oconee 

Fork radiation detectors were used at Arkansas Nuclear One (ANO) and the Oconee Nuclear Station to 

verify utility reactor records.12 The fork detectors provided a nondestructive means for characterizing the 

fuel assemblies following discharge and cooling time in the spent fuel pools. The fork detector analysis 

was conducted underwater in the spent fuel pool using a measurement taken at about the assembly 

centerline. [At ANO Unit-1 (ANO-1), a steel band located at the center of each assembly precluded a 

measurement at that level, so the measurement was taken 1 ft above the centerline of each assembly. 

Other measurements showed the neutron and gamma-ray yield to be essentially constant along the central 

section of similar PWR assemblies, extending a few feet on either side of the center.] The count time on 

each assembly was 100 s. The fork detector system used two arms that were placed on either side of the 

fuel assembly facing one another. Each arm of the fork detector contained two fission chambers for 

neutron detection and an ionization chamber for gamma-ray detection to provide information on the 

neutron and gamma radiation fields emanating from the SNF assembly being examined.13 Because the 

gamma-ray detector is less sensitive to variations in burnup, it confirms burnup with approximately 15% 

uncertainty9

 and is thus used only as a general confirmation of assembly cooling time and burnup. 

3.4.1.3 Measurement Data Analysis 

The approach used in the analysis of the SNF at both sites was to accumulate measurements from a 

number of assemblies and generate an internal calibration by comparing each assembly with the 

best-derived fit of all the site data. The self-calibration (best-fit curve) eliminated the uncertainties and 

complications that were introduced by external calibration techniques while retaining the sensitivity to 

detect measurements that were inconsistent with the utility reactor records. The analysis of the fork 

detector data made use of the reactor records for cooling time, burnup, and initial enrichment in such a 

way that errors in any of these parameters were likely to increase the deviation from the best-fit 

calibration curve. During out-of-core examinations, any observed deviations from the best-fit curves 

would incorporate the uncertainties in the out-or-core measurements and any errors in the reactor records. 

The average deviations were therefore likely to be upper bounds on the random errors in the reactor 

records for assembly burnup.9

 To correct the observed data for the variation in cooling times among the 

assemblies, the neutron data (after background subtraction) were extrapolated back to the date of 

discharge of each assembly using an exponential factor of half-life equal to 18.1 years, which is the halflife of the principal neutron emitter, 244Cm. A factor to adjust the observed count rates for the variation in 

initial enrichment among the assemblies was calculated using reactor records.9

At ANO-1, 34 assemblies were measured with the fork system in 1.5 days of operation. The initial 

enrichment of the assemblies ranged from 2.016 to 3.209 wt % 235U. The range of assembly average 

burnup was from 19.9 to 57.3 GWd/MTU, and the cooling times varied from 6.1 to 17.6 years. 

Background counts were generally less than 1% of the signal from the assembly.9,14 At ANO Unit 2 

(ANO-2), measurements were made on 39 standard assemblies with average enrichments ranging from 

1.9 to 3.9 wt % 235U. The range in fuel assembly average burnup was 12.3 to 50.7 GWd/MTU, and the 

cooling times ranged from 3.8 to 13.7 years.9

 Two assemblies that contained neutron sources and had 

been cooled for approximately 2 years were also examined during this period. The fork detector system 

identified the presence of neutron sources in these two assemblies by a rise of 25 to 40% in the signal in 

the neutron detectors at the location of the neutron sources near the midpoint of the assemblies.9
11

The neutron and gamma-ray emissions in the examined assemblies were measured in the spent fuel pool 

by raising each assembly partially out of the storage rack and performing the measurement near the center 

of the assembly (see Fig. 3.2). 

Fig. 3.2. Schematic showing the method for making measurements in a spent fuel pool using 

the fork detector system. (Source: Reference 9.)

The overall uncertainty introduced by the out-of-core measurements with corrections was about 2% of the 

reactor record indicated burnup. At ANO-1, the self-calibration curve was the power law best-fit (least 

squares) to the data and is given by 

N = (C) (B)3.83 , (2)

where N is the neutron count rate in counts per second, B is the burnup in units of GWd/MTU, and C is a 

fitted constant whose value is 0.00100. At ANO-1, the neutron signal was proportional to the 3.83 power 

of the burnup, and at ANO-2, the neutron signal was proportional to the 4.35 power of the burnup. The 

average deviation of the burnup measurements from the best-fit curve of reactor record burnup vs neutron 

signal was 2.7% at ANO-1, with a maximum deviation for a single assembly of 9.1% (see Fig. 3.3). At 

ANO-2, the average deviation of the burnup measurements from the best-fit curve of burnup vs neutron 

rate was 3.5%, with a maximum deviation for a single assembly of 8.6%. This was consistent with the 2 

to 3% random variation among the reactor records for average assembly burnup.9

Because the fork detector measurements were taken only at the assembly centerline and compared with a 

best-fit calibration curve, the true assembly full-length average burnup values were not actually measured. 

The average burnup values for the assemblies were developed as calculated numbers based on the fork 
12

detector response (measured counts) at each assembly centerline, comparison of the detector output with 

the calibration standard, and adjustment of the result for other factors such as assembly cooling times. 

Fig. 3.3. Log-Log best-fit plot of neutron signal vs reactor record burnup at Arkansas Nuclear 

One Unit-1. (Source: Reference 9.)

Another fork detector demonstration program was conducted in 1993 at the Oconee Nuclear Station. In 

conjunction with Sandia National Laboratories and Los Alamos National Laboratory, Oconee staff used 

the fork detector system to measure the burnup of 93 assemblies from Oconee Units 1 and 2 over a 

3.5-day period. The initial enrichment of the assemblies ranged from 2.91 to 3.92 wt % 235U, and the 

average burnup ranged from 20.3 to 58.3 GWd/MTU. The cooling times ranged from 4.2 to 14.8 years. 

The measurements showed an average deviation in burnup from the best-fit curve of about 2.2% with 

enrichment correction and 10% without the correction. Among the 91 assemblies fit by the curve, only 

one assembly deviated by more than 6%. Two assemblies exhibited much higher (factor of 5) neutron 

signals than expected from the burnup records, and a check of the reactor records for these two assemblies 

found that they contained small americium-curium-beryllium primary neutron sources.9

During the fork measurement programs at ANO and Oconee, both Babcock and Wilcox (B&W) and 

Combustion Engineering SNF assemblies were examined. A finding from these measurement activities 

was that the burnup dependence of the neutron signal was specific to each assembly design. For the 

B&W assemblies at ANO-1 and the Oconee Nuclear Station, the neutron signal was proportional to the 

3.83 power of the burnup. For the Combustion Engineering assemblies at ANO-2, the neutron signal was 

proportional to the 4.35 power of the burnup. Thus, a direct comparison of data between assembly types 

is possible, but only after assessing and correcting the data for the impact of the different assembly 

designs. This finding further demonstrated how these out-of-core examinations were dependent on 

accurate and detailed calibration and data interpretation.9
13

The Westinghouse measurement system used at the Surry Power Station in the early 1980s was similar to 

the fork detector system described above in that it used two fission chambers on opposite faces at the 

midplane of a fuel assembly and the associated signal conditioning, recording, and data analysis hardware 

and software. The measurement of an assembly in the reactor spent fuel pool took about 10 min from 

grappling of the assembly in the fuel storage rack; raising, moving, and placing it in the mounting fixture 

on the pool bottom; counting the assembly; and returning the assembly to its original location in the 

storage rack. Measurements were taken only at the midpoint of the fuel assembly. A feature of this 

measurement methodology is that the measurement at the midpoint of the fuel assembly would typically 

have the highest burnup instead of the assembly average burnup. When this work was conducted in the 

1980s, it was assumed that no errors were introduced by this measurement procedure if the peak-toaverage burnup ratio was a constant, independent of burnup. It was assumed that peak-to-average burnup 

ratios were not much greater than unity for high-burnup PWR fuels. Thus, the measurement of burnup in 

the midplane of the assembly would provide a good measure of assembly average burnup in high-burnup 

PWR fuels with relatively uniform axial power distributions.4

 However, near the fuel assembly ends, 

burnup is suppressed due to leakage. Consequently, the majority of PWR SNF assemblies have similar 

axial-burnup profiles (or shapes)—relatively flat in the axial midsection (with peak burnup of ~1.1 times 

the assembly average burnup) and significantly underburned fuel at the ends (with burnup of ~0.5 times 

the assembly average burnup).15 Also, the burnup is slightly higher at the bottom of the assembly than at 

the top due to the difference in the moderator density. The cooler (higher-density) water at the assembly 

inlet results in higher reactivity (which subsequently results in higher burnup) than the warmer moderator 

at the assembly outlet. Other factors, such as control rods or axial power-shaping rods, can also affect the 

axial burnup shape. Thus, measurements at the midplane of a fuel assembly may be used to give a 

reasonable estimate of the assembly average burnup, but may also introduce uncertainty into the burnup 

calculation when determining assembly average burnup.16

In the demonstration at Surry, 50 fuel assemblies were measured, covering a burnup range of 14.41 to 

41.05 GWd/MTU, a cooling time of 0.18 to 8.86 years, an initial enrichment range of 1.86 to 3.40 wt % 

235U, and a considerable variation in fuel power history and cross-assembly burnup gradients. The data 

were analyzed to provide correlations and correction terms for burnup, enrichment, and time dependence 

of the neutron source. The final analysis of results provided an estimated deviation of 0.8 GWd/MTU 

from utility-reported average burnup values on each assembly. Because the utility reactor records have an 

uncertainty of approximately 0.6 GWd/MTU, the neutron-counting technique appeared to provide an 

accurate and sensitive measure of burnup when enrichments and cooling times were known and the 

system was calibrated to identical assemblies of known characteristics.4

The Surry assessment of the neutron-counting system, which included evaluations of two detectors on 

opposite faces vs a single detector, found that the standard deviation nearly doubled with the single 

detector because of nonuniformity of burnup across assemblies and small positioning errors. The 

reproducibility of repeated count-rate measurements of the same assembly was also checked. A 6% 

variation was found for the shortest-cooled fuel, with a high gamma-ray background (0.18 years), but the 

variation appeared to be under 4% for a representative population of fuel.4

Using a reference burnup uncertainty of 2%, a count rate uncertainty of 6% (consistent with the 

6% variation noted above), and a burnup coefficient of 4 (as determined in the Surry measurements and 

confirmed by ORIGEN217 data), the following formula was used as a basis to determine the burnup 

uncertainty at Surry: 

1 m

p p

r

r r

C t B B

C t

⎛ ⎞⎛ ⎞ = ⎜ ⎟⎜ ⎟ ⎝ ⎠⎝ ⎠ 

, (3)
14

where B is burnup, Br is reference burnup, C is the net count rate, Cr is the reference net count rate, t is 

age, tr is reference age, and p and m are analytically (or experimentally) determined coefficients for 

burnup and time dependence. 

Solving the equation for uncertainty in burnup, the equation becomes 

2 2 2 2 2 2 2

2

1 ln ln r r

r rr r

B C Bp tm B C

B B pCC B p t p

⎛ ⎞ δ δ δδ ⎛ ⎞ ⎛ ⎞ ⎛ ⎞ ⎛⎞ δ δ ⎛ ⎞ ⎛⎞ ⎛⎞ ⎛⎞ ⎜ ⎟ = + ++ + ⎜ ⎟ ⎜ ⎟ ⎜ ⎟ ⎜⎟ ⎜ ⎟ ⎜⎟ ⎜⎟ ⎜⎟ ⎝ ⎠ ⎝ ⎠ ⎝ ⎠ ⎝ ⎠ ⎝⎠ ⎝ ⎠ ⎝⎠ ⎝⎠ ⎝⎠

 . (4)

The last two terms allow for the uncertainties in the sensitivity coefficients. However, because these 

terms also depend on the proximity of the burnup values and times of the measured and referenced fuel, 

they were assumed to be small (i.e., zero) relative to the other terms. The uncertainty in burnup was then 

calculated to be 2.9%. Based on the out-of-core measurement program, it was concluded that neutron 

counting of PWR assemblies in a system calibrated with identical assemblies of known burnup and age 

provided a very accurate estimate of assembly burnup. Another conclusion reached in these examinations 

was that the equipment sensitivity imposed significant requirements on the measurement process, because 

a lack of attention to measurement details could quickly generate very inaccurate results.4

3.4.2 High-Resolution Gamma-Ray Spectroscopy 

In addition to the fork detector measurements noted above, ANO also performed five measurement 

campaigns using high-resolution gamma-ray spectroscopy to provide confirmation of the operatordeclared burnup values. Pajarito Scientific Corporation (currently BNFL Instruments, Inc.) performed 

this work between April 1996 and August 1997. The first campaign was a demonstration campaign in 

which 52 assemblies were measured to test and refine the monitoring equipment. Four additional 

campaigns measuring 351 assemblies were later performed. The assembly types chosen were B&W 

15 × 15 assemblies and Combustion Engineering 16 × 16 assemblies. These assemblies had a wide range 

of irradiation histories comprising various burnup values (12.31 to 46.87 GWd/MTU), cooling times 

(2.51 to 20.57 years), and enrichments (1.921 to 3.902 wt % 235U). 

The fuel measurement procedure for these HRGS examinations initially included instrument 

standardization by means of measuring a reference source. However, this produced large errors and 

exposed the test operators to an unnecessary amount of radiation. Instrument standardization was 

eventually achieved by measuring a selected reference assembly. The data collected from the five 

measurement campaigns were not presented in a published report. However, the rough draft analysis of 

the campaign 3 measurement data, as a sample set, suggested that the dependent calibration 1σ

measurement uncertainty may be apportioned as 2.33% to the measurement process and 4.8% to the 

reactor record error to give the observed overall uncertainty, for campaign 3, of 5.37% at the 67% 

confidence interval. The distribution of the data was symmetrical and continuous. The determination of 

cooling time from the activity ratio 134Cs/154Eu was effective in its role in the burnup measurement 

procedure as a cooling time correction parameter. The average cooling time uncertainty was less than 

120 days. The uncertainty was calculated from a combination of a dependent calibration based on the 

correlation between operator-declared data and assembly-measured cooling times and statistical 

uncertainties. 

During the HRGS campaigns, the standard measurement time for the gamma-spectra acquisition was 

about 15 min per assembly. The total measurement time, including fuel-handling time, was 

approximately 1 h per assembly. The handling of the selected fuel assemblies for testing was an offnormal event. To minimize the background source strength, the fuel assemblies had to be moved to a 

location in the pool that was shielded from the bulk of the fuel assemblies. At ANO, the testing was 
15

performed in the cask pit. Fuel assemblies had to be moved to the point where the fuel rod would touch 

the detector. The detector was used as a proximity indicator. In this configuration, there were no limit 

switches that could be used to limit the movement of the fuel-handling machine to prevent inadvertent 

damage to the fuel assemblies. The examination relied heavily upon operator actions. During the 

examinations, the fuel assemblies had to be moved downward along the detector to allow acquisition of 

gamma-ray spectra from a number of points along the length of the assembly. The assembly was 

subsequently rotated 180° about its vertical axis, and a repetition of the gamma-ray scanning was 

conducted. The combination of the scanning and rotation of the assembly maximized the sampling of the 

assembly and minimized the accumulation of systematic errors arising from axial or radial burnup 

variations due to the axial burnup profile or radial tilt. Assembly hang-ups did occur during these 

activities. Additionally, rotating the fuel assemblies was difficult and was not a normal fuel-handling 

practice. During the fuel moves, the operating team received radiation doses even when examining spent 

fuel assemblies that had long cooling times and relatively low burnup values. Based on the potential for 

fuel-handling problems, ANO performed a risk, dose, and cost assessment to evaluate the measurement 

test program. The evaluation suggested that due to the high risk of fuel damage and additional dose and 

the high cost of the examination program, the value associated with performing additional measurements 

was not sufficient to justify continuing the measurement program. At that point, the measurement 

campaigns at ANO were halted. 

However, before the gamma-scan program was halted, three of the ANO-1 fuel assemblies were 

identified as “outliers” by the detector system. Subsequent evaluation of the reactor records for those 

assemblies by ANO staff revealed the possibility that the examination responses may have been perturbed 

by mistakes in the data correlation. Thus, the measurements for these assemblies were considered to be 

flawed, and the reactor burnup records for the three assemblies were assumed to be correct. 

In another documented examination program, Zion Station Reactor 2, cycle 1, data were benchmarked for 

the Electric Power Research Institute (EPRI) Advanced Recycle Methodology Program (ARMP). 

Assemblies with enrichments of 2.25% (Region 1), 2.79% (Region 2), and 3.29% (Region 3) 235U and 

burnup values averaging 19.95, 19.17, and 14.43 GWd/MTU, respectively, were measured using HRGS 

based on 140La decay. Direct burnup accuracies were not calculated for the assemblies at the time the data 

were collected but were later inferred from the data. A subsequent evaluation showed an uncertainty of 

less than 2% between the utility-calculated burnup and the gamma-ray spectroscopy-measured burnup.4,18

3.4.3 Summary of U.S. Out-of-Core Measurement Results 

None of the systems used for out-of-core measurement of SNF burnup actually measure assembly fissile 

content directly. Instead, such systems measure gamma-ray or neutron emissions from nonfissile 

materials in an assembly and then use that information to calculate the assembly fissile content. There 

have been a number of measurement campaigns using various types of equipment, with the majority of 

the examinations using the fork detector system. The available data comparing the use of either HRGS 

equipment or neutron-counting fork detector equipment with utility reactor records are significant but not 

overwhelming. The neutron-counting systems appear to have more accuracy and are less limiting than 

the gamma-ray scanning systems, but they still require the use of reactor record data for burnup, cooling 

time, and initial enrichment as input parameters. Also, the fork detector system requires a geometrically 

identical reference assembly of known burnup and cooling time or the development of a best-fit curve for 

calibration purposes before any examinations are conducted. The “known” burnup of the reference 

assemblies is taken directly from utility reactor records. 

Several of the referenced reports state that the deviation of fork detector system measurements from 

utility records is in the 2.2–5% range.4,9 The specific results noted in Section 3.4.1 of this report show the 

average accuracy of the fork detector when compared with reactor records was between 3.0 and 3.5%, 
16

with maximum assembly deviations of up to 8.6%. Even though the gamma-ray scans were conducted on 

the full length of the fuel assemblies instead of the fuel centerline as was done in the fork detector 

examinations, the high-resolution gamma-ray scans were not as accurate as the fork detector 

examinations. Furthermore, in ANO campaign 3, the gamma-ray scans demonstrated an average 

deviation from plant records of 5.37% at the 67% confidence interval. 

The accuracy of out-of-core measurement systems used for demonstration measurements is adequate for 

the purposes of verifying that fissile content has not been removed/diverted from an assembly (the 

purpose for which the equipment was developed) or for ensuring that a significantly underburned fuel 

assembly is not mistaken for a fully burned assembly and subsequently placed in a transport cask. 

Review of the various measurement campaigns indicates that these out-of-core measurement systems are 

somewhat less accurate than reactor records and, as noted, the results of the out-of-core examinations are 

highly dependent on reactor records. For any out-of-core measurement campaign, the as low as 

reasonably achievable (ALARA) radiation dose implications for the examination personnel, the value 

added by the measurement campaign, the potential increase in fuel mishaps, and the potential impacts on 

other spent fuel pool operations should all be considered. 

Since the out-of-core burnup measurement programs were conducted at Surry and ANO-1, the equipment 

vendors have upgraded their equipment to incorporate lessons learned and improve the sensitivity of the 

equipment; however, the upgraded equipment has not been demonstrated at any U.S. commercial utility 

to measure SNF assembly burnup. A description of the upgraded systems is provided in Table 3.2.19 

Another different measurement approach that is potentially relevant to burnup confirmation, albeit 

approximate, is the measurement of light intensity from Cerenkov radiation, which is discussed in the 

following section. 
17

Table 3.2. Burnup verification systems comparison 

 Westinghouse BNFL EPRI/Sandia 

System Name None specific None specific FORK+ 

System 

Description 

A miniature SiC nuclear 

detector designed for use in 

harsh radiation and temperature 

environments. SiC has several 

advantages for use as a solidstate radiation detector 

material, primarily due to its 

wider band-gap energy 

compared with that for silicon 

or germanium: 

1. Lower leakage current 

2. Operational capability at 

elevated temperatures 

3. Greater radiation resistance 

4. Capable of pulse-mode 

operation in high-radiation 

fields 

Gamma rays and neutrons can 

be measured simultaneously. 

Demonstrated linear response 

to both neutron and gamma-ray 

fluxes. Detector sensitive to 

thermal or epicadmium neutron 

energies. Does not require 

external cooling. 

An HRGS system that 

uses a high-purity Ge 

detector. The detector 

requires nitrogen 

cooling to operate. The 

system demonstrated in 

the United States 

measures only gamma 

radiation. However, 

neutron techniques 

(fission chambers) 

have been incorporated 

into the system used at 

Sellafield (Thorp). 

Demonstrated linear 

response to gamma-ray 

fluxes. 

A gamma-ray spectroscopy 

system which uses a 

CdZnTe (CZT) detector. 

CZT crystal requires 

tungsten shielding to 

improve resolution. In 

addition, a cadmium liner is 

required over the tungsten to 

absorb thermal neutrons to 

keep the thermal neutron 

fluence at the crystal to a 

minimum. (Note: Neutron 

absorption by Cd produces a 

gamma ray; therefore, the 

CZT crystal must be 

protected.) Neutron 

measurement is 

accomplished through the 

use of a fission chamber. 

The CZT crystal is cooled 

thermionically by using a 

microelectronic thermionic 

cooling device. Larger CZT 

crystals would require 

nitrogen cooling. 

Detector Types SiC (6LiF coated) High-purity Ge CdZnTe with tungsten 

shielding and lined with 

cadmium. 

Source of Signal Gamma rays 

Tritons—Neutron detection is 

achieved through the 

juxtaposition of an enriched 

6

LiF layer. Triton particle 

detection is used to infer 

neutron flux. Alpha particles 

may be stopped by an Al 

absorber to minimize SiC 

radiation damage. 

Gamma rays Gamma rays 

Neutrons 

Abbreviations: BNFL = British Nuclear Fuels plc; EPRI = Electric Power Research Institute; HRGS = high-resolution 

gamma-ray spectrometry; Thorp = Thermal Oxide Reprocessing Plant. (Source: Modified from Appendix A in Ref. 19.) 

3.5 CERENKOV RADIATION MONITORING DEVICES 

Cerenkov radiation is emitted whenever a charged particle passes through a medium at a velocity 

exceeding the phase velocity of light in that medium. In water, the phase velocity of light is about 75% of 

its value in a vacuum. An electron passing through water and having a kinetic energy greater than 

approximately 0.26 MeV will produce Cerenkov radiation. In spent fuel, these electrons include 

Compton electrons produced by gamma radiation, beta rays that escape directly into the water, and the 

interactions of high-energy neutron capture gamma rays that produce electrons from Compton scattering 
18

and pair production. The intensity of Cerenkov light generated by irradiated fuel is proportional to the 

radiation field intensity in the vicinity of the irradiated fuel. This field intensity is proportional to the 

burnup of the fuel and inversely proportional to the fuel cooling time. 

Cerenkov light measurement is a very simple and relatively nonintrusive method for verifying the 

effectiveness of spent fuel safeguards compared to other methods because it involves only the viewing of 

spent fuel assemblies using a Cerenkov viewing device (CVD), and no movement of the stored spent fuel 

is required. Current CVDs can detect Cerenkov glow images without interference from normal lighting. 

Because Cerenkov emissions range from the ultraviolet (UV) to the infrared, the CVD uses a UV light 

image intensifier, which has good sensitivity to UV light due to the use of a Cs-Te photocathode material. 

The CVD also uses a conventional night viewing device with an optical filter to minimize interference 

from lights in the facility. Use of modern CVD equipment allows the detection of a weak Cerenkov glow 

image that may result from fuel that has a low burnup or a relatively long cooling time.20

An improved CVD, the UV-I.I CVD, was tested in Japan at BWR and PWR facilities in the early 1990s 

with fuel burnup values ranging from 6.2 to 33 GWd/MTU and cooling times from 370 days (1 year) to 

6,300 days (17.2 years). The CVD was placed in a waterproof housing with a window and was 

submerged in the spent fuel pools to about 10 cm. The distance between the CVD and the spent fuel was 

about 7 m, and all facility lights were in their normal (on) position.20 The Cerenkov radiation occurs 

strongly in the space between fuel pins and is highly collimated when viewed along the axis of the 

assembly. This is the normal view for LWR fuel because the spent fuel is stored vertically and is viewed 

from above the spent fuel pool. The highly collimated nature of the light requires that the instrument used 

to measure the Cerenkov glow have its optical axis positioned along the axis of the fuel assembly to view 

the maximum light intensity emitted by the assembly.21 The results of the tests in Japan showed that the 

improved CVD was capable of providing clearer and more vivid images of the spent fuel than older 

CVDs and was able to identify Cerenkov patterns in the spent fuel for low-burnup and long-cooling-time 

assemblies. 

In 1993, the Swedish and Canadian safeguards support programs began a joint program to develop a highsensitivity Cerenkov viewing device. This instrument was called a digital CVD (DCVD). In January 

2002, a prototype DCVD was tested at Central Interim Storage for Spent Fuel in Sweden on PWR fuel 

and nonfuel, long-cooled BWR fuel, and Ågesta test reactor fuel assemblies.22 Six PWR fuel assemblies 

were measured to determine their Cerenkov intensity as a function of cooling time. The burnup values 

and cooling times of the six fuel assemblies are shown in Table 3.3. Intensities were measured by 

selecting the brightest pixels (less the brightest 1% of intensities to reduce the effects of noise). Since 

theoretical calculations of photon intensities as a function of burnup and cooling times for PWR fuel are 

not available, corrections to normalize the higher and lower burnup values (T12 and K30, respectively) 

could not be done. The authors concluded that the prototype DCVD was successful in meeting their 

measurement objective of verification of fuel with a burnup of at least 10 GWd/MTU and a cooling time 

of 40 years or less.22

On the basis of the six PWR examinations, the authors of the Swedish report indicated that it may be 

possible to determine the cooling time of an assembly knowing its burnup (from reactor records) and 

DCVD Cerenkov intensity. Considerable work remains to be done in this area, specifically with respect 

to the precision of readings and the stability of the CVD detector. Water quality in spent fuel pools is also 

an important factor, although it should be constant in any one facility over a short time period.22 It was 

noted that the spent fuel pool at Ringhals Unit 2 contained about 2,000 μg/g boron to absorb neutrons 

emitted from the spent fuel. The boron caused the light intensity of the Cerenkov glow to be reduced to 

about half the intensity received from similar spent fuel assemblies in a pool with no boron.22 
19

Table 3.3. Intensity measurement of six fuel assemblies 

Fuel 

identification 

number

Burnup 

(GWd/MTU)

Cooling time 

(years)

Intensity 

(counts)

Z04 44 1 17101

Y26 43 2 8113

W05 43 4 4053

T12 56 6 2891

R26 41 9 1047

K30 32 14 462

Source: adapted from Reference 22. 

Another consideration that must be taken into account is the “near-neighbor” effect of spent fuel 

assemblies in close proximity to the assembly being examined (either adjacent to or diagonal with the 

assembly being examined). The near-neighbor effect occurs when gamma rays from neighboring fuel 

assemblies travel into the adjacent assembly and generate Cerenkov light in the water spaces. In BWR 

Cerenkov examinations, a clear near-neighbor effect has been noted; in the limited PWR examinations 

conducted in Ref. 22, the data were mixed; thus, a clear correlation showing how much the near-neighbor 

effect has on PWR spent fuel assemblies is still unknown.22

As of 2007, the DCVD is in its third generation, has integrated hardware, is equipped with a zoom lens 

(80 mm–200 mm), and uses windows-based software. During a recent training course at IAEA, the new 

DCVD was tested, and the DCVD partial defect test capability was successfully “demonstrated” to 

perform improved safeguards functions (detect single rod defects).23 The authors noted that the DCVD 

could not clearly identify unirradiated substituted rods in spent fuel assemblies, indicating that the 

Cerenkov glow examinations would still not provide satisfactory information for lightly burned or very 

old spent fuel. 

In summary, equipment and techniques exist that can be used successfully to detect the Cerenkov light 

given off by spent fuel; however, CVDs were developed to support nuclear safeguards programs and have 

not been optimized to measure spent fuel burnup except in a broad sense. For low-burnup fuel or fuel 

cooled for significant time periods, the Cerenkov glow will be very dim—although it should still be 

detectable. Equipment alignment is critical to this type of examination and is especially critical when 

examining fuel that has been cooled for long periods of time. Near-neighbor assemblies and pool water 

quality (boron concentration) can also affect the examination results. It may be very difficult to 

distinguish between old, moderately burned fuel and new, very lightly burned fuel by the Cerenkov glow 

measurement. Nevertheless, a Cerenkov glow examination may be capable of providing “go/no-go” 

information for imprecise screening. 

3.6 COMMENTS ON INTERNATIONAL PRACTICES AND 

REGULATIONS 

The issue of how to demonstrate compliance with safety criteria, for example, minimum required burnup, 

when applying credit for fuel burnup is of common interest internationally. As evidence, during the most 

recent IAEA technical meeting24 on burnup credit, a working group was convened to review the status of 

methods used to demonstrate compliance with safety criteria among nations currently applying burnup 

credit. The main observations of the working group report24 are summarized in this section, which also 

includes recent information on German regulations and practice; the reader is referred to the working 

group report for the full details. 
20

The working group discussions focused on methods of verifying assembly burnup and reviewed examples 

of how this is put into practice as part of spent fuel operations in France, Germany, the United Kingdom, 

and the United States. The examples highlighted both similarities and differences in practice. Where 

differences were identified, the group reviewed the causes both in terms of variations in operational 

design/requirements and in terms of any differences in the underlying safety philosophy. In general it 

was found that the safety philosophy and associated methods are very similar in all the countries reviewed 

and that differences with respect to compliance issues arise mainly from differences in the operational 

environments. Examples of spent fuel operations reviewed by the working group are provided below. 

An outcome of the working group discussions was a recommendation that the IAEA Standard on 

Transport25 be reviewed with respect to the current “requirement” for a measurement of burnup. The 

recommendation stems from the recognition that the rigor required to demonstrate compliance of any 

safety criterion should be dependent upon the importance of that criterion to the overall safety, which can 

depend on many aspects of the spent fuel system and its operations. A related observation is that there is 

significant variation between standards with respect to whether a measurement of burnup is a firm 

requirement or not.*

3.6.1 Summary of Current Practices and Regulatory Requirements 

The following examples of burnup credit practice were presented and discussed by the group. 

• Dissolution in the Thorp reprocessing plant (burnup credit is not used for on-site storage) 

• Transport, storage, and dissolution at the Cap de La Hague reprocessing plant 

• Reactor pool storage in the United States and Germany 

3.6.1.1 United Kingdom—Thorp Reprocessing Plant 

Criticality evaluations for transport of SNF to the Thorp plant are based on the fresh-fuel assumption. 

Burnup credit is applied to the head-end plant, particularly in the dissolvers, where large batches of fuel 

comprising several fuel assemblies are sheared into small lengths and dropped into hot nitric acid doped 

with gadolinium. The original criticality assessment was based on the fresh-fuel assumption, which 

resulted in a requirement for significant concentrations of gadolinium to be added for criticality control. 

The current criticality assessment is based on actinide-only burnup credit, which reduces the gadolinium 

addition requirements (and hence reduces waste volumes). 

Consistent with typical practice, the safety criteria are based on a 5% administrative margin for normal 

conditions with additional allowance for code bias and uncertainty (including uncertainty associated with 

the burnup credit approach). For some low-probability-accident conditions, a reduced administrative 

margin (2% in keff) is applied. The burnup credit evaluation includes numerous conservative assumptions; 

for example, the fuel packing fraction in the dissolver is optimized to maximize the calculated keff value, 

which is reported to represent an additional margin of about 14% in keff relative to typical packing 

fractions. 

Compliance with the loading curve is made through a combination of checks against the supplier’s data 

and through measurements made on each assembly by the Thorp fuel pond feed monitors. The 

measurement is based on gamma-ray spectroscopy and neutron counting, which provide information on 

 *

It was noted that while the IAEA standard requires that “a measurement shall be performed,” (paragraph 674 of 

[25]), the advisory material explains that the “measurement technique should depend on the likelihood of misloading 

the fuel and the amount of available subcritical margin due to irradiation.” 
21

cooling time, burnup, initial enrichment, and residual enrichment (RE).*26 A go/no-go trip is set against 

the measured RE value, which prevents any assembly above the limiting RE from being fed forward to 

the shearing machine. The assessment demonstrates that the RE at the zero burnup end of the loading 

curve is bounding (i.e., a minimum) for the rest of the curve, so this value, with allowance for 

measurement and calibration uncertainties, is used as a test of compliance for all assemblies. The 

calibration is made through measurements on selected fuel assemblies before each campaign, and the 

measurements are calibrated against supplier data. Checks are made during and after each campaign to 

confirm calibration constants and uncertainties. The measured RE is based on neutron counts taken at a 

single axial location near the center of the element. 

Based on the available information, it is surmised that this system is more of an approximate check of 

burnup, through the determination of the RE, than it is an accurate verification or quantification of 

burnup. 

3.6.1.2 France—Transport, Storage, and Dissolution at the Cap de La Hague 

Reprocessing Plant 

At the Cap de La Hague reprocessing plant, burnup credit is applied for on-site pool storage of PWR SNF 

and dissolution in the rotary dissolvers, where sheared rods are dropped into one of the dissolver buckets 

that are soaking in hot, unpoisoned nitric acid. For these two applications, in which the required level of 

credited burnup is different, the burnup credit approach is also noticeably different. For transport of SNF 

to La Hague, the criticality assessments are also based on burnup credit, very similar to those applied for 

pool storage. For all applications, the safety criteria are based on an administrative margin of 5% in keff

for normal conditions and some accident conditions, with additional allowance for code bias and 

uncertainty. For some unlikely accident conditions, the administrative margin is reduced to 2 or 3% in 

keff. Some uncertainties associated with burnup credit are taken into account through additional 

conservatisms in the depletion calculations. 

The original assessments for storage and transport were based on the fresh-fuel assumption, which was 

sufficient to comply with the criticality safety criteria for low enriched fuels. However, as fuel 

enrichments have increased, the assessments have become based on actinide-only burnup credit and the 

following two conditions relative to burnup verification. 

y If the minimum required burnup is less than or equal to 3.2 GWd/MTU, a simple gross 

gamma-ray measurement (or a validated—by regulators—equivalent method to confirm the 

irradiation) is sufficient. 

y If the minimum required burnup is higher than 3.2 GWd/MTU, a validated (by regulators) 

burnup measurement must be performed. The intention of this measurement is to verify that 

the irradiation in the 50 least-irradiated centimeters (axially) of the fuel assembly is higher 

than the minimum required burnup.

These burnup verifications are performed before loading the fuel assemblies into a transport cask and are 

completed by the supplier. The type of measurement, made in the supplier’s plant, depends on the 

supplier (French, German, etc.) but must be approved by the French regulators. 

 *

RE is derived from measurement of sub-critical neutron multiplication, calibrated against data received from 

the suppliers of the fuel. These data are used to form the equivalent 235U enrichment from the residual 235U, 239Pu, 

and 241Pu in the fuel. For 239Pu and 241Pu factors are applied to create their equivalent (in reactivity) 235U mass. 

These factors are based on the relative thermal fission cross-sections. The values of RE derived from the supplier 

data is then used to convert the measurements of neutron multiplication into RE. 
22

For dissolution in rotary dissolvers, the safety assessment is based on actinide-only burnup credit, which 

leads to loading curves: maximum permissible mass per dissolver bucket as a function of burnup (one 

loading curve for each initial enrichment). Each of these curves presents a burnup limit over which 

criticality safety is ensured by the bucket geometry (without restrictions on loaded mass). A PWR fuel 

assembly is typically loaded into three or four buckets (about 110-liter volume each). 

Compliance with the loading curves and determination of the number of buckets needed to load a fuel 

assembly are made through a combination of checks against the supplier’s data and through 

measurements made on each assembly. The measurements must provide information on the initial 

enrichment, the average burnup, and the assembly axial profile. These calibrated and validated 

measurements are implemented between the storage pools and the dissolvers and consist of gamma-ray 

scanning (on two opposite faces of the fuel assembly) and passive neutron measurements (on the two 

other opposite faces). The axial scannings are interpreted by an online evaluation program. A go/no-go 

trip is set against the comparison between the average burnup measured and the supplier’s data. Details 

were not provided on the measurement calibration and associated uncertainties. 

3.6.1.3 Germany—Transport and Storage 

The requirements for pool storage facilities for LWR fuel assemblies are contained in the German safety 

standards KTA 3602 and DIN 25471.27 Transport and storage cask requirements are defined in the 

German safety standard DIN 25712.28 A review of these German regulatory standards is available in 

Ref. 29. Although they are licensed under different regulations, the requirements in these regulations with 

respect to burnup credit applications to pool storage and transport/storage casks are completely consistent 

with respect to implementation and validation of the calculations, determination of criticality safety 

acceptance criteria and loading criteria, and prevention of misloading events. An administrative margin 

of 5% is applied for normal conditions, with allowance for a lower value under certain conditions. 

Both DIN 25471 (Ref. 27) and DIN 25712 (Ref. 28) require that the misloading event has to be excluded 

as a design basis event by applying the double contingency principle directly to the misloading event: at 

least two independent, unlikely, and concurrent incidents have to happen before a misloading event can 

occur. This application of the double contingency principle and hence the exclusion of the misloading 

event as a design basis event from the criticality safety analysis is achieved by applying independent 

layers of hardware and software measures, ensuring the reliability of the reactor record data and the fuelhandling procedures applied to the pool and cask loading operations.30 Therefore, no burnup 

measurement is required. 

For transport/storage casks, DIN 25712 (Ref. 28) allows application of actinide-plus-fission-product 

burnup credit to LWR uranium oxide and mixed oxide fuel. Flooding of the transport casks, as well as of 

the storage casks, must be considered. In contrast to pool storage, a burnup verification measurement for 

fuel to be loaded in a cask is stipulated, per paragraph 674 of IAEA TS-R-1. 

In the past, when the spent fuel was transported to France for reprocessing, a burnup measurement was 

performed according to the French requirements: measurement of the shape and burnup of the leastburnt 50 cm at the top end of the fuel zone for each assembly. For this purpose the Python system was 

used at the Brokdorf (PWR) and Gundremmingen (BWR) plants. Spent fuel shipments to France no 

longer occur. Under current German regulations, only a check (e.g., by performing a gross gamma-ray 

measurement) is required for SNF with record burnup values less than 10 GWd/MTU (i.e., less than one 

cycle burnup), which requires neither a calibration against reactor records nor an evaluation of 

uncertainties. 

The following text, extracted from Ref. 29, provides particularly relevant insights and information relative 

to the German regulations and practices. 
23

2.6.2. Quantification and verification of the burnup of the fuel assemblies to be loaded in a 

transport or dry storage casks (DIN 25712) 

As laid down in the safety standard DIN 25712, quantification of the burnup of the fuel assemblies 

shall be based on the evaluation of the reactor records. Verification of the burnup shall be based 

on the evaluation of the reactor records as well and, additionally, on a consistency check by means 

of a measurement (such as a gamma scanning, a measurement of passive neutron emission, or a 

combination of both measurement procedures). Determination of the fuel’s burnup and its 

verification shall be performed in compliance with the quality assurance requirements laid down 

in the German safety code KTA 1401 (cf. section 2.6.1). 

The only reason that the requirement for a consistency check by means of a measurement has been 

included in the safety standard DIN 25712 is that this requirement is laid down in the IAEA 

regulation IAEA TS-R-1 [6]. If this were not the case this requirement would not have been 

included in the standard DIN 25712 since the evaluation of a measurement requires information 

from the reactor records [7] and since studies like [8] have shown that the utility-supplied data on 

burnup are of greater accuracy and reliability than could be provided by additional radiation 

measurement of spent fuel. The following rules are therefore laid down in the safety standard 

DIN 25712: 

• The check whether the burnup of a fuel assembly fulfills the loading criterion (loading 

curve, cf. Fig. 6 for example) is carried out by using the reactor record information only: 

The loading criterion is met when the assembly’s burnup obtained from the reactor 

records does not fall below an upper discrimination limit which is calculated, at a 

significance level of 5%, from the minimum required burnup given by the loading 

criterion and the uncertainty of the burnup value obtained from the reactor records. An 

example for determining such an upper discrimination limit is given in the informatory 

part of the standard DIN 25712. 

• The consistency of a measurement result and the reactor record information is proven 

when the measurement result falls into an interval which is given by the lower and upper 

discrimination limit calculated, on the basis of a 5% significance level, from the reactor 

record information, the uncertainty of this information and the uncertainty of the 

measurement result. An example for determining such an interval is given in the 

informatory part of the standard DIN 25712. 

• If it is obvious that the actual burnup of a fuel assembly is much greater than the 

minimum required burnup given by the loading criterion, then, instead of the consistency 

check, a measurement procedure may be used which is capable of demonstrating 

compliance of the assembly’s burnup with the loading criterion. (Such a measurement 

procedure can be based for instance on the measurement of the intensity of the 662 keV 

gamma-transition following the decay of the burnup indicator Cs-137 [7][9].

3.6.2 Summary 

Although approaches to burnup credit criticality assessment are very similar, the outcomes with respect to 

compliance procedures, particularly with regard to requirements for burnup verification measurements, 

vary considerably. In general the level of reliance placed on burnup verification depends on 

• the amount of burn-up being credited; 

• the level of confidence in other sources of information (e.g., reactor records); 

• the presence of other contingencies (e.g., loss of control over boron concentration in a storage pool, 

flooding in a dry transport cask); 

• the system and/or operational sensitivity to burnup and/or misloading; and 
24

• the presence of other margins of safety not explicitly credited in the assessment (e.g., packing fraction 

in dissolvers, soluble boron in spent fuel pools).

A common comment during the working group discussions was that the requirement for a burnup 

measurement in their regulations and/or practices is necessary for compliance with the requirements of 

IAEA TS-R-1.
25

4 REACTOR BURNUP RECORDS: 

GENERATION AND ACCURACY 

4.1 BACKGROUND 

Reactor operation records are generated, controlled, and archived per plant procedures that support 

10CFR50 Appendix B and are required to be readily retrievable for design, audit, or investigation 

activities. These records, also called core-follow data, contain operating parameters that are important to 

the safe and reliable operation of the nuclear plant and are either measured, calculated, or inferred from a 

combination of measurements and predictions. 

Measured records include reactor operation monitoring data such as reactor power level, reactor coolant 

flow, and reactor coolant boron concentration. These data are obtained from plant instruments—

thermocouples, flowmeters, radiation detectors, etc.—and electronic equipment that are calibrated and 

maintained per plant procedure to ensure a high degree of reliability and accuracy. Measurement 

uncertainties associated with these data have been quantified from extensive fuel vendor and utilities 

studies, which were approved by NRC as part of the reactor design and safety analyses methodology 

review. 

Most core physics parameters, including those used for Technical Specifications compliance, cannot be 

directly measured and are either calculated by a nodal simulator neutronics code or inferred from testing 

involving reactor manipulation before cycle startup. These include control rod reactivity worth, soluble 

boron worth, moderator temperature coefficient, and shutdown margin (SDM). Three dimensional core 

power distribution can be inferred directly from neutron flux measurements, and depending on the reactor 

type, this measurement is online (continuous) or offline (flux mapping performed monthly or as needed). 

Typically BWRs, and some PWRs, have fixed in-core detectors, which allow for online core power 

distribution and thermal limits monitoring. Most PWR plants, however, do not have fixed detection 

systems and perform monthly flux maps using movable detectors that are inserted in the core. In both 

fixed and movable detection system plants, the flux is measured in only one-third or so of the fuel 

assemblies in the core. A computer code, using data generated by the reactor reload design codes, is used 

to calculate the power distribution for the entire core using the instrumented location measurements to 

infer the predictions in the noninstrumented locations. 

The inferred core power distribution obtained from the combination of in-core measurements and 

corrected predictions is referred to by the plant engineers as the “measured” core power distribution. It 

should be noted that this characterization would be closer to correct if all the core locations were 

instrumented. Actually, only about 30% of a PWR core is instrumented. The decision to not instrument 

the entire core was made during the design phase of the previous generation of reactors and is based on 

the argument that instead of measuring all the fuel assemblies in the core, it is acceptable—using basic 

statistics—to directly measure the neutron flux in only a randomly selected representative sample of the 

core. Cycle-specific, burnup-dependent computer calculations are used to generate code bias and 

assembly-to-assembly coupling factors that are used to infer assembly average power and power peaking 

factors for uninstrumented assemblies based on the flux measurements in the nearby instrumented 

assemblies. This is accepted practice as the sample size is significant compared to the total fuel assembly 

population in the core. Also, due to the permanence and distribution of the core instrumented locations, a 

majority of the fuel assemblies will be measured for at least one cycle during their residence in the core as 

their locations are changed from one cycle to the next. A typical fuel assembly resides in the core for two 

to four reactor cycles. 
26

Assuming the reactor core power distribution is adequately determined, it is then important to determine 

an accurate burnup value for each fuel assembly. The fuel assembly burnup is simply the amount of 

cumulative energy generated per weight of fuel. The burnup cumulated in fuel assembly i, ΔBi, during a 

reactor operating time interval Δt can be expressed as 

ΔBi = [(RTP * Δt * CF) / MF] * Pi , (5)

where ΔBi is in megawatt–days per metric ton of uranium and Δt is in days. RTP is the reactor rated 

thermal power, a constant for a given core, expressed in megawatts, CF is the Capacity Factor, or the 

fraction of measured reactor thermal power to rated thermal power (CF=1 for a reactor operating at full 

power), MF is the mass of fuel in metric tons of uranium, and Pi is the relative power in Fuel Assembly i. 

In Eq. (5), MF is precisely measured in the fuel fabrication facility and is formally transmitted to the 

nuclear utility. CF is determined accurately as the operating reactor thermal power is continuously 

measured at the plant by redundant methods. Because the measured thermal power is important to the 

determination of burnup, a discussion on its measurement accuracy is presented in Section 4.2. 

All parameters in Eq. (5) are directly measured except Pi. Therefore, an accurate determination of the 

fuel assembly burnup requires accurate determination of the core power distribution. 

The fuel assembly burnup is a core-follow parameter that is particularly visible to plant personnel and is 

currently used in regulation, procedure compliance, and design activities. The nuclear utilities are 

required by existing NRC regulation to keep the peak rod average burnup under 62,000 MWd/MTU for 

fuel integrity protection. Depending on the fuel assembly design, this rod-averaged limit corresponds to 

an average bundle peak burnup limit of about 45,000 MWd/MTU. The burnup records are also used for 

storage of discharged fuel assemblies in the spent fuel pool and in dry casks when they complete their 

core residency. The inferred burnup information is also an input to the special nuclear material (SNM) 

program. In current industry practice, the burnup values are used as input to the core design and safety 

analyses of the upcoming cycle as the starting exposure values for the reinserted assemblies, which 

typically compose two-thirds of the new core. The burnup values of these reinserted assemblies are 

confirmed from measurements performed during the power ascension tests following the refueling 

outage. 

4.2 THERMAL POWER MEASUREMENT 

As stated in Eq. (5), to calculate assembly burnup, the thermal power of the reactor must be known. The 

thermal output of a nuclear reactor is determined from temperature and flow measurements of the cooling 

water circulating through the reactor core. Thermal output is based on a complete steady-state energyrate measurement across a defined envelope that includes the nuclear steam supply system. The primary 

components of thermal output are the energy rate in steady-state steam flow as measured by steady-state 

feedwater flow and the enthalpy difference between the steam (temperature and quality measurement) and 

the feedwater (temperature). Measured energy losses and credits are included for blowdown; letdown and 

makeup; seal water; component cooling water; electrical power inputs to pumps, heaters, and 

miscellaneous equipment; and radiative and convective losses. Standards from the American National 

Standards Institute (ANSI) and utility Technical Specifications detail the required accuracy of the flow 

measuring, temperature, and pressure devices in the nuclear steam supply system. The net result of these 

requirements is an overall core thermal output uncertainty of somewhat less than 1% at the time of 

measurement.31 The effects of transients, below-full-power operations, and the possibility of instrument 

calibration drift between periodic calibrations all tend to increase the uncertainty associated with powerlevel measurements. Thus, the net accuracy of core power measurements is about 1% for the operational 
27

power-level measurements that are a primary input to fuel burnup determination.4,18 The sum of the 

thermal power of all assemblies must equal the total thermal power of the reactor over any time span. 

Therefore, if the burnup of one assembly is high, another assembly burnup must be low, resulting in a 

“zero sum” characteristic of any errors. The most likely source of a possible error, one that applies to all 

assemblies, is the integral of the power of the reactor, which, as noted above, is measured very 

accurately.4

 The measurement of total core thermal power for calculation of burnup increments is done at 

least daily. The statistical error in total measured power from, for example, 100 days, gives a factor of 10 

reduction in the single 1% uncertainty, with the result that the burnup uncertainty contribution from the 

measurement uncertainty associated with total core power is less than 0.1%.32 

The time integral of the thermal power [in gigawatt-days] produced by the reactor is then used as the basis 

for the burnup assignment to individual assemblies using the SIMULATE-3,33 or similar, computer code. 

The burnup assignment based on a core irradiation analysis that matches the actual energy output is often 

referred to as the “core-follow” assembly burnup.34

4.3 IN-CORE POWER DISTRIBUTION MEASUREMENT 

Because of the differences in reactor types and designs, the in-core power measurement systems 

encountered in the nuclear industry are varied. However they share several common features, and their 

purpose and general layout are common to all units. 

Fission chamber detectors are used to directly measure neutron reaction rates at several axial locations 

along the length of the fuel assemblies in the reactor. A typical fission chamber detector consists of two 

concentric cylinders, which act as electrodes. The collecting electrode is the inner cylinder and is 

separated from the outer electrode, which is coated with highly enriched U3O8 (90% 235U), by a small gap. 

The gas in the gap is ionized by the charged particles resulting from neutron fission in the uranium. The 

negative ions are accelerated to the collector by a potential difference maintained between the two 

electrodes, and the resulting current is measured and is proportional to the neutron flux at the detector 

location. The number of detectors that are used varies and depends on the size of the reactor and whether 

the detectors are fixed inside the core (BWRs and some PWRs) or are movable (most PWRs). 

Roughly one-third of the fuel assemblies in a PWR core are instrumented. For example, a three-loop 

Westinghouse PWR (157 assemblies in the core) has 50 instrumented core locations, and a four-loop 

PWR (193 assemblies in the core) has 58 such locations. Figure 4.1 illustrates the randomly distributed 

detector locations in a three-loop Westinghouse PWR. 
28

Fig. 4.1. In-core instrumented locations in a three-loop Westinghouse pressurized water 

reactor (Courtesy of Progress Energy). 

A typical PWR movable in-core instrumentation configuration is illustrated in Fig. 4.2. Other reactor 

types have fairly similar systems. In a movable detection system, five or six fission chamber detectors are 

inserted through the instrumented core locations using retractable thimbles. The thimbles are closed at 

the leading reactor ends; thus they are dry inside and serve as the pressure barrier between the reactor 

coolant pressure and the atmosphere. Axial flux traces are recorded as the detectors are driven through 

the core. These traces typically contain 61 axial data points and provide detailed axial neutron flux 

variation along the fuel assembly. During the flux map measurement, which typically lasts several hours, 

it is imperative that the reactor conditions (power, temperature, control rod position, etc.) are held as 

constant as possible by plant operators: the goal is to obtain a core power distribution that is instantaneous 

(i.e., a snapshot of the reactor state, although several hours would have elapsed between the first and last 

trace measurements). The pertinent reactor conditions are recorded along with the detector signals for 

post-processing. For data normalization purposes, cross calibration of the detectors is performed by 

inserting all of the fission chambers through a common core location. 
29

Fig. 4.2. Typical Westinghouse pressurized water reactor in-core flux map system. 

4.4 IN-CORE MEASUREMENTS PROCESSING 

4.4.1 Flux Map Processors 

Even if all the fuel assemblies in the core were instrumented, the measured flux traces would still have to 

be processed. This processing step includes data validation checks, corrections for background and 

reactor operation fluctuations, evaluations of duplicate and symmetrical traces, and normalization to a 

single detector. As discussed previously, typically two-thirds of the fuel assemblies in the core are not 

instrumented, and power distribution at these locations needs to be determined. Several codes are 

available from the nuclear vendors to process the raw flux map data and generate the power distributions 

at the noninstrumented locations.35,36,37 Utilities, through EPRI, have developed a core analysis system 

(ARMP) to support the operational, fuel management, and SNM accountability needs of their operational 

reactors. The ARMP system, which is typical of all such operational systems, has the ability to interpret 
30

and use in-core measurements to provide flux and power distribution data and to reflect changes in fuel 

and nuclear absorber characteristics as fissionable and absorbing materials are consumed. 

Modern commercial flux map processors use proprietary methodologies that are relatively simple in 

concept and share the same general calculation flow. They are based on the use of predictions from the 

nodal simulator code and coupling factors relating noninstrumented locations to the nearest instrumented 

assemblies. The flux map processors significantly accelerate a process that in the 1960s and 1970s relied 

partially on hand calculation. Consequently, these processors eliminate the human factor element; ensure 

faster, more straightforward independent verification; and provide more easily retrievable core-follow 

records. Typically, the processors implement the following steps. 

• The “raw” measured traces are checked for validity, corrected for background interference, 

statistically evaluated for any available duplicate and symmetric traces, and normalized to a single 

detector response. 

• The assembly flux and/or power predictions are expanded from the nodal simulator number of axial 

nodes (~ 24) to that of the measured axial locations (~61). 

• Measured reaction rates at the instrumented locations are compared to their predictions from the 

nodal simulator code. 

• Correction factors are generated from the ratios of the measured to predicted reaction rates at the 

instrumented locations and are used to adjust the predicted assembly power to infer the power in the 

uninstrumented assemblies. 

• Coupling or weighing coefficients for each noninstrumented location are derived using data from the 

nearest instrumented assemblies. 

• The power distributions at the noninstrumented locations are derived using the nodal simulator 

assembly powers along with the coupling coefficients. 

• Safety factors and margins to the Technical Specification thermal limits are computed using the 

inferred core power distribution. 

Equation (5) is then used to determine the cumulative burnup from the inferred core power distribution 

and the cumulative reactor energy generation. 

Figure 4.3 is an actual plant graph that illustrates the flux map processor comparative function. The 

comparison between the predicted and measured axial power distributions in an instrumented assembly is 

typical. In this figure the discontinuities in the measured data correspond to the presence of assembly 

spacers (which depress the flux) and correspond to the higher observed differences. For this particular 

case, the top and bottom of the assembly are blanket regions consisting of natural uranium used to 

increase efficiency, a feature of low-leakage core design. These blanket regions typically have higher 

differences because they are low power areas with high measurement statistical fluctuations. Outside of 

the spacers and blanket areas, the differences between predictions and measurements are in good 

agreement (generally below 5%). The purpose of the flux map is not only to validate the core-follow 

code but to determine the correction factors used to tune the predictions to yield accurate power 

distributions, and thus accurate thermal limits margins. 
31

Fig. 4.3. Comparison of measured vs predicted axial power distribution at an instrumented core 

location (actual flux map data). The yellow points are the percent difference between the predicted 

and measured values, with negative differences not shown (Courtesy of Progress Energy). 

4.4.2 Reactor Analysis Codes 

The current methodology used to generate the core-follow data is approved by NRC for this purpose and 

is essentially the same that is used for reactor core design, augmented with in-core measurement 

processor codes. The most important among the several computer codes that constitute the design 

methodology are a lattice physics code (such as CASMO38) for fuel assembly modeling and a nodal 

reactor simulator code (such as SIMULATE33). The lattice physics code is a multigroup two-dimensional 

transport theory code used to conduct burnup calculations on an assembly or on a single fuel pin and 

handles a geometry consisting of cylindrical fuel rods with varying composition in a square-pitch array. 

Once the cross sections and other neutron transport constants are determined, they are used as input to the 

nodal simulator code. The nodal simulator code is a two-group three-dimensional program that solves the 

neutron diffusion equation for the homogenized nodal neutron flux. Typically, the reactor physics 

parameters are determined at 24 axial and 4 radial nodes in each assembly; this ensures detailed modeling 

of the core isotopic inventory and power distribution. During the core design phase, the nodal simulator 

depletes the core for the duration of the cycle at increments of about 1,000 MWd/MTU core average 

burnup and generates reactor parameters on a core, assembly, and node basis at each burnup. Using the 

cross sections and other physics constants for the major nuclides that are supplied by the lattice code, the 
32

SIMULATE code updates the isotopic inventory at each reactor burnup. This inventory is determined at 

each core node and includes the atom densities that are necessary to determine the reaction rates, and 

subsequently the nodal flux and power. The output from the reactor physics code includes safety peaking 

factors (such as FQ and FΔH for PWRs and minimum critical power ratio for BWRs), nodal power 

distribution, fuel assembly burnup, xenon and samarium reactivity worth, critical boron concentration, 

and core keff. 

The typical nodal code also has a detector model that calculates the reaction rates as measured by a 

detector inside the fuel assembly during the flux mapping. This is an important validation feature that 

confirms the acceptability of the assembly axial distribution predictions. 

 

During the core design process, reactor state points, such as 100% rated power and known rod positions 

(all control rods out of the core for PWRs, selected design rod patterns for BWRs) are modeled. The 

core-follow calculations involve depleting the nodal simulator to the current day and time in the cycle 

with the actual operating core state parameters. This ensures that the reactivity and core physics 

parameters are adjusted on a timely basis—daily, or hourly, depending on the utility—for the deviations 

between the assumed operating conditions in the design phase and the actual conditions. 

It should be noted that use of microscopic depletion techniques to track individual actinide and fission 

product inventories in three-dimensional core simulator codes is a relatively recent development. Much 

of the current and historical fuel assembly burnup predictions and core follow data are not based on such 

a method. 

4.4.3 Regular Validation of Reactor Analysis Codes 

The primary driving force for accuracy in fuel assembly burnup measurement is derived from the 

regulatory-based requirement for accurate reactivity balances, as measured by the difference between 

actual and predicted control rod position (both BWRs and PWRs) or boron concentrations (PWRs). 

Accuracy in reactivity balances requires accuracy in power distributions, from which the burnup 

increments are directly determined. Thus, the level of accuracy of individual assembly burnup values is 

driven by the regulatory-based accuracy needs in reactivity balances.31 Utility Technical Specifications 

require that the reactor reactivity be predicted within 1,000 pcm Δk/k. This is especially important for 

ensuring that the SDM, a calculated parameter that varies with cycle exposure, is accurately and 

conservatively determined. SDM is the ability to instantly bring down the reactor from full-power 

operation to shutdown by rapid insertion of the control rods. Generally, the utilities have a more 

restrictive acceptance criterion associated with reactivity prediction and typically use 500 pcm Δk/k. This 

criterion is verified regularly with reactor measurements: estimated critical position or estimated critical 

boron concentration, control rod position and daily updated boron letdown curve (required boron 

concentration at 100% rated thermal power). 

Several of the parameters generated by the nodal simulator codes during the core design phase are used as 

inputs to the safety analyses codes and need to be confirmed by plant measurements before new cycle 

startup. The startup testing scope differs among the different reactor types because of the inherent design 

differences. For PWRs, the scope of this testing and the associated acceptance criteria are defined in an 

ANSI/American Nuclear Society (ANS) standard.39 The PWR parameters that must be compared to incore measurements during the startup testing, also called the zero power physics testing (ZPPT), include 

estimated critical conditions, core reactivity, control rod worth, and moderator temperature coefficient. It 

is clear that a successful verification of the ZPPT acceptance criteria is possible only if the threedimensional nodal core power distribution is accurately predicted. For example, during one test, each 

control rod bank reactivity worth is measured at every few steps along the fuel assembly as the rod is 
33

withdrawn from the all-rod-in to the all-rod-out position. Each predicted reactivity worth along the 20 or 

so axial positions during the rod withdrawal must agree with the measured values consistent with the 

acceptance criteria in Ref. 39. Therefore, in addition to the total integral rod reactivity worth (expressed 

in percent mille), the predicted differential rod worth (percent mille/step) as a function of assembly height 

is validated. 

Following successful ZPPT acceptance criteria confirmation, the reactor is allowed to increase power to a 

holding point—typically 30% rated thermal power—and the results of the ZPPT are saved as reactor 

records. 

After completion of the ZPPT, a series of reactor measurements are taken at increasing reactor power 

levels up to full rated power as part of the power ascension testing. At about 30% rated thermal power, 

the first flux map of the cycle is performed. As noted previously, typically two-thirds of the reactor is 

composed of reinserted bundles whose burnup values were determined from the previous end-of-cycle 

flux map. One of the goals of this power ascension testing is to verify that each assembly is loaded in the 

correct position and has the burnup that was assigned to it from the last flux map. Should an assembly be 

misloaded or the burnup be significantly miscalculated, anomalous behavior should be noted in the 

reactor controls system, and per plant Technical Specifications, the situation would be evaluated and 

assemblies would be moved as necessary. Thus the startup physics testing provides a system to verify 

that the burnup of each assembly is consistent with the values used in the core design. 

Note that the startup physics testing would not detect small errors in burnup close to fluctuation levels but 

might identify burnup errors of approximately 5% or more that would impact assembly reactivity. An 

actual example of a PWR core that was loaded with a number of assemblies containing reactivity different 

from the expected reactivity was documented in NRC Licensee Event Report (LER) 26193020. In this 

instance, due to an initial manufacturing deficiency, six new fuel assemblies were made with three fuel 

rods containing integrated burnable neutron absorbers placed in the wrong assembly quadrant. For this 

17 × 17 assembly array, the three mislocated rods represent approximately 1% of the total number (264) 

of fuel rods. The uranium enrichment and composition of these rods were similar to those of the other 

rods except for the small amount (6%) of imbedded gadolinium compound. The core was loaded, and the 

reactor startup commenced. Upon completion of the first flux map at 30% reactor power, the utility 

engineers noted a radial core power tilt of 2.8%, which exceeded the acceptance criteria of 2%. The flux 

map indicated that the peaking factors were higher than expected, although still within Technical 

Specifications. The flux map program also produced a comparison of predicted vs measured fuel bundle 

relative powers, indicating a 14% higher power in the core area surrounding the misfabricated assemblies. 

The fabrication error was subsequently identified/confirmed via a review by the vendor of the fuel 

fabrication sheets. This example illustrates the sensitivity of the in-core detection system to small bundle 

reactivity differences between the design and actual core loading. 

Once the first full-power flux map is successfully completed, the reactor is allowed to operate at rated full 

power. Monthly flux maps are then taken to confirm the core design power distribution predictions and to 

adjust core power peaking factors (e.g., F∆H and FQ) to include measured deviations (e.g., core quadrant 

tilt & assembly radial and axial power variations) from design for comparison with safety limits to 

accurately determine thermal limits margins. Flux maps are also used to calibrate the ex-core detectors 

and determine additional reactor follow information such as core axial offset, a measure of the 

relationship between the amount of power generated in the top and bottom halves of the core, and 

quadrant power tilt ratio. 

In addition to in-core measurements, the utilities evaluate the accuracy of their core power distributions 

and assembly burnup predictions by how efficiently they planned the reactor energy requirements and 

refueling outage schedules. It is important to compute the core loading requirements and use the 
34

associated energy to its fullest as fuel cost is by far the single largest expense for the upcoming cycle. 

Because a large fraction of the new core is composed of reinserted fuel assemblies whose burnup values 

are input to the reload design and safety analysis of the upcoming cycle, it is important for these values to 

be accurate. Also, plant outages are periods of reduced generating capacity, and the utilities schedule the 

refueling periods years in advance to ensure that only one unit is shutdown at any given time. Falling 

outside the planned reactor operating window can have significant cost impact, both in terms of 

generating capacity shortfall and failure to maximize the use of the fuel energy. Underestimating fuel 

burnup could mean operating at less than full power at the end of cycle and possibly starting the refueling 

outage earlier than scheduled; overestimation could mean fuel assemblies will be prematurely discharged 

without the full utilization of their energy. In both cases—that is, falling outside of the operating window 

assumed in the cycle-specific safety analyses—the utility is required to again perform these analyses 

under an accelerated schedule and at a significant cost. It is therefore important for economic reasons that 

the fuel assembly burnup values be accurately determined by the nuclear utilities. Overall, the industry 

has an excellent record of accurately predicting the cycle end of full power operations and meeting the 

refueling schedules start-end windows. 

Successful reactor physics testing and good cycle length predictions give some unquantified confidence in 

the fuel assembly burnup values used in the reactor reload design process. However, it is not clear how 

far individual assembly burnup values would have to be in error in order to affect reactor performance, 

given the large number of fresh fuel assemblies present in the core at startup. 

4.5 IN-CORE BURNUP MEASUREMENT ACCURACY 

As discussed previously, the plant fuel assembly burnup record is technically not measured but inferred 

from measurements and code predictions. Approximately one-third of a core is directly measured, and 

the records do not generally differentiate between the data for instrumented and noninstrumented 

assemblies. Although a majority of the assemblies would be in an instrumented location for some portion 

of their reactor residency, very few would be in such a location in every cycle of their residency. 

However, there is a very large database of pairs of measurements and predictions at the instrumented 

locations because they are generated monthly during flux mapping. Comparisons of measurements to 

predictions are performed to quantify the accuracy with which the nodal simulator code calculates the 

power distribution and, thus, the burnup of a given instrumented assembly. A flux map summary report 

contains a detailed statistical evaluation of the measurements, predictions, and inferred data. Before 

methodology approval, vendors provide NRC with proprietary licensing topical reports 35,36,37

documenting the uncertainty associated with the flux map processor inferred power distribution based on 

actual plant data spanning several operating cycles. Consequently, most utilities have an acceptance 

criterion of about 5% between the observed measurements and predictions. 

Previous studies 4,32,34,40 have shown that utility-developed burnup data are accurate to within at least 5%. 

A 1989 study, The Incentives and Feasibility for Direct Measurement of Spent Nuclear Fuel 

Characteristics in the Federal Waste Management System,

4

 compared the accuracy of waste 

characteristics developed from utility-supplied data with the accuracy obtainable from direct 

measurements (in-core monitors). The report stated that “with respect to utility-supplied data it was 

concluded that because utilities measure total core power to within about 1% and make extensive in-core 

measurements of power distribution, assembly-average burnup values at end of life can be determined to 

within about 2% (to one standard deviation) of actual values.” In the document, Reactor Record 

Uncertainty Determination,40 developed by the U.S. Department of Energy Yucca Mountain Project in 

2004, the reactor records at nine PWR plants, consisting of 5,447 assemblies with assembly lifetime 

burnup values greater than 10 GWd/MTU, were evaluated, and their reactor records indicated an 

uncertainty in burnup of between 2 to 4.2% with a 95/95 confidence level. 
35

In the EPRI report, Determination of the Accuracy of Utility Spent Fuel Burnup Records,

32 an evaluation 

of the core-follow code SIMULATE at only the instrumented locations was performed. Actual flux maps 

from three cycles were accessed, and only the measured reaction rates at the instrumented locations were 

extracted. Normally in the processing of a flux map, the subsequent steps of correction factors and/or 

coupling coefficients generation followed by the use of the core-follow code predictions at the 

noninstrumented locations would generate inferred values at these locations that collect additional 

uncertainty at each mathematical correction. Therefore, the EPRI study removes these additional 

“interferences” and evaluates the code predictions in only the locations with in-core measurements. Postprocessing of the pertinent flux map data was performed using spreadsheets. After converting the 

measured reactions rates to measured assembly power, assembly burnup values were generated using the 

relationship between burnup and power [Eq. (5)]. The measured and predicted assembly powers were 

added and cumulated into measured and predicted burnup values at each time step, corresponding to the 

elapsed time between two successive flux maps. The uncertainty in assembly average burnup was 

expressed as the percent difference between measured and calculated burnup at each instrumented fuel 

assembly at the end of each of the three cycles. In the EPRI analysis, the area of concern was the overprediction of burnup (which would be nonconservative). Therefore, a one-sided tolerance provided the 

appropriate level of probability and confidence. The tolerance was applied to ensure that 95% of the 

measurements occurred on one side of the normal distribution. Because the distribution was assumed to 

be normal, the 95% one-sided tolerance could be defined to include 95% of the error distribution. The 

resulting uncertainty for the assembly average burnup in the instrumented locations is consistent among 

the three cycles: 2.49%, 1.67% and 1.99% for cycles 1, 2, and 3, respectively. 

The EPRI report also looked at movable detector reproducibility as that was a measure of uncertainty 

associated with repeated measurements of a signal detector in a given location. For each in-core flux map 

performed, a number of duplicate traces were obtained for reproducibility purposes at a given thimble 

location at different times during the flux map. In the three cycles of data analyzed, there were 441 pairs 

of duplicate traces. The mean error developed from this analysis was −0.147% with a standard deviation 

of 0.748%. 

Utilities have periodically conducted burnup evaluation studies for design purposes or in response to 

external or regulatory requests. In a Duke Energy Corporation response to NRC concerning a Technical 

Specification amendment for the McGuire 1 and 2 spent fuel pool, Duke discussed a statistical evaluation 

of its assembly burnup records. To quantify the bounding burnup measurement uncertainty, data were 

compiled for the discharged fuel assemblies from McGuire’s entire reactor operational history (≈ 1,900 

assemblies from both units). When the code predicted and in-core measured burnup data were compared, 

the maximum individual assembly error (burnup measurement error) observed was about 4.0%.34 Duke 

also noted that the measurement uncertainty showed a clear trend of smaller errors with higher discharge 

burnup. 

In 2004, the Tennessee Valley Authority (TVA) evaluated its reactor records for burnup accuracy at all 

three of its reactor sites. To validate the fuel database burnup values, TVA performed a comparison of its 

flux-map-generated burnup values for discharged fuel assemblies against the design code burnup values. 

This analysis was done to identify any transcription errors, data entry errors, or other input errors in the 

database. Unless significant burnup value differences were identified (≥ 2% difference), new burnup 

calculations were not conducted for these assemblies. The comparison revealed differences between the 

predicted and database burnup values that were typically less than a few tenths of a gigawatt-day per 

metric ton of uranium and were randomly distributed. A total of 1,117 fuel assemblies were reviewed. 

Most burnup errors were less than 1 GWd/MTU, with the largest deviation observed from design values 

being 1.51 GWd/MTU on an assembly with approximately 41.8 GWd/MTU burnup (3.6% difference). 

Only seven assemblies met TVA’s criteria for burnup value revision and required adjustments to their 

assigned reactor record burnup values. 
36

Utility reactor records generally contain core and assembly-specific information that includes assembly 

design, burnup, initial assembly enrichment, power history, boron concentration (parts per million), 

moderator temperature, and discharge date. Record keeping requirements at individual utilities, for both 

the content and storage media of reactor records, have varied over time, as have the requirements for 

information required by NRC. For very old SNF, the reactor records may include “batch averages” for 

burnup information. The use of batch average values arose because the utilities were allowed to group 

assemblies with similar characteristics (a batch) into a single record for the purposes of reporting fuel 

inventories to NRC. As was noted in the report Source of Burnup Values for Commercial Spent Nuclear 

Fuel Assemblies, Rev. 3,41 scoping examination of utility records for Three Mile Island Unit 1 (TMI-1) 

for plant core-follow burnup records produced mixed results.42 This examination involved 12 discharged 

reloads of fuel, representing approximately 740 assemblies. About two-thirds of the reported assembly 

burnup values agreed well with the measured burnup values. About one-third of the reported burnup 

values reflected batch average values rather than individual assembly burnup values, as seen in Fig. 4.4.43 

The differences were generally between +5 and −5 GWd/MTU, with maximum differences upward of 

−16 GWd/MTU for some assemblies.42 With regard to burnup calculations, the differences noted in the 

TMI-1 batch average data are not conservative and are cause for concern. Although not confirmed, it is 

expected that utilities have sufficient information to determine the individual burnup values for these 

assemblies currently reported with batch average values. 

-20000

-15000

-10000

-5000

0

5000

10000

1 38 75 112 149 186 223 260 297 334 371 408 445 482 519 556 593 630 667 704

Measured BU - RW859 BU (MWd/tU)

Assembly No. Reload 1 to 12

Fig. 4.4. Measured vs RW-85944 reported assembly burnup (BU) values. (Source: Ref. 18.)

Two other factors affecting the accuracy of fuel burnup calculations are the use of increased irradiation 

times in reactors (increased fuel burnup) and the installation in many reactors of more precise feedwater 

flowmeters. Increasing the burnup has the effect of “equalizing” burnup uncertainty. As was noted 

earlier, in the EPRI report, Determination of the Accuracy of Utility Spent Fuel Burnup Records,

32 an 

analysis was conducted on discharged fuel from three cycles of operation. This covered a startup cycle, a 

transition cycle, and a near-equilibrium cycle. At the end of each cycle, about one-third of the fuel 

assemblies in the core were discharged. Since only about 30% of the core locations were instrumented, 

this meant that about 34% of the assemblies were never located in an instrumented location. Of the 

remaining assemblies, 44% were instrumented during one cycle, 19% were instrumented during two of 

the cycles, and 3% were instrumented in all three cycles. The uncertainty in burnup evaluated over three 
37

cycles of operation demonstrated a general decrease in uncertainty with an increase in residence time or 

burnup. For assemblies discharged after one cycle of burnup, the mean burnup uncertainty was 1.9%; 

after two cycles of burnup, the mean burnup uncertainty was 0.98%; and after three cycles of burnup, the 

mean burnup uncertainty was 1.02%. The data were not weighted for different cycle lengths, and all data 

were given equal weight. This decrease in uncertainty is indicative of the self-correcting nature of 

burnup, which is seen as the decrease in assembly average burnup uncertainty as the in-core residence 

time of assemblies increases.32 The use of more precise feedwater flowmeters, which were approved for 

use in the 1980s, reduces the uncertainty in determining feedwater flow. The feedwater flow is a direct 

input to the calculated reactor thermal power, which is used in core-follow fuel burnup calculations. Thus 

since the advent of the new flowmeters, reactor thermal power can be calculated with less uncertainty, 

and the resultant assembly core-follow calculations for burnup can be more accurately calculated. 

Another way to evaluate the accuracy of total core burnup measurements would be to look at the change 

in boron concentrations in the core throughout a fuel cycle. As noted previously, the differences between 

true burnup values and measured burnup values in the entire core are expected to be small. In a nominal 

PWR core, at the start of a fuel cycle, boron concentrations are about 1,600 ppm and steadily decrease as 

fissile content in the fuel is depleted. By the end of a 15-month fuel cycle (16,000 MWd/MTU), the 

boron concentration in the core has been reduced to nearly zero—which is a decrease of 3.5 ppm per day. 

If the total core burnup were miscalculated by 5%, this would equate roughly to missing the end-of-cycle 

full-power capability by more than 22 days, which would be very obvious. By evaluating the boron 

concentration in the core, a utility can show that integral core power, and thus true core burnup, over any 

fuel cycle has been calculated accurately. 

One factor that impacts the uncertainties in assembly burnup calculations is the evolution of the computer 

codes used to calculate core power measurements. In the 1970s and 1980s, SNF assembly burnup values 

were inferred from monthly in-core flux measurements using cycle-specific data, typically supplied by the 

reactor vendors. Due to computing resource limitations in those days, these constants were prepared 

using two-dimensional fine-mesh (pin-by-pin) models. The methods used were adequate for the original 

purposes of the in-core flux measurements, which were primarily to identify misloaded assemblies, 

dropped control rod clusters, and cross-core power tilts. Since that time, however, there have been 

significant improvements in computing capabilities and in the software used to model reactor cores to 

provide continuous real-time monitoring capability. Burnup credit has been used for storage of 

commercial SNF at nuclear power plants since the 1980s as utilities worked to increase their on-site 

storage capacity. Reliance on the plant-supplied assembly burnup values for burnup credit has raised 

awareness of the importance of determining accurate burnup values. An important factor responsible for 

the improved accuracy of assembly burnup values is the improvement in quality assurance (QA) 

programs throughout the nuclear power industry, encouraged by NRC. Improved computing resources 

and methods, use of continuous real-time core monitoring systems, increased awareness of the need for 

accurate assembly burnup data, and significant improvements in QA practices have resulted in improved 

reliability of the assembly burnup data. 

Factors that could negatively impact the accuracy of reactor burnup records include human error, 

transcription errors, and degradation of information storage media. Since the 1980s, utilities have 

maintained burnup records for individual assemblies using flux map processor data. While these data 

may be as accurate as noted in the reports above, the assembly burnup data have often been transcribed 

(possibly several times), input into new SNF tracking software, and/or stored on non-QA-audited media. 

Transcription errors for the burnup data or assembly identifiers or legibility issues with the media could 

result in “incorrect” information being used in cask loading operations. 
38

4.6 SUMMARY 

The burnup values assigned to fuel assemblies by the utilities are not measured. They are inferred from 

neutron flux measurements and from core thermal power, which is itself inferred from a reactor energy 

balance. Uncertainties associated with determining burnup values include (1) the uncertainty in the flux 

measurements, (2) the uncertainty in the conversion of the flux to assembly power, (3) uncertainty in the 

extrapolation and/or interpolation of assembly power to noninstrumented assemblies, (4) uncertainty in 

the determination of the core thermal power, and (5) uncertainty in the integration over time of assembly 

power to generate burnup values. Largely qualitative comparisons of the assembly-specific inferred 

burnup values, measured reactor physics test results, and core follow data with design predictions give 

some confidence that the inferred burnup values are close to the true burnup values. Quantitative 

comparisons showing the sensitivity of physics test results and core follow data to variation in assembly 

burnup would be helpful in bounding fuel assembly burnup uncertainty. 

Accurate determination of the assembly burnup, the amount of cumulative generated power per mass of 

fuel, depends on an accurate determination of core power distribution and core thermal power output 

history. Although the core power distribution that is predicted by validated and approved core design 

codes is acceptable, in-core reactor measurements are used on a regular basis to update these predictions 

to reflect the actual reactor operating conditions. Commercial flux map processing codes claim high 

accuracy for the resulting power distribution, and utility reactor records appear to be accurate to within 

5%. Multiple independent comparisons performed at different utilities between 1985 and 2004 involving 

several thousand in-core measured assembly burnup values report reactor record deviations of less than 

4.2% from core design predictions. A notable exception involves records from some utilities that report 

batch average burnup values. Also, the industry records do not take into account any inaccuracies in 

recording or transcribing the assembly burnup data into utility databases. Finally, note that there is some 

expectation that the accuracy of the reactor record values is fairly constant over the years due to 

competing effects; computer codes have become more accurate over time, tending to improve accuracy, 

while fuel designs and, to some extent, operations have become more complex, tending to reduce 

accuracy. 
39

5 CONSEQUENCES OF FUEL ASSEMBLY MISLOADING 

As discussed in Section 1, a loading curve (see Fig. 1.1) represents combinations of minimum burnup and 

initial enrichment that correspond to a limiting value of the effective neutron multiplication factor (keff) 

for a given configuration. Assemblies with insufficient burnup, as compared with the minimum required 

burnup dictated by the loading curve, are not acceptable for loading, and thus if loaded, would be 

considered to be “misloaded.” The goal of the NRC recommendation in ISG-8 for verifying reactor 

records by out-of-core measurement is to prevent inadvertent loading (i.e., misloading) of underburned 

assemblies. To understand the significance, and corresponding diligence with which such misloadings 

should be prevented, it is important to understand the consequences of potential assembly misloading on 

the system keff value and to evaluate the associated increases in keff against inherent margins (e.g., an 

administrative margin), where present. To support this understanding, a study45 was recently performed 

to determine the changes in keff that can result from a wide variety of postulated fuel misloading events in 

a representative high-capacity transport cask. The purpose of the study was to provide quantitative 

information for the impact of such events on the subcritical margin and aid in assessing the appropriate 

role for burnup measurements in ensuring safety. The discussion that follows summarizes the analysis 

and findings in Ref. 45. 

Misloading an underburned fuel assembly will cause an increase in the cask reactivity. The exact 

magnitude of the increase is dependent on several factors, including cask design, assembly design, 

assembly irradiation conditions, post-irradiation cooling time, assembly enrichment, assembly burnup, 

and assembly position within the cask. However, the magnitude of the increase is dominated by the 

amount by which the actual assembly burnup is less than the minimum required burnup value for loading 

acceptance and the position of the assembly within the cask. Therefore, the computational study was 

performed to investigate a variety of cases involving misloading of underburned assemblies, including 

cases involving misloading of assemblies with no burnup (i.e., with fresh fuel), into the center positions 

of the generic 32-PWR assembly (GBC-32) cask design. The GBC-32 was developed to be 

representative of typical high-capacity rail casks being considered by industry. Detailed specifications for 

the GBC-32 cask are provided in Computational Benchmark for Estimation of Reactivity Margin from 

Fission Products and Minor Actinides in PWR Burnup Credit.

46 In all cases considered for the study, the 

GBC-32 cask was assumed to be fully flooded with full-density water, and for all assemblies in all cases, 

Westinghouse 17 × 17 fuel and a cooling period of 5 years were assumed. To provide a bounding 

estimate of the effect of misloading an underburned assembly, the misloaded assemblies were assumed to 

be in the central positions for all cases. For all other potentially dependent factors, representative 

conditions were assumed (i.e., a representative high-capacity cask design, representative burnup credit 

depletion conditions, representative fuel assembly specifications, and a 5-year cooling time).45

The investigation of misloading underburned assemblies was performed to provide estimates of the effect 

of misloading events involving errors in assembly burnup verification, perhaps due to inaccurate reactor 

records, while the investigation of misloading fresh fuel with different enrichment values provides 

estimates for worst-case misloading events, potentially due to improper assembly identification. 

Although no attempt was made to quantify the probability of such misloading events, others47 have 

estimated that the probability of a single misloading can be as high as 10−3

 or as low as 10−5 for a large 

cask. If it is valid to assume that multiple misloading events in a single cask are statistically independent, 

the probability of misloading n assemblies can be estimated by raising the probability of a single 

misloading to the power n. Examination of known misloading events (see Section 6) reveals examples in 

which multiple misloads are not independent. 

The study considers two representative burnup points on cask loading curves developed with and without 

the principal fission products present and includes an investigation of (1) the effect of misloading fuel that 
40

has burnup values that are 90, 80, 50, 25, 10, and 0% of the minimum value required by the loading curve 

(i.e., underburned); (2) the effect of misloading fresh fuel with enrichment values of 2, 3, 4, and 5 wt % 

235U; and (3) for each scenario, the effect of misloading multiple (1, 2, 3, and 4) assemblies. The actual 

burnup and enrichment combinations used in the study were 30 GWd/MTU, 3.00 wt % 235U and 

45 GWd/MTU, 3.66 wt % 235U for conditions representing actinide-only burnup credit and analysis 

assumptions consistent with ISG-8 and 30 GWd/MTU, 3.88 wt % 235U and 45 GWd/MTU, 4.89 wt % 

235U for conditions representing credit for actinides and the principal fission products. 

For the representative burnup-enrichment combinations corresponding to actinide-only burnup credit, it 

was found that misloading one assembly with 0% of the required burnup (i.e., fresh fuel with enrichment 

of either 3.0 or 3.66 wt % 235U) results in an increase in keff of 0.02–0.035. Likewise, misloading two 

assemblies that are underburned by 75% (i.e., have only 25% of the required minimum burnup) results in 

an increase in keff of 0.025–0.035, while misloading four assemblies that are underburned by 50% also 

results in an increase in keff of 0.025–0.035. Because the reactivity reduction due to burnup is increased 

with the inclusion of credit for fission products, the impact of underburned fuel is also larger compared 

with corresponding cases without fission products present (i.e., actinide only). Consequently, misloading 

one assembly with 0% of the required burnup (i.e., fresh fuel with enrichment of either 3.88 or 4.89 wt % 

235U) results in an increase in keff of 0.035–0.055 compared with 0.02–0.035 for the actinide-only case. 

Misloading two assemblies underburned by 75% or four assemblies underburned by 50% results in an 

increase in keff of 0.035–0.045. For all cases considered, four simultaneous misloads involving 20% 

reduced burnup results in a maximum increase in keff of 0.0125, which is well within the 0.05 

administrative margin that is typically used in these types of criticality safety evaluations. Another 

notable observation is that for the cask and conditions considered, a reduction in burnup of 20% in all 

assemblies results in an increase in keff of less than 0.035. Note, in all cases considered, the “misloaded” 

assemblies were assumed to have reduced burnup; no cases were considered in which the burnup was 

increased as compared with the minimum required value. Also, in all cases, the remainder of the cask 

was assumed to be loaded with assemblies that just met the burnup-enrichment loading criteria (i.e., 

burnup and enrichment combination on the loading curve). Hence, the analysis made no attempt to credit 

the practical reality that a significant portion of the loaded assemblies may have higher-than-required 

burnup, which can compensate for the positive reactivity insertion associated with an underburned 

assembly. 

For misload conditions involving fresh fuel assemblies, without regard to the burnup-enrichment 

combinations from a loading curve, the results indicate that misloading a single fresh assembly with 3, 4, 

or 5 wt % 235U enrichment results in an increase in keff of approximately 2, 4, or 6%, respectively. 

Notably, a single fresh fuel assembly with 5.0 wt % 235U enrichment loaded into the cask center will 

result in an increase in keff of more than the 0.05 administrative margin, while misloading two fresh 

5 wt % 235U assemblies results in an increase in keff of more than 0.10. Because the burnup-enrichment 

combinations correspond to the same value of keff and the misloaded fresh fuel assemblies are not 

dependent on the reference burnup-enrichment combination or the nuclides included in the spent fuel 

compositions, these results were not dependent upon the presence of fission products. Throughout the 

analysis, the misloaded assemblies were placed in the most reactive (central) positions within the cask. 

Therefore, the impact of misloading assemblies into noncentral positions (e.g., nearer to the radial cask 

periphery) is bounded by the cases considered.45

In summary, the consequences to keff of loading assemblies that have slightly reduced burnup (e.g., 5% 

due to uncertainties in the burnup verification process), as compared with the required burnup, are fairly 

small (≤ 1%). On the other hand, loading one or more highly enriched (i.e., > 4 wt %) fresh fuel 

assemblies has a significant consequence on criticality safety. These findings suggest that while it may 

not be overly important to precisely verify the burnup value, it is important to ensure that fresh or 

very-low-burnup (nearly fresh) fuel assemblies are not misloaded into a cask. 
41

6 FUEL MOVEMENT AND MISLOADING EXPERIENCE 

The recommendation in ISG-8 for administrative procedures to include a measurement that confirms the 

reactor record for each assembly is intended to prevent unauthorized loading (misloading) of assemblies 

due to inaccuracies in reactor burnup records and/or improper assembly identification, thereby ensuring 

that the appropriate subcritical margin is maintained. Section 5of this report detailed the effects a misload 

(or multiple misloads) could have on criticality. It is important to recognize that misloading of SNF, 

although rare, does occur. 

Fuel-handling problems can be categorized into one of the following categories. 

1. The desired (correct) assembly can be selected and moved, but it is moved to an incorrect 

location. This category has two subsets. 

a. It can occur through human error (placing the assembly in location B when it was intended 

for placement in location A). 

b. It can occur through procedural/personnel error. When the assembly is purposely placed in 

location A, later review of the Technical Specifications may reveal that the assembly does not 

meet the requirements for being in location A, and it should have been placed in location B. 

This error can occur if incorrect data are used to select the assembly (e.g., incorrect burnup, 

cooling period, or initial assembly enrichment information), incorrectly transcribed 

information from databases/calculations is used, incorrect procedures are used, or correct 

information/procedures are applied incorrectly. 

2. Fuel can be misidentified (wrong assembly selected and then moved to the designated location). 

Utilities implement a number of safeguard methods to prevent fuel misloads. These methods include 

hardware, procedures, and redundancy of assembly verification. Each fuel assembly is manufactured 

with a unique identification number (ID) permanently installed on the upper assembly plate that enables 

effective visual confirmation of each assembly ID (Figs. 6.1–6.2). Plant procedures require independent 

verifications (by the qualified person performing the work and a second qualified person) of the assembly 

ID before each fuel movement. When all the fuel bundles have been moved to their intended locations—

either in the core, in the spent fuel pool, or in the spent fuel cask—independent verifications by another 

crew member are performed to compare the IDs of the as-loaded configuration to those in the map 

provided in the implementing design document. 

To evaluate the frequency and causes of such misloading/misplacement incidents, NRC’s Sequence 

Coding and Search System (SCSS) database for LERs was searched for incidents of misloaded fuel 

assemblies. Commercial nuclear plant licensees must submit LERs pursuant to 10 CFR 50.73 to 

document specific types of operational events, including (among others) operations or conditions 

prohibited by the plant’s Technical Specifications, degraded conditions of the plant or its principal safety 

barriers, operation in an unanalyzed condition that degrades plant safety, or events or conditions that 

could have prevented the fulfillment of the function of structures or systems important to safety. 

Misloaded assemblies of SNF typically occur in the spent fuel pool and are reported as violations of 

Technical Specifications. Core misplacements, though rare, would be reported as operation in an 

unanalyzed condition. It should be noted that the database of LERs covers a limited period of spent fuel 

pool movement experience, and that it is not clear that all misload events that occurred during that period 

are accurately reflected in the database. Given that this is the case, it is difficult to draw quantitative 

conclusions about spent fuel pool misload frequency and trends. 
42

Fig. 6.1. Drawing of a boiling water 

reactor fuel assembly with assembly 

identification number. 

Fig. 6.2. Picture of the top of a pressurized 

water reactor fuel assembly in a spent fuel pool 

showing the unique assembly identification 

number. 

The NRC’s LER database was searched to find and retrieve LERs reporting fuel-handling errors and 

misloaded SNF assemblies from the approximately 50,000 LERs reported from 1980 to early 2003, when 

the SCSS was discontinued. All subsequent LERs were manually reviewed to find additional pertinent 

events in the 2003–2006 time period. A number of LERs were identified as potential 

misload/misplacement events. Upon review, it was found that in all but one event the fuel assemblies 

were selected and handled correctly (i.e., consistent with the intent). However, the assemblies were 

subsequently determined to have been loaded in locations in the spent fuel pool contrary to technical 

specification requirements as a result of calculational, procedural, or other human errors. Thus, the 

assemblies were mispositioned; they were not incorrectly identified or mishandled. Reactor records were 

responsible for two of the LERs. A total of 19 LERs that reported pertinent fuel-handling errors were 

located and are summarized below.48

• (LER 36682036) In 1982, Hatch Unit 2 personnel loaded the wrong fuel assembly around a source 

range monitor. Bridge personnel mistakenly picked up a bundle adjacent to the desired bundle in the 

spent fuel pool. Thus, the cause was personnel error. 

• (LER 36683031) In 1983, Hatch Unit 2 personnel loaded eight fuel assemblies in an improper core 

orientation around a source range monitor. Plant personnel entered the fuel assembly data into the 

plant computer incorrectly by entering the data as Unit 1 data instead of Unit 2 data. Thus, the cause 

was personnel error. 

• (LER 36985035) In 1985, McGuire Unit 1 personnel identified seven spent B&W fuel assemblies 

stored side by side in the spent fuel pool in a manner contrary to Technical Specification 

requirements. The seven assemblies did not meet the burnup criteria for unrestricted placement in 

Region 2 of the spent fuel pool and should have been stored in a checkerboard pattern. This occurred 

due to personnel error in not adequately reviewing the Technical Specifications before approving the 

fuel movement. 
43

• (LER 26685005) In 1985, a Point Beach Unit 1 QA audit determined that three spent fuel assemblies 

that had been discharged from the reactor less than 1 year had been stored next to the spent fuel pool 

wall in violation of plant Technical Specifications. 

• (LER 21987006) In 1987, Oyster Creek Unit 1 personnel temporarily stored 184 fresh fuel assemblies 

in the spent fuel pool that had enrichments above Technical Specification requirements. The safety 

analysis for the fresh fuel was prepared based on the assumption that the fuel would be stored only in 

the dry storage facility. The cause of the violation was personnel error. A Technical Specification 

modification was subsequently requested to allow storage of the more highly enriched fuel in the 

spent fuel pool. 

• (LER 28687008) In 1987, Indian Point Unit 3 personnel identified a fresh fuel assembly that had been 

inadvertently loaded in the spent fuel pool in a manner contrary to Technical Specification 

requirements. The misloading was due to personnel error. 

• (LER 30287026) In 1987, Crystal River Unit 3 personnel moved a fresh fuel assembly in the spent 

fuel pool in a manner contrary to Technical Specification requirements. The misloading of the 

assembly was promptly identified, and the assembly was relocated to an approved location. The 

misloading was due to personnel error. 

• (LER 28088028) In 1988, Surry Unit 1 personnel found that they had placed a spent fuel assembly in 

a location in the spent fuel pool in a manner contrary to Technical Specification requirements. The 

spent fuel assembly had not decayed for a sufficiently long time period to be stored in the region in 

which it was placed. Subsequently all spent fuel in that region of the spent fuel pool was confirmed 

to meet plant Technical Specifications. The cause of the misloading was procedural errors. 

• (LER 36991016) In 1991, McGuire Unit 1 personnel identified 11 fuel assemblies that had been 

stored in the Unit 1 spent fuel pool in a manner contrary to Technical Specification requirements. 

The vacant row provision of the Technical Specifications had not been met. The problem was 

determined to be due to a procedural deficiency because the guidance provided by the applicable 

procedure was obscure. 

• (LER 45494006) In 1994, Byron Unit 1 personnel identified a fuel assembly located in the wrong 

region of the spent fuel pool. The assembly did not meet the Technical Specification requirement for 

burnup. The Nuclear Component Transfer List (NCTL) incorrectly specified the placement of the 

assembly into a region not approved for the actual assembly burnup. The cause was personnel error. 

A search by plant staff found one previous event of a misloaded fuel assembly due to an error in the 

NCTL. (This second event was not found in this LER database review but was noted in the subject 

LER by plant staff.) 

• (LER 45696007) In 1996, Braidwood Unit 1 personnel placed spent fuel in an inappropriate 

configuration during a Boraflex neutron absorber testing procedure. Staff did not consider the effects 

of the assemblies on lower-burnup fuel in adjacent storage locations. The cause was procedural, 

personnel, and managerial error. 

• (LER 45496008) In 1996, Byron Unit 1 personnel identified three fuel assemblies located in a region 

of the spent fuel pool in a manner contrary to Technical Specification requirements. The assemblies 

did not meet the Technical Specification requirements for minimum burnup. The computer 

spreadsheet used to verify minimum burnup contained erroneous information and had not been 

independently verified. Subsequently, all SNF assemblies in region 2 of the spent fuel pool were 

verified to be in their correct locations per Technical Specifications. 
44

• (LER 45696008) In 1996, Braidwood Unit 1 personnel identified a spent fuel assembly located in a 

region of the spent fuel pool in a manner contrary to Technical Specification requirements. The cause 

was personnel error. The fuel assembly did not meet the required checkerboard configuration based 

on burnup vs initial enrichment specified in the Technical Specifications. A calculation contained the 

wrong burnup information and was not independently verified. In this LER, Braidwood staff noted 

one other fuel misloading event (457-200-94-016), in which new fuel was misloaded in the spent fuel 

pool during transfer from the new fuel storage vault. The cause of the event was personnel error in 

not following procedures. 

• (LER 52999003) In 1999, Palo Verde Unit 2 personnel identified a spent fuel assembly stored in the 

wrong position in the spent fuel pool (based on the initial enrichment and assembly burnup factors). 

This misloading occurred due to personnel error and poor independent verification. A subsequent 

review of the spent fuel configuration of all Unit 1, 2, and 3 assemblies confirmed that out of 

approximately 2052 assemblies, only the one assembly was in the incorrect location. 

• (LER 34800004) In 2000, Farley Unit 1 personnel identified three spent fuel assemblies in spent fuel 

pool locations not allowed by Technical Specifications. The assemblies had insufficient burnup for 

the storage locations in which they were placed. The cause was personnel error, lack of detail in the 

procedure, and insufficient independent review. Subsequently all Unit 2 spent fuel pool assemblies 

were confirmed to be in the correct locations per Technical Specifications. 

• (LER 27502006) In 2002, Diablo Canyon Unit 1 personnel determined that a spent fuel assembly had 

been stored in a manner contrary to Technical Specification requirements. The assembly had 

insufficient burnup for its location adjacent to other spent fuel assemblies. The cause was personnel 

error. Subsequent to this event, the utility verified the spent fuel pool storage configuration for the 

previous 5 years and found no other discrepancies. 

• (LER 24404002) In 2004, Ginna personnel identified three (old) consolidated rod storage canisters in 

spent fuel pool locations not allowed by Technical Specifications. A subsequent review indicated that 

in early storage calculations, the total energy generated by the fuel rods contained in the canisters was 

incorrectly calculated. Later, when the canisters were moved in the spent fuel pool, only the total 

canister data (for each canister) was reviewed, not the data on the individual assemblies in each 

canister. The cause was personnel error for using incorrect burnup values calculated for the fuel 

assemblies inside the three canisters. 

• (LER 44504001) In 2004, while performing a complete verification of fuel assembly configurations 

in the spent fuel pool, Comanche Peak personnel identified one fuel assembly loaded in a manner 

contrary to Technical Specification requirements. The cause was an incorrect burnup value that had 

been calculated for the fuel assembly. The incorrect burnup value resulted from inadequate 

conversion of the data files during a computer code migration in 1998. During a subsequent 

verification of all assembly configurations in the spent fuel pool, several other discrepancies in fuel 

assemblies were identified. These discrepancies were compared against the applicable Technical 

Specifications, and it was concluded that all the affected fuel assemblies (other than the initial 

assembly) were acceptable in their existing storage locations. 

• (LER 31106004) In 2006, while reviewing the Salem Unit 2 move sheet for the fifteenth refueling 

outage core reload, it was discovered that a fuel assembly had been placed in a region of the spent 

fuel pool without meeting the minimum burnup requirements for that storage location. During the 

subsequent investigation, it was discovered that 11 other fuel assemblies had also been 

inappropriately discharged into similar locations in the Unit 2 spent fuel pool without meeting the 
45

minimum burnup requirements, and 1 assembly had been inappropriately placed in the Salem Unit 1 

spent fuel pool. The cause of the occurrences was attributed to a lack of technical rigor associated 

with the review and verification of assembly burnup values before discharge into the spent fuel pools; 

ultimately, personnel error. 

Another misload event was identified in the NRC “Morning Reports.” In November 2000, the utility 

Consumer’s Power notified Region III that a total of 11 spent fuel assemblies had been placed into five 

VSC-24 spent fuel storage casks at the Palisades Plant before undergoing 5 years of cooling time (as 

required by the cask CoC) following their discharge from the reactor core. (They had cooled for just over 

4 years.) The event was due to an error in recording the core discharge dates for these assemblies. This 

event was not treated as a criticality issue but was evaluated from a heat-load standpoint and does 

represent a misload event. 

During the preparation of this report, a Certificate of Compliance Violation Report49 related to the 

misloading of 34 fuel assemblies in four HI-STORM 100 model casks at Grand Gulf was filed by 

Entergy. The fuel assemblies “exceeded the maximum allowed Decay Heat per Fuel Storage Location 

Limit and/or the Fuel Burnup Limit.” The misloading was preliminarily attributed to an error in the Cask 

Loader Database, which was used to select the fuel assemblies that were loaded into each cask. The Cask 

Loader Database was relied upon to ensure that assemblies selected for loading met the applicable 

exposure and cooling time limits provided in the CoC. Similar to the situation with the VSC-24 cask, this 

event does not represent a criticality concern, but does represent a misload event. 

A review of the above-listed LERs illuminates a number of issues. Only one LER (36682036) deals with 

the misidentification of a spent fuel assembly. Of the other LERs, 16 LERs describe spent fuel pool 

misloadings of assemblies due to deficient procedures or human error in not following approved 

procedural requirements. Two LERs were related to a reactor record error. The details behind LER 

24404002 (2004, Ginna) show that the spent fuel assemblies had been sent off-site as part of a testing 

program and were returned several years later to the site spent fuel pool. The SNF was stored in 

consolidated-rod storage canisters in the spent fuel pool. The site Nuclear Fuel Accounting Code 

contained canister data but not data for the individual fuel assemblies that were contained in the canisters, 

even though the site evidently did have individual assembly records on file. 

Another important observation from this review is that there is very little data on misloads in spent fuel 

casks—of all the information reviewed, only two fuel assembly misloading events involved spent fuel 

casks. Hence, the applicability of the above misload data to cask loading operations is not entirely clear. 

The existence of relatively little data for cask misloading events is expected since fuel movements into 

spent fuel casks have been relatively few, as compared to fuel movements associated with the core and 

spent fuel pool. 

Note that the review of the LER database for this report certainly did not find all SNF misloading events 

that have occurred at utilities. The two misloading events referenced in the LERs listed above and in the 

“Morning Reports” discussing the assembly-cooling-time issue are indications of other misloading 

events. Even so, the frequency of misloads of assemblies appears to be low. Per the 2002 RW-859 

report,44 as of the end of 2002, 163,646 SNF assemblies in the United States had been permanently 

discharged from reactors. The U.S. reactors discharged approximately 29,000 additional spent fuel 

assemblies between 2003 and 2006. It seems reasonable to assume that each assembly has been moved at 

least eight times (once into the spent fuel pool as fresh fuel, once into the core for cycle 1, shuffled two 

times each for cycles 2 and 3, once back into the spent fuel pool as burned fuel, and once within the spent 

fuel pool as the fuel ages and cools). A significant number of assemblies will have been moved more 

frequently, as spent fuel pool contents are rearranged due to heat load distribution issues, spent fuel pool 

reracking, space management realignments, support for outage needs, etc. Thus, a reasonable estimate of 
46

fuel assembly moves may range from about 8 moves per assembly to 10 or more moves per assembly. 

Considering the large number of assembly moves and the relatively small number of misloading events, 

the frequency of occurrence is rather small and given the numerous lessons learned, particularly in terms 

of procedural development and compliance, this frequency should decrease with time. However, more 

complex loading patterns and increases in the number and types of fuel assemblies present in a spent fuel 

pool appear to act against such a decrease in this frequency, making it difficult to draw quantitative 

conclusions about spent fuel pool misload frequency and trends in the absence of more data. 

The LERs evaluated show that the probability of selecting an assembly with the wrong ID or placing the 

assembly in a position other than the position specified by procedure is very low (only one occurrence in 

the 19 LERs listed). The probability of the procedures being in error or of human error occurring in 

implementing the procedures, while also very low, was the predominant factor in the other assembly 

misloading events. Additionally, note that as corrective actions for many of these LERs, the affected 

utilities verified that the assemblies in the spent fuel pools were stored in their correct locations according 

to plant Technical Specifications. In so doing, because other fuel misloadings were not discovered, the 

utilities demonstrated that their spent and fresh fuel movement and storage programs did not have 

additional and/or systematic errors. 

A major criticality concern when loading casks is the potential for misloading fresh or very low burnup 

(nearly fresh) fuel into positions in lieu of spent fuel assemblies. Although several of the noted LERs 

dealt with fresh fuel being placed in locations that were not appropriate per plant Technical 

Specifications, none of the LERs reviewed indicated any occurrence of misloading fresh fuel when the 

intent was to load (or move) a spent fuel assembly. Note, however, that this provides little confidence 

that such a misload has not occurred. It is important to recognize that the physical appearance of fresh 

fuel is significantly different from fuel that has been irradiated for even a relatively short period of time 

(less than one cycle). A fresh fuel assembly has a shiny surface, whereas the surface of an irradiated fuel 

assembly will have undergone slight oxidation, resulting in a dull reddish appearance. Also, visible 

encrustation may have built up on an irradiated fuel assembly, and the assembly may be slightly bent or 

twisted. However, there may be cases where plants removed a leaking or damaged fuel assembly only 

days or weeks after it was freshly loaded into the core and subsequently stored it on-site for months, or 

even years. This “slightly burned” SNF may or may not appear to be more heavily irradiated fuel. If this 

slightly burned assembly was mistaken for heavily burned SNF due to human or administrative error, the 

consequences could be serious. Procedures requiring visual inspection of the fuel before it is loaded into 

a cask exist at utilities to prevent the misloading of assemblies into a spent fuel cask. Frequently utility 

procedures require two-person independent verification of the assembly ID, and the loading procedure is 

often videotaped. No instances of fresh fuel being mistaken for spent fuel were found in the 

documentation. 

Although the information reviewed for this report did not include an extensive amount of non-U.S. data, 

the following information was noted. In a paper included in an ANS conference proceedings,50 the 

French authors noted that the Dampierre Nuclear Power Plant experienced a fuel misloading event during 

a core reload activity. In April 2001, a total of 113 assemblies had been placed in the wrong positions 

before the error was discovered. This error occurred when the 26th assembly was inadvertently 

substituted for the 25th assembly in the reload sequence, with the result that all of the following 

assemblies were also placed in incorrect locations within the core.50 The error was detected when an 

operator noticed that the distribution of rod cluster assemblies in the reactor vessel was not in accordance 

with the expected symmetrical pattern. The cause of the misload incident was operator error. It was 

noted that the neutron source range monitors could not detect a local increase in reactivity unless a 

reactive pattern was formed close to one of the monitors. There had been no attempt to take the reactor 

critical because the incident was discovered before any actions were taken to reinstall the vessel head. 
47

7 SUMMARY DISCUSSIONS 

The current NRC regulatory guidance for burnup credit (ISG-8, Rev. 2) includes a recommendation for a 

measurement confirming assembly burnup to ensure that an incorrect loading in a burnup credit cask with 

a more reactive assembly than permitted by the loading criteria does not occur. A burnup measurement is 

recommended, in part, because of concerns regarding the completeness and accuracy of utility reactor 

burnup records for all SNF and concerns associated with the potential for calculation, procedural, and 

other human errors in selection of assemblies for loading. The purpose of this report is to detail 

information and issues relevant to preshipment burnup measurements when using burnup credit in PWR 

SNF transport and storage casks. In particular, this report reviews the role of burnup measurements in the 

regulatory guidance for demonstrating compliance with burnup loading criteria, burnup measurement 

capabilities and experience, generation and accuracy of utility burnup records, fuel movement and 

misloading experience, and the consequences of misloading assemblies in casks designed for burnup 

credit. 

7.1 OUT-OF-CORE MEASUREMENTS 

As it is not possible to nondestructively measure assembly fissile content or burnup directly, burnup 

measurement techniques/devices measure neutron and gamma radiation emitted from an assembly and 

use this information, along with calibration data, to infer the assembly burnup. The various 

techniques/devices used to measure assembly burnup were originally developed for nuclear materials 

safeguards purposes, have limitations in their applicability for highly accurate burnup determination, and 

have advantages and disadvantages relative to each other. Generally, during fork detector examinations, 

assemblies have been measured at only one position along their length. If the measurements at this 

position are representative of burnup (and the equipment calibrations against assemblies of like geometry 

and known burnup accurately account for axial variations in assembly burnup), the relative burnup values 

of the assemblies can be obtained. Previously established correlations between burnup and the passive 

radiation levels then verify that the measurements are consistent with operator-declared values for burnup. 

Out-of-core measurements of gamma and/or neutron radiation that infer assembly burnup require 

calibration of the counting equipment with an assembly of known burnup and age and of identical 

geometry. Thus, the accuracy of the assembly burnup inferred from the gamma-ray and/or neutron 

measurements cannot be more accurate than the accuracy of the burnup of the reference assembly used 

for calibration. Additionally, the approach used in the analysis of the SNF during each measurement 

campaign was to accumulate measurements from a number of assemblies and generate an internal 

calibration by comparing each assembly with the best-derived fit of all the site data. During out-of-core 

examinations, any observed deviations from the best-fit curves would incorporate the uncertainties in the 

out-of-core measurements as well as any errors in the reactor records. The average deviations were 

therefore likely to be upper bounds on the random errors in the reactor records for assembly burnup. 

Several of the referenced reports state that the fork detector system demonstrates an average deviation 

from the burnup calibration curves of 2.2–5% (Refs. 4,9,14). The specific results noted in this report 

show the average deviation of the fork detector, when compared with calibration curves developed from 

reactor record burnup data, was between 2.7 and 3.5%, with maximum assembly deviations of up to 

9.1%. The high-resolution gamma-ray scans were not as accurate and in the ANO campaign averaged a 

deviation from the calibration curve of 5.37% at the 67% confidence level. In any case, physical 

measurements of the assembly neutron and gamma radiation do not provide an absolute burnup 

measurement but instead provide a measurement of burnup based on information provided in the utility 

reactor records and serve to confirm the information in the reactor records. The effort to conduct the 

out-of-core measurement of SNF involves additional handling of the inspected assemblies, which 

increases the risk of assembly damage or misloading, increases the radiation dose to personnel conducting 
48

examinations, adversely impacts other fuel-handling activities, and increases the financial costs to the 

utilities. 

Although not discussed previously in this report, subcritical measurement techniques51 could be used for 

verification of subcriticality, as opposed to burnup, and hence have some relevance. However, the 

viability and feasibility of subcritical measurements for large systems such as transport casks have not 

been demonstrated and it is recognized that significant technical and operational challenges would be 

associated with any attempt to use subcritical measurements for verification of subcriticality in storage 

and transport casks. 

7.2 REACTOR BURNUP RECORDS 

Utility-developed assembly burnup records are based on either (1) the measured core thermal output, with 

burnup distributed to individual assemblies using validated computer codes, or (2) a combination of 

information provided by in-core detectors, measured core thermal output, and validated computer codes. 

The total core power calculation has an uncertainty of about 1% (Refs. 4, 9, and 31). In the 1989 report, 

The Incentives and Feasibility for Direct Measurement of Spent Nuclear Fuel Characteristics in the 

Federal Waste Management System,

4

 an analysis of data for Zion Unit 2, cycle 1, indicated that it is 

appropriate to use the extensive in-core measurements of power distribution (flux maps) to develop the 

end-of-cycle assembly average burnup values. Using the total core power information and in-core flux 

maps, assembly average burnup values were determined to within about 2% of the predicted burnup 

values when computer models, correctly normalized to start-of-cycle conditions and adjusted periodically 

on the basis of in-core measurements, were used.4

 Multiple, independent comparisons4,5,9,12,14,34,40,41 show 

that the in-core measured burnups are within 1.79% and 4.2% of predicted burnups, and based on 

qualitative indications, the measured burnups are probably closer to the true burnups than to the nodal 

predicted burnups. Based on these comparisons, it may be concluded that the uncertainty in the 

utility-assigned burnup values is less than 5%. 

One area of concern for reactor record accuracy is older fuel, including fuel that has been treated 

differently from the vast majority of SNF at the site. For example, some spent fuel at utilities was 

shipped off-site for various test programs and was subsequently returned to the site at a much later date. 

The records for this fuel are much more suspect than the records for fuel that has not been handled by 

multiple organizations. Also, as noted in Section 6of this report, Ginna and possibly other utilities store 

some spent fuel in consolidated rod storage canisters in the spent fuel pool. The site records may contain 

canister data but not data for individual assemblies contained in the canisters. If burnup is to be used as a 

loading criterion for such fuel, it is important that utilities verify the characteristics of the assemblies that 

have been handled or stored in these atypical conditions. Also, “batch-average” assembly burnup data, 

which is much less accurate than individual assembly burnup data, has been used in at least one case and 

should not be used in operations to support the loading of assemblies into burnup credit transport casks 

where loading criteria include assembly-specific burnup values. The core-follow and in-core monitoring 

data for SNF is stored by utilities and should be used to reconstruct assembly-specific burnup data where 

appropriate. 

7.3 MISLOADS IN A PWR BURNUP CREDIT CASK 

The analyses in the report Criticality Analysis of Assembly Misload in a PWR Burnup Credit Cask45 show 

that fresh fuel misloads into a transport cask can have a significant effect on criticality safety. For 

misload conditions involving fresh fuel assemblies, the results indicate that misloading a single fresh 

assembly with 3, 4, or 5 wt % 235U enrichment results in an increase in keff of approximately 2, 4, or 6%, 
49

respectively. Notably, a single fresh fuel assembly with 5.0 wt % 235U enrichment loaded into the cask 

center will result in an increase in keff of more than 0.05, while misloading two fresh 5 wt % 235U 

assemblies results in an increase in keff of more than 0.10. Also, misloading two assemblies that are 

underburned by 75% results in an increase in keff of 0.025–0.035, while misloading four assemblies that 

are underburned by 50% results in an increase in keff of 0.025–0.035.45

Concerning the possibility of mistaking a fresh assembly for an irradiated assembly when loading a 

transport cask, it is important to note that the physical appearance of fresh fuel is significantly different 

from fuel that has been irradiated, even for a relatively short time (less than one cycle). A fresh fuel 

assembly has a shiny surface, whereas the surface of an irradiated fuel assembly will have undergone 

slight oxidation resulting in a dull reddish layer. Also, visible encrustation may have built up on an 

irradiated fuel assembly, and the assembly may be slightly bent or twisted. “Slightly burned” fuel that 

has received only days or weeks of irradiation may exist in some utilities’ spent fuel pools. This slightly 

burned SNF may or may not give the appearance of more heavily irradiated fuel. If this slightly burned 

fuel were mistaken for heavily burned SNF due to human or administrative error, the consequences could 

be a serious impact to the safety margin. 

Rigorous and comprehensive procedural requirements for preparing, reviewing, and conducting fuel 

transfers to ensure that the proper fuel assemblies are loaded into the correct cask position could suffice to 

preclude a fuel misload that violates the criticality loading curve criteria. For example, two fuel-handling 

operators should independently verify assembly identifiers at each stage of the burnup verification and 

cask loading procedures. Reactor record assembly burnup information should be independently verified. 

This can be done by independent out-of-core measurement of the assembly or by verifying the original 

core-follow burnup data against the in-core measurement data. These procedural measures ensure proper 

assembly selection and records assignment. 

Procedures requiring visual inspections and/or simple field measurements (e.g., gross radiation 

measurements, Cerenkov radiation detection) to be performed before loading SNF into a cask may be 

sufficient to prevent fresh or nearly fresh fuel from being mistakenly loaded into a burnup credit cask. 

Any additional measurement program would have to be evaluated to determine its limitations to detect the 

various ages and irradiation levels of SNF. For example, the DCVDs may not be able to differentiate 

between very old, moderately burned SNF and new, very lightly burned SNF. Also, equipment alignment 

is critical to Cerenkov examinations and is especially critical when examining fuel that has been cooled 

for long periods of time. Near-neighbor assemblies and pool water quality (boron concentration) can also 

affect the Cerenkov examination results. 

7.4 MISLOADING EVENTS 

A review of known fuel misloading experience indicated that utility fuel-handling errors are rare but have 

occurred. More than 1.5 million fuel-handling activities have occurred since the NRC LER tracking 

system was developed, with 19 LERs noting assembly misload events. These included 18 LERs for 

assemblies not meeting the Technical Specifications for the locations in which they were stored and 

1 LER involving moving an incorrectly identified assembly. Two misloading events involving SNF casks 

were identified. It should be noted that the database of LERs covers a limited period of spent fuel pool 

movement experience, and that it is not clear that all misload events that occurred during that period are 

accurately reflected in the database. Given that this is the case, it is difficult to draw quantitative 

conclusions about spent fuel pool misload frequency and trends. Considering the number of fuelhandling activities, the frequency of reported misload events is low. 
50

7.5 ISSUES RELATED TO OLDER SPENT NUCLEAR FUEL 

Utilities use various NRC-approved methods to determine the end-of-cycle assembly average burnup 

values that are recorded in reactor records. However, the methodologies used have changed over time 

and older data are potentially less accurate than newer data. Utility use of flux map data, design codes 

(and code constants), and full-core measured/inferred radial power distributions has improved over time, 

decreasing the burnup uncertainty of newer spent fuels. Because burnup values of older fuels were not 

recalculated with these incremental improvements, the burnup uncertainty factors used for older fuels 

may be larger than those for newer fuels by an undetermined amount. Other characteristics of older SNF 

include less complex fuel designs, lower initial assembly enrichments of 235U (2–3%), and lower 

assembly average burnup values. Because utilities have not uniformly or consistently calculated SNF 

burnup using the same methods, it may be useful for the industry to provide information on the codes and 

methods that have been historically used to develop assembly reactor record burnup information. 

Also, the extent and/or rigor of QA verification given to the data entry, transcription, and information 

verification may be inconsistent among utilities. A number of utilities have already made commitments 

to NRC to develop and implement plans to validate databases used in the control of spent fuel pool 

storage. As the utilities migrate their SNF databases to newer software, they frequently conduct 

verification programs to ensure that the data are transcribed correctly. As noted in several of the 

referenced LERs, these verification programs can involve confirmation of the utility spent fuel pool 

configurations by verifying assembly IDs and assembly burnup against Technical Specifications for 

assembly location within the spent fuel pool. This verification of both the spent fuel pool configurations 

and the SNF databases serves to validate not only the assembly locations within the spent fuel pool but 

also the accurate assignment of the burnup and cooling time to the proper assemblies in the reactor 

records. 

7.6 METHODS FOR CALCULATING BURNUP 

A number of existing reports discuss varying methods and criteria for developing a “final” assembly 

burnup value when there is a difference between reactor record burnup values and out-of-core burnup 

measurement values.7,14,32,52 Two methods for using reactor records of assembly burnup for developing 

this final assembly burnup value are as follows. 

• Method 1—Use the burnup value from the reactor record, when confirmed to be within a predefined 

range by an out-of-core burnup measurement, reduced by the uncertainty in the reactor record. 

• Method 2—Use the burnup value from an out-of-core burnup measurement, as calibrated by the 

reactor record data, reduced by the uncertainty in the out-of-core burnup measurement, as determined 

statistically from all of the measurements in the calibration. 

Each of the two methods relies on a calibration of the measurement system by measuring a large number 

of geometrically identical assemblies and using reactor record burnup data for these same assemblies to 

establish the calibration. 

The first method has the advantage of simplicity in that the assembly burnup can be determined before 

out-of-core measurement from reactor data and is subject only to confirmation by the out-of-core 

measurement. The second method has the advantage that both the burnup and its uncertainty are 

measurement based.52 

A third possible method using a less robust measurement system than that required in method 2 above 

would involve using verified reactor burnup records, which would be required to have a clearly defined 
51

and quantified uncertainty optionally coupled with a simple measurement (e.g., Cerenkov light 

measurement) to prevent the misloading of fresh and slightly burned fuel assemblies. This method would 

have the advantages of simplicity, potentially no additional assembly-handling operations, and the 

confirmation that fresh fuel is not mistaken for spent fuel. 
52
53

8 OBSERVATIONS 

Over the past 20 years, multiple SNF out-of-core examinations have been conducted at U.S. utilities to 

compare utility records for assembly burnup and cooling time with out-of-core measurement data. The 

measurement equipment used [neutron/gamma-ray detection systems (e.g., the “Fork Detector” system 

from Los Alamos National Laboratory) and HRGS systems available commercially from several vendors] 

was developed to detect fissile material diversion and was not specifically designed for ease of use in 

reactor spent fuel pools or optimized to accurately verify or quantify SNF burnup independent of utility 

reactor records. However, the data gathered from these examination campaigns have been fairly uniform 

and consistent and are useful for evaluating the effectiveness of these measurement techniques/equipment 

and the general accuracy of reactor burnup records. Although the out-of-core burnup measurement 

techniques and equipment used in the examination campaigns are rather dated, published information on 

SNF burnup measurement campaigns involving more modern techniques/equipment could not be found. 

In addition to the out-of-core measurement campaigns to evaluate assembly burnup versus reactor 

records, some utilities have retroactively applied significant effort to “verify” the data in their SNF 

records. They have done this by comparing the assembly burnup records with burnup calculations 

performed using NRC-accepted design codes and in-core monitoring measurements including 

consideration of cycle-specific total power output and full-core measured/inferred radial power 

distributions. In a number of cases, utilities have upgraded the software used to track their SNF and have 

used this opportunity to verify the data input accuracy of their reactor records. Other utilities have 

reverified their assembly records to ensure that the burnup information on their spent fuel pool assemblies 

matches the burnup calculations calculated for the fuel when it was removed from the reactor cores. 

The following observations highlight key points from the information reviewed for this report. 

Reactor Burnup Records

• Utility records for fuel burnup are based on either (1) the measured core thermal output, with burnup 

distributed to individual assemblies using validated computer codes, or (2) a combination of 

information provided by in-core detectors, measured core thermal output, and validated computer 

codes. 

• There is a significant amount of data available from 1980 to the present to support a finding that 

utility records for fuel burnup are accurate for individual spent fuel assemblies to at least 5% of “true” 

assembly burnup. These data originate primarily from reactor core-follow data, ex-core burnup 

measurement programs, comparisons between calculated burnup values (on which reactor record 

values are based) and burnup values inferred from in-core measurements, and retrospective 

evaluations based on comparisons between existing reactor record values and calculated values. 

• Utilities do not all use the same methods to calculate and verify the assembly burnup values that are 

recorded in their reactor records. Moreover, the computational methods used by many utilities have 

evolved over time such that burnup values for older and newer fuel are based on different methods. 

Hence, for SNF assemblies to be transported in burnup credit casks, utilities need to demonstrate how 

the burnup values in the reactor records were developed and recorded and document information as to 

the accuracy of their recorded burnup values. Examples of this type of activity are the reactor record 

verification programs of Duke Energy and TVA described in this report. 

• In at least one case, burnup values in a utility’s reactor records for some assemblies are based on the 

“batch average” burnup, which is definitely not an accurate representation of the individual burnup 
54

for each of the assemblies. The number of utilities that may have used batch average data is 

unknown. However, based on a review of various data sources for discharged fuel, it is expected that 

the use of batch averaged values is limited in number and would apply only to older fuel records. 

Contemporary reactor records use individual, assembly-specific burnup calculations and values. 

• Unless a QA program is implemented to ensure the accuracy of reactor records, reactor records may 

become less accurate over time as a result of transcription errors, as record media degrade or are 

changed to newer media, or as software is updated. 

Burnup Verification Measurements

• Out-of-core measurement systems can adequately verify reactor record burnup information. However, 

as is also the case for in-core measurement systems, out-of-core measurement systems cannot measure 

fuel burnup directly. Instead, these out-of-core measurement systems measure gamma-ray and/or 

neutron emissions from the assemblies, which are then compared to a calibration curve to develop an 

estimated fissile content and corresponding assembly burnup. 

• The burnup measurement programs reviewed for this report concluded that out-of-core measurement 

systems provided somewhat less accurate burnup values, as compared to reactor record burnup values 

developed using reactor in-core monitoring systems and design codes. The modern in-core systems, 

which use core measurements and some data from design codes, produce burnup values that are 

generally within approximately 2% of the burnup values predicted by design codes, whereas the outof-core measurement systems produced average assembly burnup results that were within the expected 

fork detector accuracy range of 2.2–5% of predicted burnup. However, there were some fork detector 

examinations with maximum assembly deviations as high as 9.1%. Regardless, these examination 

programs demonstrated that out-of-core measurements could identify substantially underburned fuel 

assemblies and hence could be used to prevent them from being accidentally loaded into spent fuel 

transport casks. 

• Careful calibration against an assembly of known burnup, known cooling time, and identical geometry 

is required to achieve the reported accuracies with the fork detector used in the U.S. studies. Thus, in 

practice, these out-of-core measurements are dependent on and calibrated against reactor burnup 

records. Consequently, the burnup values inferred from these measurements tend to have a higher 

uncertainty than reactor record assembly burnup values determined using in-core measurements. 

• Fuel assembly axial-burnup profiles have a significant impact on reactivity and are therefore an 

important component in determining “average assembly burnup.” For the fork detector examination 

programs evaluated in this report, personnel used measurements at the assembly centerline (midplane) 

and assumed that the axial profile of the reference assembly could be used to estimate the assembly 

average burnup (i.e., it was assumed that the axial profiles were the same for the reference and 

measured assemblies). This approach could give erroneous burnup measurement results for 

assemblies that have different axial burnup characteristics. Hence, if out-of-core measurements are 

used, care should be taken to ensure that correct average burnup information is collected, 

commensurate with the measurement accuracy goals and criteria for approved contents. 

• The costs and risks associated with performing out-of-core burnup measurements should be balanced 

with the risks and potential costs of not performing the measurements. Out-of-core measurement 

campaigns require utility resources for planning and execution, increase the dose to personnel, 

increase the risk of damage to assemblies and potential fuel mishandling events due to the increased 

assembly movements, and have associated financial cost to the utility. The risks and potential costs 
55

associated with loading a significantly underburned assembly into a transport or storage configuration 

have not been explored in this report. 

• The neutron-counting measurement systems appear to be more accurate than the HRGS systems and 

require less skilled operators for handling the SNF assemblies. 

• Considerable efforts, primarily motivated by interests related to nuclear material safeguards, are 

ongoing to develop better and more accurate measurement systems. 

Consequences of Fuel Assembly Misloading 

• The consequences to keff of loading assemblies that have slightly reduced burnup (e.g., 5% due to 

uncertainties in the burnup verification process), as compared with the required burnup, are fairly 

small (≤ 1%). On the other hand, loading one or more highly enriched (i.e., > 4 wt %) fresh fuel 

assemblies has a significant consequence on criticality safety. These findings suggest that while it 

may not be overly important to precisely verify the burnup value, it is important to ensure that fresh or 

very-low-burnup (i.e., nearly fresh) fuel assemblies are not misloaded into a cask. 

Fuel Movement and Operational Considerations

• Although utilities’ record of reliability in selecting and moving assemblies during fuel-handling 

processes may be characterized as “good” from the standpoint that no inadvertent criticalities have 

occurred as a result of fuel misloading, there are a number of documented examples of fuel 

misloading. Considering the large number of fuel assembly movements that have been executed, 

relatively few mishandling events involving movement of an incorrectly identified fuel assembly have 

been reported. Procedural violations related to SNF movements that resulted in violations of plant 

Technical Specifications have occurred with a fairly low frequency. Most of these events were not a 

result of incorrect burnup values assigned to the SNF but were instead the result of personnel error in 

selecting assemblies for movement. 

• Fresh fuel is visibly different from SNF due to the oxidation, crud buildup, and bending/twisting of the 

latter, so visual inspection should easily differentiate new assemblies from SNF. However, there is 

some uncertainty about the appearance of an assembly with very limited burnup (e.g., removed 

promptly after start-up due to leaking fuel rods or some other problem) after it has resided in a spent 

fuel pool for a number of years. 

• Visual inspections and/or simple field measurements (e.g., gross radiation measurements, Cerenkov 

radiation detector) could be performed during cask loading to detect and prevent accidental loading of 

fresh or nearly fresh fuel. A basic detector system could be devised to ensure that an assembly has 

some minimum activity level. One examination tool that may suffice is the DCVD. Any 

measurement program would have to be evaluated to determine its limitations in detecting SNF. For 

example, the DCVDs may not be able to differentiate between very old, moderately burned SNF and 

very-lightly-burned (i.e., nearly fresh) SNF. Also, any examination limitations, such as equipment 

alignment for Cerenkov examinations, must be evaluated to ensure that adequate examination results 

are achievable for all SNF to be evaluated. 
56
57 

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OCTOBER 2009