ASTM E944-19

Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance

SIGNIFICANCE AND USE
3.1 Adjustment methods provide a means for combining the results of neutron transport calculations with neutron dosimetry measurements (see Test Method E1005 and NUREG/CR-5049) in order to obtain optimal estimates for neutron damage exposure parameters with assigned uncertainties. The inclusion of measurements reduces the uncertainties for these parameter values and provides a test for the consistency between measurements and calculations and between different measurements (see 3.3.3). This does not, however, imply that the standards for measurements and calculations of the input data can be lowered; the results of any adjustment procedure can be only as reliable as are the input data.  
3.2 Input Data and Definitions:  
3.2.1 The symbols introduced in this section will be used throughout the guide.  
3.2.2 Dosimetry measurements are given as a set of reaction rates (or equivalent) denoted by the following symbols:
These data are, at present, obtained primarily from radiometric dosimeters, but other types of sensors may be included (see 4.1).  
3.2.3 The neutron spectrum (see Terminology E170) at the dosimeter location, fluence or fluence rate Φ(E) as a function of neutron energy E , is obtained by appropriate neutronics calculations (neutron transport using the methods of discrete ordinates or Monte Carlo, see Guide E482). The results of the calculation are customarily given in the form of multigroup fluences or fluence rates.
where:
Ej and Ej+1  are the lower and upper bounds for the j-th energy group, respectively, and k is the total number of groups.  
3.2.4 The reaction cross sections of the dosimetry sensors are obtained from an evaluated cross section file. The cross section for the i-th reaction as a function of energy E will be denoted by the following:
Used in connection with the group fluences, Eq 2, are the calculated group-averaged cross sections σij. These values are defined through the following equation:
...
SCOPE
1.1 This guide covers the analysis and interpretation of the physics dosimetry for Light Water Reactor (LWR) surveillance programs. The main purpose is the application of adjustment methods to determine best estimates of neutron damage exposure parameters and their uncertainties.  
1.2 This guide is also applicable to irradiation damage studies in research reactors.  
1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
1.4 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

General Information

• Status: Published
• Publication Date: 30-Sep-2019
• ICS: 27.120.20 - Nuclear power plants. Safety
• Technical Committee: E10 - Nuclear Technology and Applications
• Drafting Committee: E10.05 - Nuclear Radiation Metrology

Relations

• Replaces: ASTM E944-13e1 - Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance
• Effective Date: 01-Oct-2019
• Refers: ASTM E1018-20 - Standard Guide for Application of ASTM Evaluated Cross Section Data File
• Effective Date: 01-Mar-2020
• Refers: ASTM E854-19 - Standard Test Method for Application and Analysis of Solid State Track Recorder (SSTR) Monitors for Reactor Surveillance
• Effective Date: 01-Nov-2019
• Refers: ASTM E704-19 - Standard Test Method for Measuring Reaction Rates by Radioactivation of Uranium-238
• Effective Date: 01-Oct-2019
• Refers: ASTM E705-18 - Standard Test Method for Measuring Reaction Rates by Radioactivation of Neptunium-237
• Effective Date: 01-Dec-2018
• Refers: ASTM E263-18 - Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Iron
• Effective Date: 01-Dec-2018
• Refers: ASTM E844-18 - Standard Guide for Sensor Set Design and Irradiation for Reactor Surveillance
• Effective Date: 01-Jun-2018
• Refers: ASTM E910-18 - Standard Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance
• Effective Date: 01-Feb-2018
• Refers: ASTM E526-17 - Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Titanium
• Effective Date: 01-Aug-2017
• Refers: ASTM E262-17 - Standard Test Method for Determining Thermal Neutron Reaction Rates and Thermal Neutron Fluence Rates by Radioactivation Techniques
• Effective Date: 01-Aug-2017
• Refers: ASTM E170-17 - Standard Terminology Relating to Radiation Measurements and Dosimetry
• Effective Date: 01-Jun-2017
• Refers: ASTM E170-16a - Standard Terminology Relating to Radiation Measurements and Dosimetry
• Effective Date: 01-Oct-2016
• Refers: ASTM E170-16 - Standard Terminology Relating to Radiation Measurements and Dosimetry
• Effective Date: 15-Feb-2016
• Refers: ASTM E170-15a - Standard Terminology Relating to Radiation Measurements and Dosimetry
• Effective Date: 01-Sep-2015
• Refers: ASTM E1005-15 - Standard Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance
• Effective Date: 01-Jul-2015

Overview

ASTM E944-19 – Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance provides comprehensive guidance on the analysis and interpretation of neutron dosimetry in light water reactor (LWR) surveillance programs as well as research reactor irradiation studies. The primary purpose of this standard is to outline the application of neutron spectrum adjustment methods, which combine computational neutron transport results and physical dosimetry measurements to achieve best estimates of neutron damage exposure parameters with well-defined uncertainties. The guide emphasizes that reduced uncertainties through spectrum adjustment do not compensate for inadequacies in input data quality-the reliability of the outcome depends on the precision of both measurements and calculations.

Key Topics

• Neutron Spectrum Adjustment Methods: The guide explains how adjustment algorithms refine estimates of neutron fluence and reaction rates by integrating both measured and calculated data. Statistical techniques like the least squares method are highlighted for producing unbiased minimum variance estimates.
• Input Data Quality: Details are provided on the necessary input data, including radiometric measurements from dosimetry sensors, neutron spectra derived from transport calculations, and associated cross section data. Proper documentation of variances and covariances for all input parameters is required.
• Uncertainty Analysis: Emphasis is placed on the reduction and propagation of uncertainties. Adjustment methods are used to detect inconsistencies in input data, supporting robust validation processes.
• Selection and Use of Computer Codes: The guide discusses requirements and comparative properties of commonly used neutron spectrum adjustment and unfolding codes, such as STAY’SL, FERRET, LEPRICON, LSL-M2, among others, stressing the importance of using codes that provide comprehensive output uncertainties.

Applications

ASTM E944-19 is essential for professionals involved in:

• Reactor Vessel Surveillance: Provides methodologies to accurately estimate neutron exposure and damage parameters, supporting the safe operation and longevity of reactor pressure vessels.
• Dosimetry Program Development: Offers best practices for selecting and preparing sensor sets, conducting radiometric dosimetry, and performing data analysis to monitor neutron fields effectively.
• Regulatory Compliance & Benchmarking: Supports meeting regulatory requirements by ensuring validated, consistent estimations of neutron exposure using internationally recognized techniques.
• Research Reactor Studies: Applicable to experimental studies, where accurate irradiation damage quantification is critical for material testing and development.

By applying ASTM E944-19, reactor operators and engineers can achieve:

• Reduced uncertainties in damage parameter estimation
• Increased reliability in lifetime assessments of reactor components
• Early detection and correction of inconsistencies in dosimetry data

Related Standards

Several other standards and referenced documents complement and support the guidance in ASTM E944-19, including:

• ASTM E170: Radiation measurements and dosimetry terminology
• ASTM E1005: Radiometric monitors for reactor vessel surveillance
• ASTM E482: Neutron transport methods for reactor surveillance
• ASTM E844: Sensor set design for reactor surveillance
• ASTM E1018: Application of evaluated cross section data files
• NUREG/CR-1861, NUREG/CR-5049: U.S. Nuclear Regulatory Commission guidance on dosimetry model fitting and pressure vessel fluence analysis

For a full list of relevant referenced standards, users are encouraged to consult the standard itself.

Practical Value

Applying ASTM E944-19 enhances the precision and reliability of neutron dosimetry in reactor environments. By following its rigorous methods for input data selection, uncertainty quantification, and use of validated computational tools, organizations can support the integrity, safety, and regulatory compliance of both commercial power and research reactors.

Keywords: neutron spectrum adjustment, dosimetry, neutron fluence, reactor surveillance, uncertainty analysis, irradiation damage, ASTM E944-19, light water reactor, research reactor, neutron exposure parameters

Standards Content (Sample)

This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the
Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.
Designation: E944 − 19
Standard Guide for
Application of Neutron Spectrum Adjustment Methods in
Reactor Surveillance
This standard is issued under the fixed designation E944; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision. A number in parentheses indicates the year of last reapproval. A
superscript epsilon (´) indicates an editorial change since the last revision or reapproval.
1. Scope E265 Test Method for Measuring Reaction Rates and Fast-
Neutron Fluences by Radioactivation of Sulfur-32
1.1 This guide covers the analysis and interpretation of the
E266 Test Method for Measuring Fast-Neutron Reaction
physics dosimetry for LightWater Reactor (LWR) surveillance
Rates by Radioactivation of Aluminum
programs. The main purpose is the application of adjustment
E393 Test Method for Measuring Reaction Rates by Analy-
methods to determine best estimates of neutron damage expo-
sis of Barium-140 From Fission Dosimeters
sure parameters and their uncertainties.
E481 Test Method for Measuring Neutron Fluence Rates by
1.2 This guide is also applicable to irradiation damage
Radioactivation of Cobalt and Silver
studies in research reactors.
E482 Guide for Application of Neutron Transport Methods
1.3 This standard does not purport to address all of the
for Reactor Vessel Surveillance
safety concerns, if any, associated with its use. It is the
E523 Test Method for Measuring Fast-Neutron Reaction
responsibility of the user of this standard to establish appro-
Rates by Radioactivation of Copper
priate safety, health, and environmental practices and deter-
E526 Test Method for Measuring Fast-Neutron Reaction
mine the applicability of regulatory limitations prior to use.
Rates by Radioactivation of Titanium
1.4 This international standard was developed in accor-
E693 Practice for Characterizing Neutron Exposures in Iron
dance with internationally recognized principles on standard-
and Low Alloy Steels in Terms of Displacements Per
ization established in the Decision on Principles for the
Atom (DPA)
Development of International Standards, Guides and Recom-
E704 Test Method for Measuring Reaction Rates by Radio-
mendations issued by the World Trade Organization Technical
activation of Uranium-238
Barriers to Trade (TBT) Committee.
E705 Test Method for Measuring Reaction Rates by Radio-
2. Referenced Documents
activation of Neptunium-237
E706 MasterMatrixforLight-WaterReactorPressureVessel
2.1 ASTM Standards:
E170 Terminology Relating to Radiation Measurements and Surveillance Standards
Dosimetry E844 Guide for Sensor Set Design and Irradiation for
E262 Test Method for Determining Thermal Neutron Reac-
Reactor Surveillance
tion Rates and Thermal Neutron Fluence Rates by Radio-
E853 Practice forAnalysis and Interpretation of Light-Water
activation Techniques
Reactor Surveillance Neutron Exposure Results
E263 Test Method for Measuring Fast-Neutron Reaction
E854 Test Method for Application and Analysis of Solid
Rates by Radioactivation of Iron
State Track Recorder (SSTR) Monitors for Reactor Sur-
E264 Test Method for Measuring Fast-Neutron Reaction
veillance
Rates by Radioactivation of Nickel
E910 Test Method for Application and Analysis of Helium
Accumulation Fluence Monitors for Reactor Vessel Sur-
veillance
This guide is under the jurisdiction of ASTM Committee E10 on Nuclear
E1005 Test Method for Application and Analysis of Radio-
Technology and Applications and is the direct responsibility of Subcommittee
E10.05 on Nuclear Radiation Metrology. A brief overview of Guide E944 appears metric Monitors for Reactor Vessel Surveillance
in Master Matrix E706 in 5.4.1.
E1018 Guide for Application of ASTM Evaluated Cross
Current edition approved Oct. 1, 2019. Published October 2019. Originally
ɛ1 Section Data File
approved in 1983. Last previous edition approved in 2013 as E944 – 13 . DOI:
10.1520/E0944-19. E2005 Guide for Benchmark Testing of Reactor Dosimetry
For referenced ASTM standards, visit the ASTM website, www.astm.org, or
in Standard and Reference Neutron Fields
contact ASTM Customer Service at [e-mail]. For Annual Book of ASTM
E2006 Guide for BenchmarkTesting of LightWater Reactor
Standards volume information, refer to the standard’s Document Summary page on
the ASTM website. Calculations
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
E944 − 19
2.2 Nuclear Regulatory Commission Documents: where:
NUREG/CR-1861 PCA Experiments and Blind Test E and E are the lower and upper bounds for the j-th energy
j j+1
NUREG/CR-2222 Theory and Practice of General Adjust- group, respectively, and k is the total number of groups.
ment and Model Fitting Procedures
3.2.4 The reaction cross sections of the dosimetry sensors
NUREG/CR-3318 LWR Pressure Vessel Surveillance Do-
are obtained from an evaluated cross section file. The cross
simetry Improvement Program: PCA Experiments, Blind
section for the i-th reaction as a function of energy E will be
Test, and Physics-Dosimetry Support for the PSF Experi-
denoted by the following:
ment
σ E , i 51,2, … (3)
~ !
i
NUREG/CR-3319 LWR Power Reactor Surveillance
Used in connection with the group fluences, Eq 2, are the
Physics-Dosimetry Data Base Compendium
NUREG/CR-5049 Pressure Vessel Fluence Analysis and calculated group-averaged cross sections σ . These values are
ij
defined through the following equation:
Neutron Dosimetry
4 E
j11
2.3 Electric Power Research Institute:
σ 5 Φ E σ E dE/Φ (4)
* ~ ! ~ !
ij i j
E
j
EPRI NP-2188 Development and Demonstration of an Ad-
i 51,2,.n;
vanced Methodology for LWR Dosimetry Applications
j 5 1,2, … k
2.4 Government Document:
NBSIR 85–3151 Compendium of Benchmark Neutron
3.2.5 Uncertainty information in the form of variances and
Fields for Reactor Dosimetry
covariances must be provided for all input data. Appropriate
corrections must be made if the uncertainties are due to bias
3. Significance and Use
producing effects (for example, effects of photo reactions).
3.1 Adjustment methods provide a means for combining the
3.3 Summary of the Procedures:
resultsofneutrontransportcalculationswithneutrondosimetry
3.3.1 Anadjustmentalgorithmmodifiesthesetofinputdata
measurements(seeTestMethodE1005andNUREG/CR-5049)
as defined in 3.2 in the following manner (adjusted quantities
in order to obtain optimal estimates for neutron damage
are indicated by a tilde, for example, ã):
i
exposureparameterswithassigneduncertainties.Theinclusion
a˜ 5 a 1∆a (5)
of measurements reduces the uncertainties for these parameter
i i i
values and provides a test for the consistency between mea-
˜
Φ E 5 Φ E 1∆Φ E (6)
~ ! ~ ! ~ !
surements and calculations and between different measure-
ments (see 3.3.3). This does not, however, imply that the or for group fluence rates
standards for measurements and calculations of the input data
˜
Φ 5 Φ 1∆Φ (7)
j j j
can be lowered; the results of any adjustment procedure can be
σ˜ E 5 σ E 1∆σ E , (8)
~ ! ~ ! ~ !
only as reliable as are the input data. i i i
3.2 Input Data and Definitions:
or for group-averaged cross sections
3.2.1 The symbols introduced in this section will be used
σ˜ 5 σ 1∆σ (9)
ij ij ij
throughout the guide.
The adjusted quantities must satisfy the following condi-
3.2.2 Dosimetry measurements are given as a set of reaction
rates (or equivalent) denoted by the following symbols: tions:
`
a , i 51,2, … (1)
˜
i
a˜ 5 Φ~E!σ˜ ~E!dE, i 51,2, …n (10)
*
i i
These data are, at present, obtained primarily from radio-
or in the form of group fluence rates
metric dosimeters, but other types of sensors may be included
k
(see 4.1).
˜
a˜ 5 σ˜ Φ , i 51,2, …n (11)
i ( ij j
3.2.3 The neutron spectrum (see Terminology E170)atthe
j51
dosimeterlocation,fluenceorfluencerate Φ(E)asafunctionof
Since the number of equations in Eq 11 is much smaller than
neutron energy E, is obtained by appropriate neutronics calcu-
the number of adjustments, there exists no unique solution to
lations (neutron transport using the methods of discrete ordi-
the problem unless it is further restricted. The mathematical
nates or Monte Carlo, see Guide E482). The results of the
algorithms in current adjustment codes are intended to make
calculation are customarily given in the form of multigroup
theadjustmentsassmallaspossiblerelativetotheuncertainties
fluences or fluence rates.
of the corresponding input data. Codes like STAY’SL,
E
j11
Φ 5 * Φ ~E!dE, j 51,2, … k (2) FERRET, LEPRICON, and LSL-M2 (see Table 1) are based
j
E
j
explicitly on the statistical principles such as “Maximum
Likelihood Principle” or “Bayes Theorem,” which are gener-
alizations of the well-known least squares principle, and are
Available from U.S. Government Printing Office, Superintendent of
taking into account variances and correlations of the input
Documents, 732 N. Capitol St., NW, Washington, DC 20401-0001, http://
fluence, dosimetry, and cross section data (see 4.1.1, 4.2.2, and
www.access.gpo.gov.
4.3.3). A detailed discussion of the mathematical derivations
Available from the Electric Power Research Institute, 3420 Hillview Avenue,
Palo Alto, California 94304, http://www.epri.com. can be found in NUREG/CR-2222 and EPRI NP-2188. Even
E944 − 19
TABLE 1 Available Neutron Spectrum Adjustment and Unfolding Codes
Code Available Refer-
Program Solution Method Comments
From ences
A
SAND-II semi-iterative RSICC Prog. No. CCC- (1, 2) contains trial spectra library. No output uncertainties in the
112, CCC-619, PSR- original code, but modified Monte Carlo code provides output
345 uncertainties (3)
SPECTRA statistical, linear estimation RSICC Prog. No. CCC- (4, 5) minimizes deviation in magnitude, no output uncertainties.
IUNFLD/ statistical, linear estimation (6) constrained weighted linear least squares code using B-spline
UNFOLD basic functions. No output uncertainties.
WINDOWS statistical, linear estimation, linear RSICC Prog. No. PSR- (7) minimizes shape deviation, determines upper and lower bounds
programming 136, 161 for integral parameter and contribution of foils to bounds and
estimates. No statistical output uncertainty.
RADAK, statistical, linear estimation RSICC Prog. No. PSR- (8, 9, 10, 11) RADAK is a general adjustment code not restricted to spectrum
SENSAK 122 adjustment.
STAY’SL statistical linear estimation RSICC Prog. No. PSR- (12) permits use of full or partial correlation uncertainty data for
113 activation and cross section data.
NEUPAC(J1) statistical, linear estimation RSICC Prog. No. PSR- (13, 14) permits use of full covariance data and includes routine of
177 sensitivity analysis.
FERRET statistical, least squares with log normal RSICC Prog. No. PSR- (15, 16) flexible input options allow the inclusion of both differential and
a priori distributions 145 integral measurements. Cross sections and multiple spectra may
be simultaneously adjusted. FERRET is a general adjustment
code not restricted to spectrum adjustments.
LEPRICON statistical, generalized linear least RSICC Prog. No. PSR- (17, 18, 19) simultaneous adjustment of absolute spectra at up to two
squares with normal a priori and a 277 dosimetry locations and one pressure vessel location. Combines
posteriori distributions integral and differential data with built-in uncertainties. Provides
reduced adjusted pressure vessel group fluence covariances
using built-in sensitivity database.
LSL-M2 statistical, least squares, with log normal RSICC Prog. No. 20 simultaneous adjustment of several spectra. Provides
a priori and a posteriori distributions PSR-233 covariances for adjusted integral parameters. Dosimetry cross-
section file included.
UMG Statistical, maximum entropy with output RSICC Prog. No. (21, 22) Two components. MAXED is a maximum entropy code. GRAVEL
uncertainties PSR-529 (23) is an iterative code.
NMF-90 Statistical, least squares IAEA NDS (24, 25) Several components, STAY’NL, X333, and MIEKE. Distributed by
IAEA as part of the REAL-84 interlaboratory exercise on
spectrum adjustment (26).
GMA Statistical, general least squares RSICC Prog. No. (27 ) Simultaneous evaluation with differential and integral data,
PSR-367 primarily used for cross-section evaluation but extensible to
spectrum adjustments.
A
The boldface numbers in parentheses refer to the list of references appended to this guide.
theoldercodes,notablySAND-IIandCRYSTALBALL,apply fluences and simultaneous adjustment. LSL-M2 also allows
a minimization algorithm although the statistical assumptions simultaneous adjustment, but cross correlations must be given.
are not spelled out explicitly in the supporting documentation. 3.3.2 The adjusted data ã, etc., are, for any specific
i
Table 1 lists some of the available unfolding codes; however, algorithm, unique functions of the input variables. Thus,
the first four codes listed: SAND-II, SPECTRA, IUNFLD/ uncertainties (variances and covariances) for the adjusted
UNFOLD,andWINDOWShaveseverelimitationsinthatthey parameters can, in principle, be calculated by propagation the
do not typically provide uncertainty characterization of the uncertainties for the input data. Linearization may be used
resulting unfolded spectrum and the adjusted damage exposure before calculating the uncertainties of the output data if the
parameters. adjusted data are nonlinear functions of the input data.
3.3.1.1 An important problem in reactor surveillance is the 3.3.2.1 The algorithms of the adjustment codes tend to
determinationofneutronfluenceinsidethepressurevesselwall decrease the variances of the adjusted data compared to the
at locations which are not accessible to dosimetry. Estimates correspondinginputvalues.Thelinearleastsquaresadjustment
for exposure parameter values at these locations can be codes yield estimates for the output data with minimum
obtained from adjustment codes which adjust fluences simul- variances, that is, the “best” unbiased estimates. This is the
taneouslyatmorethanonelocationwhenthecrosscorrelations primary reason for using these adjustment procedures.
between fluences at different locations are given. LEPRICON 3.3.3 Properly designed adjustment methods provide means
has provisions for the estimation of cross correlations for to detect inconsistencies in the input data which manifest
E944 − 19
themselves through adjustments that are larger than the corre- exposure parameters and their variances should ideally be part
spondinguncertaintiesorthroughlargevaluesofchi-square,or of the adjustment code.
both. (See NUREG/CR-3318 and NUREG/CR-3319.) Any
detection of inconsistencies should be documented, and output 4. Selection of Input Data
data obtained from inconsistent input should not be used. All
4.1 Sensor Sets:
input data should be carefully reviewed whenever inconsisten-
4.1.1 Radiometric Measurements (RM)—This is at present
cies are found, and efforts should be made to resolve the
the primary source for dosimetry data in research and power
inconsistencies as stated below.
reactors. RM sensor selection, preparation, and measurement,
3.3.3.1 Input data should be carefully investigated for evi-
including determination of variances and covariances, should
dence of gross errors or biases if large adjustments are
be made according to Guide E844 and the standards describing
required. Note that the erroneous data may not be the ones that
thehandlingoftheparticularfoilmaterial(TestMethodsE262,
required the largest adjustment; thus, it is necessary to review
E263, E264, E265, E266, E393, E481, E523, E526, E704, and
all input data. Data of dubious validity may be eliminated if
E705). Other passive dosimetry sensors of current interest in
proper corrections cannot be determined. Any elimination of
research and power reactors and in ex-vessel environments are
data must be documented and reasons stated which are
solid state track recorders (SSTR), helium accumulation flu-
independent of the adjustment procedure. Inconsistent data
ence monitors (HAFM), and damage monitors (DM). Use of
may also be omitted if they contribute little to the output under
these sensors is described in separate ASTM standards as
investigation.
follows:
3.3.3.2 Inconsistencies may also be caused by input vari- 4.1.2 SSTR—see Test Method E854.
ances which are too small. The assignment of uncertainties to 4.1.3 HAFM—see Test Method E910.
the input data should, therefore, be reviewed to determine
4.1.4 The preceding list does not exclude the use of other
whether the assumed precision and bias for the experimental
integral measurements, for example, from fission chambers or
and calculational data may be unrealistic. If so, variances may
nuclear emulsions (see NUREG/CR-1861).
be increased, but reasons for doing so should be documented.
4.1.5 Accurate dosimetry measurements and proper selec-
Note that in statistically based adjustment methods, listed in
tions of dosimetry sensors are particularly important if the
Table 1 the output uncertainties are determined only by the
uncertainties in the calculated spectrum are large (see Ref 28).
input uncertainties and are not affected by inconsistencies in
In this case, it is necessary either to have several dosimetry
the input data (see NUREG/CR-2222). Note also that too large
sensors which respond to various parts of the neutron energy
adjustments may yield unreliable data because the limits of the
range of interest or to utilize a sensor which closely approxi-
linearization are exceeded even if these adjustments are con-
mates the energy response of the damage exposure parameters.
sistent with the input uncertainties.
Since determination of a variety of damage exposure param-
eters is desirable, some combination of dosimeter responses is
3.3.4 Using the adjusted fluence spectrum, estimates of
usually necessary to achieve the smallest possible output
damage exposure parameter values can be calculated. These
uncertainties. Reactions currently used which are regarded as
parameters are weighted integrals over the neutron fluence
providing the best overlap with the iron dpa cross section are
`
˜
237 93 93m
p 5 * Φ~E!w~E!dE (12)
Np(n,f) and Nb(n,n') Nb. Other reactions used to mea-
o
63 46 54
sureneutronsabove1MeVare Cu(n,α), Ti(n,p), Fe(n,p),
58 238
or for group fluences
Ni(n,p), and U(n,f). (See Practice E853.) If the calculated
k
spectrum has small uncertainties, the requirements of good
˜
p 5 Φ w (13)
( j j
spectral coverage or good overlap with damage response are
j51
not as critical, but redundant dosimetry is still recommended to
with given weight (response) functions w(E)or w , respec-
j minimize chances of erroneous results. (See Refs 28, 29.)
235 239
tively.TheresponsefunctionfordpaofironislistedinPractice
4.1.6 Non-threshold sensors such as U(n,f), Pu(n,f),
E693. Fluence greater than 1.0 MeVor fluence greater than 0.1
andall(n,γ)reactionsarefrequentlyused.Thesedetectorshave
MeV is represented as w(E)=1 for E above the limit and
the highest sensitivity at low neutron energies (below 1 keV)
w(E)=0for E below.
andareusefulforthevalidationofcalculatedspectrainthelow
3.3.4.1 Finding best estimates of damage exposure param- energy range and for the estimation of effects caused by low
235 239
eters and their uncertainties is the primary objective in the use
energy neutrons (for example, U fission and Pu fission in
of adjustment procedures for reactor surveillance. If calculated U, etc.).They are not as important as the threshold reactions
according to Eq 12 or Eq 13, unbiased minimum variance
for the determination of damage exposure parameters values
estimates for the parameter p result, provided the adjusted but can serve as useful supplements, particularly in the
fluence Φ˜ is an unbiased minimum variance estimate. The determination of iron dpa (see Ref 28).
variance of p can be calculated in a straightforward manner
4.1.7 The number of reactions used in an adjustment pro-
from the variances and covariances of the adjusted fluence cedure need not be large as long as the energy range under
spectrum. Uncertainties of the response functions, w, if any,
investigation is adequately covered. A small number of well-
j
should not be considered in the calculation of the output established dosimetry sensors combined with high-quality
variances when a standard response function, such as the dpa
measuring procedures is preferable to a large number of
for iron in Practice E693, is used. The calculation of damage measurements which include inconsistent or irrelevant data.
E944 − 19
4.2 Calculations: grossly inco
...

This document is not an ASTM standard and is intended only to provide the user of an ASTM standard an indication of what changes have been made to the previous version. Because
it may not be technically possible to adequately depict all changes accurately, ASTM recommends that users consult prior editions as appropriate. In all cases only the current version
of the standard as published by ASTM is to be considered the official document.
´1
Designation: E944 − 13 E944 − 19
Standard Guide for
Application of Neutron Spectrum Adjustment Methods in
Reactor Surveillance
This standard is issued under the fixed designation E944; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision. A number in parentheses indicates the year of last reapproval. A
superscript epsilon (´) indicates an editorial change since the last revision or reapproval.
ε NOTE—The title of this guide and the Referenced Documents were updated editorially in May 2017.
1. Scope
1.1 This guide covers the analysis and interpretation of the physics dosimetry for Light Water Reactor (LWR) surveillance
programs. The main purpose is the application of adjustment methods to determine best estimates of neutron damage exposure
parameters and their uncertainties.
1.2 This guide is also applicable to irradiation damage studies in research reactors.
1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility
of the user of this standard to establish appropriate safety safety, health, and healthenvironmental practices and determine the
applicability of regulatory limitations prior to use.
1.4 This international standard was developed in accordance with internationally recognized principles on standardization
established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued
by the World Trade Organization Technical Barriers to Trade (TBT) Committee.
2. Referenced Documents
2.1 ASTM Standards:
E170 Terminology Relating to Radiation Measurements and Dosimetry
E262 Test Method for Determining Thermal Neutron Reaction Rates and Thermal Neutron Fluence Rates by Radioactivation
Techniques
E263 Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Iron
E264 Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Nickel
E265 Test Method for Measuring Reaction Rates and Fast-Neutron Fluences by Radioactivation of Sulfur-32
E266 Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Aluminum
E393 Test Method for Measuring Reaction Rates by Analysis of Barium-140 From Fission Dosimeters
E481 Test Method for Measuring Neutron Fluence Rates by Radioactivation of Cobalt and Silver
E482 Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance
E523 Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Copper
E526 Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Titanium
E693 Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements Per Atom (DPA)
E704 Test Method for Measuring Reaction Rates by Radioactivation of Uranium-238
E705 Test Method for Measuring Reaction Rates by Radioactivation of Neptunium-237
E706 Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards
E844 Guide for Sensor Set Design and Irradiation for Reactor Surveillance
E853 Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Neutron Exposure Results
E854 Test Method for Application and Analysis of Solid State Track Recorder (SSTR) Monitors for Reactor Surveillance
E910 Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance
E1005 Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance
This guide is under the jurisdiction of ASTM Committee E10 on Nuclear Technology and Applications and is the direct responsibility of Subcommittee E10.05 on Nuclear
Radiation Metrology. A brief overview of Guide E944 appears in Master Matrix E706 in 5.4.1.
Current edition approved Jan. 1, 2013Oct. 1, 2019. Published January 2013October 2019. Originally approved in 1983. Last previous edition approved in 20082013 as
ɛ1
E944 – 08.E944 – 13 . DOI: 10.1520/E0944-13E01.10.1520/E0944-19.
For referenced ASTM standards, visit the ASTM website, www.astm.org, or contact ASTM Customer Service at [e-mail]. For Annual Book of ASTM Standards
volume information, refer to the standard’s Document Summary page on the ASTM website. DOI: 10.1520/E0944-08.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
E944 − 19
E1018 Guide for Application of ASTM Evaluated Cross Section Data File
E2005 Guide for Benchmark Testing of Reactor Dosimetry in Standard and Reference Neutron Fields
E2006 Guide for Benchmark Testing of Light Water Reactor Calculations
2.2 Nuclear Regulatory Commission Documents:
NUREG/CR-1861 PCA Experiments and Blind Test
NUREG/CR-2222 Theory and Practice of General Adjustment and Model Fitting Procedures
NUREG/CR-3318 LWR Pressure Vessel Surveillance Dosimetry Improvement Program: PCA Experiments, Blind Test, and
Physics-Dosimetry Support for the PSF Experiment
NUREG/CR-3319 LWR Power Reactor Surveillance Physics-Dosimetry Data Base Compendium
NUREG/CR-5049 Pressure Vessel Fluence Analysis and Neutron Dosimetry
2.3 Electric Power Research Institute:
EPRI NP-2188 Development and Demonstration of an Advanced Methodology for LWR Dosimetry Applications
2.4 Government Document:
NBSIR 85–3151 Compendium of Benchmark Neutron Fields for Reactor Dosimetry
3. Significance and Use
3.1 Adjustment methods provide a means for combining the results of neutron transport calculations with neutron dosimetry
measurements (see Test Method E1005 and NUREG/CR-5049) in order to obtain optimal estimates for neutron damage exposure
parameters with assigned uncertainties. The inclusion of measurements reduces the uncertainties for these parameter values and
provides a test for the consistency between measurements and calculations and between different measurements (see 3.3.3). This
does not, however, imply that the standards for measurements and calculations of the input data can be lowered; the results of any
adjustment procedure can be only as reliable as are the input data.
3.2 Input Data and Definitions:
3.2.1 The symbols introduced in this section will be used throughout the guide.
3.2.2 Dosimetry measurements are given as a set of reaction rates (or equivalent) denoted by the following symbols:
a , i 5 1,2,… (1)
i
These data are, at present, obtained primarily from radiometric dosimeters, but other types of sensors may be included (see 4.1).
3.2.3 The neutron spectrum (see Terminology E170) at the dosimeter location, fluence or fluence rate Φ(E) as a function of
neutron energy E, is obtained by appropriate neutronics calculations (neutron transport using the methods of discrete ordinates or
Monte Carlo, see Guide E482). The results of the calculation are customarily given in the form of multigroup fluences or fluence
rates.
E
j11
Φ 5 Φ ~E!dE, j 5 1,2, … k (2)
*
j
E
j
where:
E and E are the lower and upper bounds for the j-th energy group, respectively, and k is the total number of groups.
j j+1
3.2.4 The reaction cross sections of the dosimetry sensors are obtained from an evaluated cross section file. The cross section
for the i-th reaction as a function of energy E will be denoted by the following:
σ ~E!, i 5 1,2, … (3)
i
Used in connection with the group fluences, Eq 2, are the calculated group-averaged cross sections σ . These values are defined
ij
through the following equation:
E
j11
σ 5 Φ E σ E dE/Φ (4)
* ~ ! ~ !
ij i j
E
j
i 5 1,2,.n;
j 51,2,…k
3.2.5 Uncertainty information in the form of variances and covariances must be provided for all input data. Appropriate
corrections must be made if the uncertainties are due to bias producing effects (for example, effects of photo reactions).
3.3 Summary of the Procedures:
3.3.1 An adjustment algorithm modifies the set of input data as defined in 3.2 in the following manner (adjusted quantities are
indicated by a tilde, for example, ã ):
i
˜a 5 a 1Δa (5)
i i i
Available from Superintendents of Documents, U. S. U.S. Government Printing Office, Washington, DC 20402.Superintendent of Documents, 732 N. Capitol St., NW,
Washington, DC 20401-0001, http://www.access.gpo.gov.
Available from the Electric Power Research Institute, P. O. Box 10412, 3420 Hillview Avenue, Palo Alto, CA 94303.California 94304, http://www.epri.com.
E944 − 19
˜
Φ E 5Φ E 1ΔΦ E (6)
~ ! ~ ! ~ !
or for group fluence rates
˜
Φ 5Φ 1ΔΦ (7)
j j j
σ˜ E 5σ E 1Δσ E , (8)
~ ! ~ ! ~ !
i i i
or for group-averaged cross sections
σ˜ 5σ 1Δσ (9)
ij ij ij
The adjusted quantities must satisfy the following conditions:
`
˜
a˜ 5 Φ E σ˜ E dE, i 5 1,2,…n (10)
* ~ ! ~ !
i i
or in the form of group fluence rates
k
˜
a˜ 5 σ˜ Φ , i 5 1,2,…n (11)
i ( ij j
j51
Since the number of equations in Eq 11 is much smaller than the number of adjustments, there exists no unique solution to the
problem unless it is further restricted. The mathematical algorithmalgorithms in current adjustment codes are intended to make the
adjustments as small as possible relative to the uncertainties of the corresponding input data. Codes like STAY’SL, FERRET,
LEPRICON, and LSL-M2 (see Table 1) are based explicitly on the statistical principles such as “Maximum Likelihood Principle”
or “Bayes Theorem,” which are generalizations of the well-known least squares principle. Using principle, and are taking into
account variances and correlations of the input fluence, dosimetry, and cross section data (see 4.1.1, 4.2.2, and 4.3.3), even the ).
A detailed discussion of the mathematical derivations can be found in NUREG/CR-2222 and EPRI NP-2188. Even the older codes,
notably SAND-II and CRYSTAL BALL, can be interpreted as application of the least squares principle apply a minimization
algorithm although the statistical assumptions are not spelled out explicitly (seein Table 1). A detailed discussionthe supporting
documentation. Table 1 of the mathematical derivations can be found in NUREG/CR-2222 and EPRI NP-2188.lists some of the
available unfolding codes; however, the first four codes listed: SAND-II, SPECTRA, IUNFLD/UNFOLD, and WINDOWS have
severe limitations in that they do not typically provide uncertainty characterization of the resulting unfolded spectrum and the
adjusted damage exposure parameters.
3.3.1.1 An important problem in reactor surveillance is the determination of neutron fluence inside the pressure vessel wall at
locations which are not accessible to dosimetry. Estimates for exposure parameter values at these locations can be obtained from
adjustment codes which adjust fluences simultaneously at more than one location when the cross correlations between fluences at
different locations are given. LEPRICON has provisions for the estimation of cross correlations for fluences and simultaneous
adjustment. LSL-M2 also allows simultaneous adjustment, but cross correlations must be given.
3.3.2 The adjusted data ã , etc., are, for any specific algorithm, unique functions of the input variables. Thus, uncertainties
i
(variances and covariances) for the adjusted parameters can, in principle, be calculated by propagation the uncertainties for the
input data. Linearization may be used before calculating the uncertainties of the output data if the adjusted data are nonlinear
functions of the input data.
3.3.2.1 The algorithms of the adjustment codes tend to decrease the variances of the adjusted data compared to the
corresponding input values. The linear least squares adjustment codes yield estimates for the output data with minimum variances,
that is, the “best” unbiased estimates. This is the primary reason for using these adjustment procedures.
3.3.3 Properly designed adjustment methods provide means to detect inconsistencies in the input data which manifest
themselves through adjustments that are larger than the corresponding uncertainties or through large values of chi-square, or both.
(See NUREG/CR-3318 and NUREG/CR-3319.) Any detection of inconsistencies should be documented, and output data obtained
from inconsistent input should not be used. All input data should be carefully reviewed whenever inconsistencies are found, and
efforts should be made to resolve the inconsistencies as stated below.
3.3.3.1 Input data should be carefully investigated for evidence of gross errors or biases if large adjustments are required. Note
that the erroneous data may not be the ones that required the largest adjustment; thus, it is necessary to review all input data. Data
of dubious validity may be eliminated if proper corrections cannot be determined. Any elimination of data must be documented
and reasons stated which are independent of the adjustment procedure. Inconsistent data may also be omitted if they contribute
little to the output under investigation.
3.3.3.2 Inconsistencies may also be caused by input variances which are too small. The assignment of uncertainties to the input
data should, therefore, be reviewed to determine whether the assumed precision and bias for the experimental and calculational
data may be unrealistic. If so, variances may be increased, but reasons for doing so should be documented. Note that in statistically
based adjustment methods, listed in Table 1 the output uncertainties are determined only by the input uncertainties and are not
affected by inconsistencies in the input data (see NUREG/CR-2222). Note also that too large adjustments may yield unreliable data
because the limits of the linearization are exceeded even if these adjustments are consistent with the input uncertainties.
E944 − 19
TABLE 1 Available Neutron Spectrum Adjustment and Unfolding Codes
Code Available Refer-
Program Solution Method Comments
From ences
A
SAND-II semi-iterative RSICC Prog. No. CCC- 1 contains trial spectra library. No output uncertainties in the
112, CCC-619, PSR- original code, but modified Monte Carlo code provides output
345 uncertainties (2, 3, 4)
A
SAND-II semi-iterative RSICC Prog. No. CCC- (1, 2) contains trial spectra library. No output uncertainties in the
112, CCC-619, PSR- original code, but modified Monte Carlo code provides output
345 uncertainties (3)
SPECTRA statistical, linear estimation RSICC Prog. No. CCC- 5, 6 minimizes deviation in magnitude, no output uncertainties.
SPECTRA statistical, linear estimation RSICC Prog. No. CCC- (4, 5) minimizes deviation in magnitude, no output uncertainties.
IUNFLD/ statistical, linear estimation 7 constrained weighted linear least squares code using B-spline
UNFOLD basic functions. No output uncertainties.
IUNFLD/ statistical, linear estimation (6) constrained weighted linear least squares code using B-spline
UNFOLD basic functions. No output uncertainties.
WINDOWS statistical, linear estimation, linear RSICC Prog. No. PSR- 8 minimizes shape deviation, determines upper and lower bounds
programming 136, 161 for integral parameter and contribution of foils to bounds and
estimates. No statistical output uncertainty.
WINDOWS statistical, linear estimation, linear RSICC Prog. No. PSR- (7) minimizes shape deviation, determines upper and lower bounds
programming 136, 161 for integral parameter and contribution of foils to bounds and
estimates. No statistical output uncertainty.
RADAK, statistical, linear estimation RSICC Prog. No. PSR- 9, 10,11,12 RADAK is a general adjustment code not restricted to spectrum
SENSAK 122 adjustment.
RADAK, statistical, linear estimation RSICC Prog. No. PSR- (8, 9, 10, 11) RADAK is a general adjustment code not restricted to spectrum
SENSAK 122 adjustment.
STAY’SL statistical linear estimation RSICC Prog. No. PSR- 13 permits use of full or partial correlation uncertainty data for
113 activation and cross section data.
STAY’SL statistical linear estimation RSICC Prog. No. PSR- (12) permits use of full or partial correlation uncertainty data for
113 activation and cross section data.
NEUPAC(J1) statistical, linear estimation RSICC Prog. No. PSR- 14, 15 permits use of full covariance data and includes routine of
177 sensitivity analysis.
NEUPAC(J1) statistical, linear estimation RSICC Prog. No. PSR- (13, 14) permits use of full covariance data and includes routine of
177 sensitivity analysis.
FERRET statistical, least squares with log normal RSICC Prog. No. PSR- 2, 3 flexible input options allow the inclusion of both differential and
a priori distributions 145 integral measurements. Cross sections and multiple spectra may
be simultaneously adjusted. FERRET is a general adjustment
code not restricted to spectrum adjustments.
FERRET statistical, least squares with log normal RSICC Prog. No. PSR- (15, 16) flexible input options allow the inclusion of both differential and
a priori distributions 145 integral measurements. Cross sections and multiple spectra may
be simultaneously adjusted. FERRET is a general adjustment
code not restricted to spectrum adjustments.
LEPRICON statistical, generalized linear least RSICC Prog. No. PSR- 16, 17, 18 simultaneous adjustment of absolute spectra at up to two
squares with normal a priori and a 277 dosimetry locations and one pressure vessel location. Combines
posteriori distributions integral and differential data with built-in uncertainties. Provides
reduced adjusted pressure vessel group fluence covariances
using built-in sensitivity database.
LEPRICON statistical, generalized linear least RSICC Prog. No. PSR- (17, 18, 19) simultaneous adjustment of absolute spectra at up to two
squares with normal a priori and a 277 dosimetry locations and one pressure vessel location. Combines
posteriori distributions integral and differential data with built-in uncertainties. Provides
reduced adjusted pressure vessel group fluence covariances
using built-in sensitivity database.
LSL-M2 statistical, least squares, with log normal RSICC Prog. No. 19 simultaneous adjustment of several spectra. Provides
a priori and a posteriori distributions PSR-233 covariances for adjusted integral parameters. Dosimetry cross-
section file included.
LSL-M2 statistical, least squares, with log normal RSICC Prog. No. 20 simultaneous adjustment of several spectra. Provides
a priori and a posteriori distributions PSR-233 covariances for adjusted integral parameters. Dosimetry cross-
section file included.
UMG Statistical, maximum entropy with output RSICC Prog. No. 20, 21 Two components. MAXED is a maximum entropy code. GRAVEL
uncertatinties PSR-529 (22) is an iterative code.
UMG Statistical, maximum entropy with output RSICC Prog. No. (21, 22) Two components. MAXED is a maximum entropy code. GRAVEL
uncertainties PSR-529 (23) is an iterative code.
NMF-90 Statistical, least squares IAEA NDS 23, 24 Several components, STAY’NL, X333, and MIEKE. Distributed by
IAEA as part of the REAL-84 interlaboratory exercise on
spectrum adjustment (25).
E944 − 19
Code Available Refer-
Program Solution Method Comments
From ences
NMF-90 Statistical, least squares IAEA NDS (24, 25) Several components, STAY’NL, X333, and MIEKE. Distributed by
IAEA as part of the REAL-84 interlaboratory exercise on
spectrum adjustment (26).
GMA Statistical, general least squares RSICC Prog. No. 26 Simultaneous evaluation with differential and integral data,
PSR-367 primarily used for cross-section evaluation but extensible to
spectrum adjustments.
GMA Statistical, general least squares RSICC Prog. No. (27 ) Simultaneous evaluation with differential and integral data,
PSR-367 primarily used for cross-section evaluation but extensible to
spectrum adjustments.
A
The The boldface numbers in parentheses refer to the list of references appended to this guide.
3.3.4 Using the adjusted fluence spectrum, estimates of damage exposure parameter values can be calculated. These parameters
are weighted integrals over the neutron fluence
`
˜
p 5* Φ~E!w~E!dE (12)
o
or for group fluences
k
˜
p 5 Φ w (13)
( j j
j51
with given weight (response) functions w(E) or w , respectively. The response function for dpa of iron is listed in Practice E693.
j
Fluence greater than 1.0 MeV or fluence greater than 0.1 MeV is represented as w(E) = 1 for E above the limit and w(E) = 0 for
E below.
3.3.4.1 Finding best estimates of damage exposure parameters and their uncertainties is the primary objective in the use of
adjustment procedures for reactor surveillance. If calculated according to Eq 12 or Eq 13, unbiased minimum variance estimates
for the parameter p result, provided the adjusted fluence Φ ˜ is an unbiased minimum variance estimate. The variance of p can be
calculated in a straightforward manner from the variances and covariances of the adjusted fluence spectrum. Uncertainties of the
response functions, w , if any, should not be considered in the calculation of the output variances when a standard response
j
function, such as the dpa for iron in Practice E693, is used. The calculation of damage exposure parameters and their variances
should ideally be part of the adjustment code.
4. Selection of Input Data
4.1 Sensor Sets:
4.1.1 Radiometric Measurements (RM)—This is at present the primary source for dosimetry data in research and power reactors.
RM sensor selection, preparation, and measurement, including determination of variances and covariances, should be made
according to Guide E844 and the standards describing the handling of the particular foil material (Test Methods E262, E263, E264,
E265, E266, E393, E481, E523, E526, E704, and E705). Other passive dosimetry sensors of current interest in research and power
reactors and in ex-vessel environments are solid state track recorders (SSTR), helium accumulation fluence monitors (HAFM), and
damage monitors (DM). Use of these sensors is described in separate ASTM standards as follows:
4.1.2 SSTR—see Test Method E854.
4.1.3 HAFM—see Test Method E910.
4.1.4 The preceding list does not exclude the use of other integral measurements, for example, from fission chambers or nuclear
emulsions (see NUREG/CR-1861).
4.1.5 Accurate dosimetry measurements and proper selections of dosimetry sensors are particularly important if the
uncertainties in the calculated spectrum are large (see Ref (2728)). In this case, it is necessary either to have several dosimetry
sensors which respond to various parts of the neutron energy range of interest or to utilize a sensor which closely approximates
the energy response of the damage exposure parameters. Since determination of a variety of damage exposure parameters is
desirable, some combination of dosimeter responses is usually necessary to achieve the smallest possible output uncertainties.
Reactions currently used which are regarded as providing the best overlap with the iron dpa cross section are Np(n,f) and
93 93m 63 46 54 58
Nb(n,n') Nb. Other reactions used to measure neutrons above 1 MeV are Cu(n,α), Ti(n,p), Fe(n,p), Ni(n,p), and
U(n,f). (See Practice E853.) If the calculated spectrum has small uncertainties, the requirements of good spectral coverage or
good overlap with damage response are not as critical, but redundant dosimetry is still recommended to minimize chances of
erroneous results. (See Refs (2728, 2829.).))
235 239
4.1.6 Non-threshold sensors such as U(n,f), Pu(n,f), and all (n,γ) reactions are frequently u
...