ML003739334
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| Issue date: | 09/30/1993 |
| From: | Office of Nuclear Regulatory Research |
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Text
U.S. NUCLEAR REGULATORY COMMISSION September 1993 OFFICE OF NUCLEAR REGULATORY RESEARCH Division 1 Task DG-1025 DRAFT REGULATORY GUIDE
Contact:
A. Taboada (301) 492-3838
___________________I DRAFT REGULATORY GUIDE DG-1025 CALCULATIONAL AND DOSIMETRY METHODS FOR DETERMINING PRESSURE VESSEL NEUTRON FLUENCE This regulatory guide is being issued in draft form to involve the public in the early stages of the develop ment of a regulatory position in this area. It has not received complete staff review and does not represent an official NRC staff position.
associ Public comments are being solicited on the draft guide (including any implementation schedule) and its ated regulatory analysis or value/impact statement. Comments should be accompanied by appropriate supporting data. Written comments may be submitted to the Regulatory Publications Branch, OFIPS, Office of Administra tion, U.S. Nuclear Regulatory Commission, Washington, DC 20555. Copies of comments received may be examined at the NRC Public Document Room, 2120 L Street NW., Washington, DC. Comments will be most helpful if received by December 17, 1993.
Requests for single copies of draft guides (which may be reproduced) or for placement on an automatic distri
_*- bution list for single copies of future draft guides in specific divisions should be made in writing to the Distribution U.S. Nuclear Regulatory Commission, Washington, DC 20555, Attention: Office of Administration, and Mall Services Section,
TABLE OF CONTENTS Pagie A. INTRODUCTION ........ .................... . .. . . . 1 B. DISCUSSION .............................. 2 C. REGULATORY POSITION ............... 4
- 1. Neutron Fluence Calculational Methods . . . . . . . . . . . 4
- . . . . . . . 4 1.1 Input Data . . . . . . . . . . . . . .
- . . . 7 1.2 Core Neutron Source .......... . . . . . . .
1.3 Fluence Calculation ..........
1.4 Qualification .... ........... .. . . . . . . . .. . 14
- 2. Neutron Fluence Measurement Methods . 18 2.1 Measurement Procedures ..... 18 2.2 Validation in Standard and Reference Neutron Fields 21 2.3 Fluence Determination from Detector Measurements . 22 2.4 Sites for Updated Dosimetry .... ............. 23 2.5 Experimental Benchmarks for Validating Calculations 23
- 3. Reporting . . . . . . . . . . . . . . . . . . . . . . . . . . 24 3.1 Neutron Fluences and Uncertainties ........... 24 3.2 Specific Activities and Average Reaction Rates ..... 25 D. IMPLEMENTATION ............. ........................ 25 References .. . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 Regulatory Analysis ........... ...................... S. .. . R/A-1 iii
LIST OF TABLES Table Page
- 1. Summary of Calculation and Dosimetry Regulatory Positions ......... 27
- 2. Threshold Detectors Recommended for Pressure Vessel Dosimetry. . . . 31 iv
I A. INTRODUCTION 2 The U.S. Nuclear Regulatory Commission (NRC) has promulgated regulations 3 that ensure the structural integrity of the reactor pressure vessel for light 4 water-cooled power reactors. Specific fracture toughness requirements for 5 normal operation and for anticipated operational occurrences for power reac 6 tors are set forth in Appendix G, "Fracture Toughness Requirements," of 10 CFR 7 Part 50, !'Domestic Licensing of Production and Utilization Facilities." Addi 8 tionally, in response to concerns over potential pressurized thermal shock 9 (PTS) events in pressurized water reactors (PWRs), the NRC issued 10 CFR 10 50.61, "Fracture Toughness Requirements for Protection Against Pressurized 11 Thermal Shock Events."
12 To satisfy the requirements of both Appendix G and 10 CFR 50.61, methods 13 for determining the fast neutron fluence (E > 1 MeV) are necessary to estimate 14 the fracture toughness of the pressure vessel materials. Appendix H, "Reactor 15 Vessel Material Surveillance Program Requirements," of 10 CFR Part 50 requires 16 the installation of surveillance capsules, including material test specimens 17 and flux dosimeters, to provide data on material damage correlations as a 18 function of fluence.
19 The fracture toughness of pressure vessel materials is related to a 20 parameter called the material's "reference nil-ductility temperature," or 21 simply reference temperature, and is denoted as RTNDT. The RTNDT is defined 22 by a correlation of the fast neutron fluence (E > 1 MeV), material chemistry 23 (concentrations of Cu and Ni), initial reference temperature, and margin 24 to account for uncertainties in the correlation and input values. In 25 10 CFR 50.61, evaluation of the reference temperature based on the best 26 estimate of the fast neutron fluence at the end of the license period is 27 required, and the corresponding reference temperature is termed RTPTS.
28 This guide describes methods and assumptions acceptable to the NRC staff 29 for determining the pressure vessel neutron fluence. These methods are 30 directly applicable to the determination of RTNDT and RTPTs. Cases of 31 unusual plant characteristics or factors that require different methods and 32 assumptions will be considered on a plant-specific basis.
33 Compliance with this guide is not a regulatory requirement of the USNRC.
34 However, if a licensee elects to use the final version of this guide to 35 determine pressure vessel neutron fluence, implementation of the guide would 36 not be satisfied unless the licensee complies with certain specific provisions 1
I identified in the Regulatory Position of the guide. The use of the following 2 terms is explained to clarify compliance with these regulatory positions.
3 Must - Necessary provisions if implementation is to be satisfied.
4 Should - Provisions that are expected to be complied with unless it is 5 not possible because of specific circumstances (for example, 6 data needed to meet the position are not available).
7 May - Provisions that are acceptable and recommended, but are to be 8 applied at the option of the licensee.
9 This draft regulatory guide contains voluntary information collections 10 that are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et 11 seq.). This guide has been submitted to the office of Management and Budget 12 for review and approval of the paperwork requirements.
13 The public reporting burden for this collection of information over and 14 above the burden previously required for this activity is estimated to be an 15 average of 880 hours0.0102 days <br />0.244 hours <br />0.00146 weeks <br />3.3484e-4 months <br /> per respondent, including the time for reviewing instruc 16 tions, searching existing data sources, gathering and maintaining the data 17 needed, and completing and reviewing the collection of information. Send 18 comments regarding this burden estimate or any other aspect of this collection 19 of information, including suggestions for further reducing the reporting 20 burden, to the Information and Records Management Branch (MNBB-7741), U.S.
21 Nuclear Regulatory Commission, Washington DC 20555; and to the Desk Officer, 22 Office of Information and Regulatory Affairs, NEOB-3019, (3150-0011), Office 23 of Management and Budget, Washington, DC 20503.
24 B. DISCUSSION 25 The methods and assumptions described in this guide are for the calcula 26 tion and measurement of vessel fluence for core and vessel geometrical and 27 material configurations that are typical of current PWR and BWR power reactor 28 designs. This guide does not address the determination of surveillance speci 29 men material properties or the correlation between material properties and 30 neutron fluence. The methodology presented is intended as a best-estimate, 31 rather than a bounding or conservative, fluence determination. When required, 32 for example, in the RTPTs correlation called for in 10 CFR 50.61, an uncer 33 tairnty margin should be included separately. While the E > I MeV fluence has 2
1 been selected as the exposure parameter for the RTNDT and RTPTs correlations, 2 the procedures described in this guide determine the entire damage fluence 3 spectrum (from 0.1 to 15 MeV) and are generally applicable to other exposure 4 units, such as displacements per atom (dpa).
5 The determination of the pressure vessel fluence is based on both calcu 6 lations and measurements; the fluence prediction is made with a calculation 7 and the measurements are used to qualify the calculational methodology.
8 Because of the importance and the difficulty of these calculations, the 9 method's qualification by comparison to measurement must be made to ensure a 10 reliable and accurate vessel fluence determination. In this qualification, 11 calculation-to-measurement comparisons are used to identify biases in the cal 12 culations and to provide reliable estimates of the fluence uncertainties.'
13 When the measurement data are of sufficient quality and quantity, the com 14 parisons to measurement may be used to (1) determine the effect of the 15 various modeling approximations and any calculational bias and, if appropri 16 ate, (2) modify the calculations by application of a bias or by model adjust 17 ment or both. As an additional methods qualification, the sensitivity of the 18 calculation to the important input and modeling parameters must be determined 19 and used to provide an independent estimate of the overall calculational 20 uncertainty. The prediction of the vessel fluence must be made by an absolute 21 fluence calculation rather than a simple spatial extrapolation of the fluence 22 measurements.
23 The calculations of the pressure vessel fluence consist of the following 24 steps: (1) determination of the geometrical and material input data, (2) 25 determination of the core neutron source, (3) propagation of the neutron 26 fluence from the core to the vessel and into the cavity, and (4) qualification 27 of the calculational procedure. These steps are discussed in detail in 28 Regulatory Positions 1.1 through 1.4. In Regulatory Position 2, the use of 29 surveillance dosimetry as an in situ verification for the calculations is 30 described. Reporting is discussed in Regulatory Position 3. The major 31 regulatory positions are summarized in Table 1.
32 As an indication of current practice, selected codes and cross-section 33 libraries are listed in the references; however, it is the responsibility of 34 lin this guide, the term "uncertainty" refers to the random difference 35 between the estimated fluence (based on either calculation or measurement) and 36 the true fluence value. The term "bias" refers to the nonrandom or systematic 37 difference between the fluence estimate and the true value.
3
I I 1 the licensee to demonstrate their acceptability in a specific application.
2 Additional material related to the determination of pressure vessel fluence 3 and material damage, but considered outside the scope of this guide, is 4 contained in other regulatory guides and ASTM and ANSI/ANS Standards.
5 C. REGULATORY POSITION 6 1. NEUTRON FLUENCE CALCULATIONAL METHODS 7 The calculational methodology for estimating reactor vessel fluence 8 should agree with available (benchmark quality) measurements to within -20%
9 (1 sigma) 2 for the determination of RTPTs as described in 10 CFR 50.61. For 10 other applications, the accuracy should be determined using the approach 11 described in Regulatory Position 1.4 as an uncertainty allowance included in 12 the fluence estimate, as appropriate for the specific application.
13 1.1 Input Data 14 1.1.1 Materials and Geometry 15 Detailed material and geometrical input data should be used to define the 16 physical characteristics that determine the attenuation of the neutron flux 17 from the core to the locations of interest on the pressure vessel. These data 18 include descriptions of the pressure vessel, core, and internals; material 19 compositions; regional temperatures; and geometry. The geometrical input data 20 include the dimensions and locations of the fuel assemblies, reactor internals 21 (baffle, core support barrel, thermal shield, and neutron pads), the pressure 22 vessel (including identification and location of all welds and plates) and 23 cladding, and surveillance capsules. For cavity dosimetry, input data should 24 also include the width of the reactor cavity and the material compositions of 25 the support structure and concrete shielding, including water content, rebar, 26 and steel. The data should be based on documented and verified as-built 27 plant-specific dimensions and materials. The isotopic compositions of impor 28 tant constituent nuclides within each region should be based on measured data.
29 In the absence of plant-specific information, "generic" compositions and 30 2 It is recognized that a lower fluence calculational uncertainty may be 31 needed as the vessel approaches its projected end of life.
4
1 dimensions may be used; however, in this case conservative estimates of the 2 variations in the compositions and dimensions should be made and accounted for 3 in the determination of the fluence uncertainty (Regulatory Position 1.4.3).
4 The determination of the concentrations of the major isotopes responsible for 5 the fluence attenuation (e.g., iron and water) should be emphasized. The 6 water number densities should be based on plant full-power operating tem 7 peratures and pressures, as well as standard steam tables. The data should 8 account for axial and radial variations in water density caused by temperature 9 differences and the presence of in-channel and downcomer voids in the case of 10 BWRs.
11 1.1.2 Cross-Sections 12 The calculational method to estimate vessel damage fluence should use the 13 neutron cross-sections over the energy range from -0.1 MeV to -15 MeV and 14 apply the latest version of the Evaluated Nuclear Data File (ENDF/B-VI).
15 These data have been thoroughly reviewed and tested relative to experimental 16 benchmarks. 3 Cross-section sets based on earlier or equivalent nuclear data 17 sets that have been thoroughly benchmarked for a specific application may be 18 used for that specific application. However, when the reevaluated cross 19 section data change, the effect of these changes on the licensee-specific 20 methodology must be evaluated and the fluence estimates updated as 21 appropriate.
22 Since the discrete ordinates transport codes that are used to determine 23 the neutron fluence employ a multigroup approximation, the basic data con 24 tained within the ENDF files must be preprocessed into a multigroup structure.
25 The development of a multigroup library should consider the adequacy of the 26 group structure, the energy dependence of the flux used to average the cross 27 sections over the individual groups, and the order of the angular expansion of 28 the scattering cross-section. Sufficient details of the energy and angular 29 dependence of the differential cross-sections (e.g., the minima in the iron 30 3 1t should be noted that in many applications the ENDF/B-IV and the first 31 three MODs of the ENDF/B-V iron cross-sections result in -20% underprediction 32 of the vessel inner-wall fluence and -35% underprediction of the cavity flu 33 ence (Refs. 1-3). Updated ENDF/B-V iron cross-section data (Ref. 4) have been 34 demonstrated to provide a more accurate determination of the flux attenuation 35 through iron (Refs. 2, 3) and are strongly recommended. These new iron data 36 are included in ENDF/B version VI.
5
1 total cross-section) should be included to preserve the prescribed accuracy in 2 attenuation characteristics.
3 The construction of the multigroup library involves the selection of a 4 problem-independent multigroup "master" library containing data for all 5 required isotopes. This library should include a sufficiently large number of 6 groups (-100-200) so that differences between the shape of the assumed flux 7 spectrum and the true flux have a negligible effect on the multigroup data.
8 This library typically includes 100 energy groups above -0.1 MeV. In 9 addition, a minimum of a P-3 Legendre expansion of the scattering cross 10 section should be used for typical LWR configurations. Several libraries 11 satisfying these provisions are available from RSIC -- the Oak Ridge National 12 Laboratory Radiation Shielding Information Center (Refs. 5, 6, 7).
13 The number of groups may be reduced, with little loss in accuracy, by 14 collapsing the data in the "master" library over spectra that more closely 15 approximate the true spectra. This reduction may be accomplished with a one 16 dimensional calculation that includes the discrete regions of the core, vessel 17 internals, by-pass and downcomer water, pressure vessel, reactor cavity, 18 shield, and support structures. The resulting "job" library should consist of 19 macroscopic multigroup (<50) cross-sections based on the region-specific iso 20 topic compositions. This library should include -20 energy groups above -0.1 21 MeV. It is the responsibility of the licensee to demonstrate the adequacy of 22 the "job" library, and this may be accomplished by comparing calculations for 23 a representative configuration performed with both the "master" and "job" 24 libraries. In these comparisons, the threshold detector cross-sections and 25 reaction rates obtained with the fine group structure should be preserved in 26 the multigroup calculations.
27 There are several group libraries available from RSIC that were 28 generated using light-water reactor (LWR) spectra for the group collapsing 29 from "master" libraries (Refs. 8, 9) and may be used for LWR application.
30 These libraries contain microscopic as well as some premixed macroscopic 31 cross-sections for relevant isotopes and materials. Because these prepackaged 32 libraries were designed specifically for LWR pressure vessel fluence calcula 33 tions, their applicability in any atypical application should be verified 34 prior to use.
6
1 1.2 Core Neutron Source 2 The determination of the fixed neutron source for the pressure vessel 3 fluence calculations should entail specification of the temporal, spatial, and 4 energy dependence together with the absolute source normalization.
5 The spatial dependence of the source should be from core-follow calcula 6 tions or from measured data. Core-follow calculations should be performed 7 with a three-dimensional (3-D) coarse mesh simulator code and provide the 8 relative power over large rectangular nodes (typically -20 x 20 x 30 cm in the 9 x, y, and z dimensions, respectively). Plant process computers provide 10 similar power distribution data obtained from in-core instrumentation.
11 The core neutron source should be determined by the power distribution, 12 which varies significantly with fuel burnup, power level, and the fuel man 13 agement scheme. The detailed state-point dependence should be accounted for 14 (Refs. 10, 11); however, if this is not feasible, an approximate averaging 15 over the operating power distribution is acceptable and may be obtained by (1) 16 averaging representative power distributions within the cycle or (2) assuming 17 the cycle-average assembly power distribution is well approximated by the 18 accumulated exposure distribution at the end of the cycle.
19 A best-estimate power distribution may be used for reactor vessel neutron 20 damage fluence calculations. However, this best-estimate power distribution 21 must be updated if changes in core loadings, surveillance measurements, or 22 other information indicate a significant change in projected fluence values.
23 This updating may be avoided by using a conservative "generic" average power 24 distribution, provided no measured distribution yields higher power levels for 25 the important peripheral assemblies.
26 The peripheral assemblies, which contribute the most to the vessel 27 fluence, have strong radial power (source) gradients, and these gradients 28 should not be neglected since this will result in an overprediction of the 29 fluence if the average assembly power is used (Ref. 12). The pin-wise source 30 distribution should be used for best-estimate calculations and the peripheral 31 assembly pin-wise source data should be obtained from quarter-core fine mesh 32 calculations in which each fuel-pin cell is explicitly represented. The pin 33 wise power distribution and the nodal power should be combined to define the 34 absolute power distribution in the assembly.
35 The energy dependence of the source (i.e., the spectrum) and the 36 normalization of the source to the number of neutrons per megawatt must 7
1 account for the fact that changes in the isotopic fission fractions with fuel 2 exposure result in variations in the energy-dependent fission spectra, the 3 number of neutrons produced per fission, and the energy released per fission.
4 These effects tend to increase the fast neutron source per megawatt of power 5 for high-burnup assemblies. The variations in these physics parameters with 6 fuel exposure may be obtained from standard lattice physics depletion calcula 7 tions (Refs. 13, 14). This effect should also be accounted for in plants that 8 have adopted the PWR low-leakage refueling schemes (vs. out-in-in three-batch 9 fuel management) in which once, twice, or thrice burned fuel is located in the 10 high importance peripheral locations (Refs. 15-18). The harder spectrum in 11 the BWR fuel regions having a high void fraction will have a similar effect on 12 the isotopic fission fractions and on the neutron source normalization and 13 spectrum.
14 The horizontal core geometry may be described using an (r,O) representa 15 tion of the nominal plane. A planar octant representation is acceptable for 16 the octant symmetric fuel-loading patterns typically employed in LWRs. Fuel 17 loading patterns that are not octant symmetric may be represented in octant 18 geometry using the octant having the highest fluence. To accurately represent 19 the important peripheral assembly geometry, a 0-mesh of 40-80 angular inter 20 vals must be applied. The (r,O) representation should reproduce the true 21 physical assembly area to within -0.5% and the pin-wise source gradients to 22 within -10%. The assignment of the (x,y) pin-wise powers to the individual 23 (r,6) mesh intervals should be made on a fractional area or equivalent basis 24 (Refs. 19, 20). Reference 20 is particularly useful if the radial mesh is a 25 function of 0).
26 The overall source normalization should be performed with respect to the 27 (r,O) source so that differences between the core area in the (r,0) 28 representation and the true core area do not bias the fluence predictions.
29 Determination of the 3-D fluence at the vessel using (r,z) and (r) 30 geometry calculations may also be appropriate, (see Regulatory Position 31 1.3.2). If these calculations are used to provide an axial correction factor, 32 the source specification may be less stringent if consistent sources are used.
8
1 1.3 Fluence Calculation 4
2 1.3.1 Transport Calculation 3 The transport of neutrons from the core to locations of interest in the 4 pressure vessel should be determined with a two-dimensional discrete ordi 5 nates 5 transport program (Refs. 22-24) in (r,O) and, when appropriate, in 6 (r,z) and (r) geometries. 6 When calculating a horizontal plane of the core/
7 vessel geometry in which the rectangular (x,y) geometry of the core boundary 8 and the cylindrical (r) geometry of the vessel are mixed, a more accurate 9 description is provided by the variable (r,O) mesh option (Ref. 23) and may be 10 applied.
11 The azimuthal (0) mesh using 40-80 intervals over an octant in (r,e) 12 geometry in the horizontal plane must provide an accurate representation of 13 the spatial distribution of the material compositions and source described in 14 Regulatory Position 1.2. The radial mesh in the core region should be -2 15 intervals per inch for peripheral assemblies, and may be much coarser for 16 assemblies more than approximately two assembly pitches removed from the core 17 reflector interface. In excore regions, a spatial mesh that ensures the flux 18 in any group changes by less than a factor of -2 between adjacent intervals 19 should be applied, and radial mesh of at least -3 intervals per inch in water 20 and -1.5 intervals per inch in steel should be used. Because of the rela 21 tively weak axial variation of the fluence, a coarse axial mesh of -0.5 inter 22 val per inch may be used except near material and source interfaces, where 23 flux gradients can be large. An S8 fully symmetric angular quadrature should 24 be used as a minimum for determining the fluence at the vessel. However, in 25 reactor cavity fluence calculations a higher order quadrature may be needed, 26 4Additional information concerning the application of transport methods 27 to reactor vessel surveillance is provided in ASTM Standard E482-89 (Ref. 21).
28 5The discrete ordinates transport method is generally used for the 29 calculation of pressure vessel fluence. It is recognized, however, that 30 alternative methods such as Monte Carlo are available, and analyses performed 31 with these methods will be reviewed on a case-by-case basis.
32 6 If DOT (Ref. 23) is used, the "0-weighted" option (MODE-5 in DOT 4.3) is 33 considered to be more accurate than the "weighted" option (MODE-3 in DOT 3.5 34 or 4.3) for flux extrapolation and is recommended. Also, the default 0 = 0.9 35 value is adequate.
9
1 depending on the width of the cavity and the axial location at which the 2 fluence is being calculated.
3 Where computer storage limitations prevent the implementation of these 4 mesh densities, the calculation should be performed in two or more "bootstrap" 5 steps rather than compromising the spatial mesh or quadrature (the number of 6 groups used usually does not affect the storage limitations, only the execu 7 tion time). In this approach, the problem volume is divided into over-lapping 8 regions. In a two-step bootstrap calculation, for example, a transport cal 9 culation is performed for the cylinder defined by 0 < r < R' with a fictitious 10 vacuum boundary condition applied at R'. From this initial calculation a 11 boundary source is determined at the radius R" = R' - A and is applied as the 12 left-hand boundary condition for a second transport calculation from R" to R 13 (the true outer boundary of the problem). The adequacy of the overlap region 14 must be tested (e.g., by decreasing the inner radius of the outer region) to 15 ensure that the use of the fictitious boundary condition at R' has not unduly 16 affected the boundary source at R" or the results at the vessel.
17 A point-wise flux convergence criterion of Z 0.001 should be used, and a 18 sufficient number of iterations should be allowed within each group to ensure 19 convergence. To avoid negative fluxes and improve convergence, a weighted 20 difference model should be used. 6 The adequacy of the spatial mesh and 21 angular quadrature, as well as the convergence, must be demonstrated by 22 tightening the numerics until the resulting changes are negligible. In dis 23 crete ordinates codes, the spatial mesh and the angular quadrature should be 24 refined simultaneously. In many cases, these evaluations can be adequately 25 performed with a one-dimensional model.
26 In performing calculations of surveillance capsule fluence (Regulatory 27 Position 1.4), it should be noted that the capsule fluence is extremely sen 28 sitive to the geometrical representation of the capsule geometry and internal 29 water region (if present), and the adequacy of the capsule representation and 30 mesh must be demonstrated using sensitivity calculations. In addition, the 31 capsule fluence and spectra are sensitive to the radial location of the cap 32 sule and its proximity to material interfaces (e.g., at the vessel, thermal 33 shield, and concrete shield in the cavity), and these should be represented 34 accurately. To account for the neutron spectrum dependence of RTNoT when the 35 fluence is extrapolated from the surveillance capsule or from the inside of 36 the pressure vessel to the T/4 and 3T/4 vessel locations, a spectral lead 37 factor (which accounts for the change in neutron spectrum between downcomer 10
1 and vessel internal locations) should be applied for the calculation of ARTNDT 2 (Ref. 25).
3 The transport calculations may be performed in either the forward or 4 adjoint modes. When several transport calculations are needed for a specific 5 geometry, assembly importance factors may be precalculated by either perform 6 ing calculations with a unit source specified in the assembly of interest or 7 by performing adjoint calculations. The adjoint fluxes determine the fluence 8 at a specific (field) location, while the forward fluxes from the unit source 9 calculations determine the fluence at all locations in the problem. Once cal 10 culated, these factors contain the required information from the transport 11 solution, and by weighing the assembly importance factors with the source dis 12 tribution of interest, the vessel (or capsule) fluence may be determined with 13 out additional transport calculations, assuming the in-vessel geometry and 14 material remain the same.
15 When fluence reduction schemes have introduced strong axial or azimuthal 16 heterogeneities into the source (e.g., an axially zoned replacement of fuel 17 with stainless steel for fluence reduction), a finer spatial mesh and tighter 18 convergence criteria may be appropriate to ensure an accurate solution. These 19 schemes may also entail a 3-D flux calculation (Regulatory Position 1.3.2).
20 1.3.2 Synthesis of the 3-D Fluence 21 When 3-D calculations are not performed, a 3-D fluence representation may 22 be constructed by synthesizing calculations of lower dimensions using the 23 expression 0g(r,6,z) = Og(r,O)*Lg(r,z) (Equation 1) 24 where Og(r,O) is the group-g transport solution in (r,6) geometry for a 25 representative plane and Lg(r,z) is a group-dependent axial shape factor. Two 26 simple methods available for determining Lg(r,z) are defined by the 27 expressions Lg(r,z) = P(z) (Equation 2) 28 where P(z) is the peripheral assembly axial power distribution, and Lg(r,z) = og(r,z)/og(r) (Equation 3) 11
1 where Og(r) and Og(r,z) are one- and two-dimensional group-g flux solutions, 2 respectively, for a cylindrical representation of the geometry that preserves 3 the important axial source and attenuation characteristics (Ref. 26). The 4 (r,z) plane should correspond to the azimuthal location of interest (e.g.,
5 peak vessel fluence or dosimetry locations) or a conservatively 0-averaged 6 (r,z) plane. The source per unit height for both the (r,O) and (r) models 7 should be identical, and the true axial source density should be used in the 8 (r,z) model.
9 Equation 2 is applicable when (a) the axial source distribution for all 10 important peripheral assemblies is approximately the same or is bounded by a 11 conservative axial power shape and (b) the attenuation characteristics do not 12 vary axially over the region of interest. In addition, since the axial 13 fluence distribution tends to flatten as it propagates from the core through 14 the pressure vessel, this approximation will tend to overpredict axial fluence 15 maxima and underpredict minima. This nonconservative underprediction could be 16 large near the top and bottom reflectors, as well as if minima are strongly 17 localized as occurs in some fluence reduction schemes.
18 Equation 3 is applicable when the axial source distribution and attenu 19 ation characteristics vary radially but do not vary significantly in the 20 azimuthal (0) direction within a given radial annulus. For example, this 21 approximation is not appropriate when strong axial fuel enrichment variations 22 are present only in selected peripheral assemblies.
23 In summary, an (r,O) geometry fluence calculation and a knowledge of the 24 peripheral assembly axial power distribution are needed when using Equation 2.
25 It will result in fluence overpredictions near the midplane at relatively 26 large distances from the core, e.g.,in the cavity, and underpredictions at 27 axial locations beyond the beltline at relatively large radial distances from 28 the core. Conservatism may be included in the latter case by using the peak 29 axial power for all elevations.
30 Both radial and axial fluence calculations, are needed when using 31 Equation 3; thus, it is generally more accurate in preserving the integral 32 properties of the three-dimensional fluence. Both Equation 2 and Equation 3 33 assume separability between the axial and azimuthal fluence calculations, 34 which is generally only approximately true.
35 When these simple synthesis techniques are not applicable, multichannel 36 synthesis methods may be used. In the multichannel synthesis calculation, the 37 fluence is represented as the sum 12
N 0g(r,O,z) = ajigj(r,O)Ogj(r,z)/Ogj(r) (Equation 4)
- i=1 I where the Ogi are basis flux solutions, typically representing specific 2 regions of the core/vessel geometry, and the weighing coefficients a, are 3 determined to provide an optimum prediction of the vessel fluence. It should 4 be emphasized, however, that the accuracy of this method is sensitive to the 5 selection of the basis functions, especially at region interfaces; and three 6 dimensional calculations should be considered where strong axial or azimuthal 7 heterogeneities exist. This synthesis technique has been applied to an 8 experimental benchmark in Reference 27.
9 1.3.3 Cavity Fluence Calculations 10 Accurate cavity fluence calculations are relied on to analyze dosimetry 11 located in the reactor cavity and to determine the fluence accumulated by the 12 reactor support structures (Ref. 28). The calculation of the neutron trans 13 port in the cavity is made difficult by (a) the strong attenuation of the 14 E > I MeV fluence through the vessel and the resulting increased sensitivity 15 to the iron inelastic cross-section and (b) the possibility of neutron stream 16 ing (i.e., strong directionally dependent) effects in the cavity. Because of 17 this increased sensitivity to the iron cross-sections, the most accurate ENDF/
18 B-VI cross-section data must be used for cavity fluence calculations. 3 Typi 19 cally, the width of the cavity together with the close-to-beltline locations 20 of the dosimetry capsules result in minimal cavity streaming effects, and an 21 S8 angular quadrature is acceptable. However, when off-beltline locations or 22 narrow cavities are analyzed, the adequacy of the S8 quadrature must be demon 23 strated with higher-order Sn calculations. In addition, since the radial mesh 24 in the (r,z) calculation is generally finer than the z-mesh in the cavity 25 resulting in narrow spatial mesh intervals, a 0-weighted difference model 26 should be used. 6 27 The cavity fluence is sensitive to both the material compositions 28 (Regulatory Position 1.1.1) and the local geometry (e.g., the presence of 29 detector wells) of the concrete shield, and these should be represented as 30 accurately as possible. Benchmark measurements involving simulated reactor 31 cavities that are recommended for methods evaluation are described in Refer 32 ence 3. The measured energy spectrum for a typical PWR cavity is given in 33 Reference 29. When both in-vessel and cavity dosimetry measurements are 13
I available, an additional verification of the measurements and calculations may 2 be made by comparing the vessel inner-wall fluence determined from the extra 3 polation of the two measurements. Measurements performed in reactor cavities 4 are described in References I and 2.
5 1.4 Qualification 6 The neutron transport calculation must be qualified, and fluence uncer 7 tainty estimates must be determined. The neutron fluence undergoes several 8 decades of attenuation before reaching the vessel, and the calculation is 9 sensitive to the material and geometrical representation of the core and 10 vessel internals, the neutron source, and the numerical schemes used in its 11 determination. The uncertainty estimates are used to determine the appro 12 priate uncertainty allowance to be included in the application of the fluence 13 estimate. While adherence to the guidelines described here will generally 14 result in accurate fluence estimates, the overall methodology must be quali 15 fied in order to quantify uncertainties, identify any potential biases in the 16 calculations, and provide confidence in the fluence calculations. In addi 17 tion, while the methodology, including computer codes and data libraries used 18 in the calculations, may have been found to be acceptable in previous applica 19 tions, the qualification ensures that the licensee's implementation of the 20 methodology is valid.
21 The methods qualification consists of three parts: (1) the analytic 22 uncertainty analysis, (2) the comparison with benchmarks and plant-specific 23 data, and (3) the estimate of uncertainty in calculated fluence.
24 1.4.1 Calculational Uncertainty Analysis 25 To demonstrate the accuracy of the methodology, an analytic uncertainty 26 analysis must be performed. This analysis includes identification of the 27 important sources of uncertainty. For typical fluence calculations, these 28 sources include:
29 o Nuclear data (cross-sections and fission spectrum) 30 0 Geometry (locations of components and deviations from the nominal 31 dimensions) 14
1 o Isotopic composition of material (density and composition of coolant 2 water, core barrel, thermal shield, pressure vessel with cladding, 3 and concrete shield) 4 o Neutron sources (core leakage, space and energy distribution, and 5 burnup dependence) 6 o Methods error (mesh density, angular expansion, convergence 7 criteria, macroscopic group cross-sections, fluence perturbation by 8 surveillance capsules, spatial synthesis, and cavity streaming) 9 This list is not necessarily exhaustive, and other uncertainties that are 10 specific to a particular reactor or a particular calculational method should 11 be considered. In typical applications, the fluence uncertainty is dominated 12 by a few uncertainty components, such as the geometry, which are usually 13 easily identified and substantially simplify the uncertainty analysis.
14 The effect of the significant component uncertainties should be deter 15 mined by a series of sensitivity calculations in which the calculational model 16 input data and modeling assumptions are varied and the numerical effect on the 17 calculated fluence is determined. (A typical sensitivity would be 15%
18 decrease in vessel >1 MeV fluence per centimeter increase in vessel inside 19 radius.) Estimates of the expected uncertainties in these input parameters 20 must be made and combined with the corresponding fluence sensitivities to 21 determine the expected total fluence uncertainty (i.e., standard deviation).
22 The independent random uncertainties should be combined in a statistical or 23 root-mean-square fashion, and the systematic errors (or biases) should be 24 combined algebraically, recognizing the sign of each contribution. The com 25 ponent uncertainties should be based on measurement or on the acceptable 26 deviations included in the design specifications. The sensitivity calcula 27 tions may be performed in one dimension when the model sensitivity does not 28 involve a detailed two-dimensional representation.
29 A sensitivity analysis, in which the influence of each of these uncer 30 tainties on the calculated group fluences has been considered, is included in 31 References 1 and 2 for several power reactor benchmarks. Since the uncertain 32 ties used in these analyses are common to many pressurized water reactors, the 33 uncertainties (including correlations) may be used as initial uncertainty 34 estimates. These variance estimates should be modified as additional 15
1 experience is obtained or if the reactor of interest differs substantially 2 from the benchmark reactor. The referenced benchmark sensitivity analysis 3 provides guidance for such modifications.
4 1.4.2 Comparisons with Benchmark and Plant-Specific Data 5 Calculational methods must be validated by comparison with measurement 6 and calculational benchmarks. The fluence calculational methods should be 7 validated against (1) a power reactor benchmark that provides in-vessel sur 8 veillance capsule dosimetry or ex-vessel cavity measurements or both, (2) a 9 pressure vessel simulator benchmark that provides measurements at the inner 10 surface and at the T/4 and 3T/4 positions within the vessel (see Regulatory 11 Position 2.5 for a discussion of such benchmarks), and (3) available fluence 12 calculational benchmarks. The results of the validation should include com 13 parisons of reaction rates, fluences, and group fluxes for the locations of 14 interest (Refs. 30, 31). Any adjustments made to the calculations should be 15 justified and reported.
16 1.4.2.1 Surveillance Capsule Measurements. The vessel surveillance 17 capsules provide the most important source of fluence measurement data, and 18 calculations should be performed for capsule measurements of the specific 19 reactor or reactors of similar design. These measurements have the advantage 20 of including the as-built materials and geometry and the actual operating 21 conditions. In calculating the capsule dosimeter reaction rates, the latest 22 dosimeter cross-sections and a detailed modeling of the capsule materials and 23 geometry should be employed.
24 1.4.2.2 Operating Reactor Fluence Measurements. Several fluence 25 measurements for operating reactors may be used for methods and data quali 26 fication (examples are provided in Reference 32). These measurements include 27 the three-dimensional geometry and operating conditions, and in some cases, 28 include both in-vessel and ex-vessel measurements. An especially accurate 29 determination of the fluence attenuation through the vessel (e.g., to the T/4 30 and 3T/4 locations) can be obtained when both in-vessel and cavity dosimetry 31 are available.
32 1.4.2.3 Calculational Benchmarks. The two-dimensional vessel fluence 33 standard problem provided by the NRC (Ref. 33) and the one-dimensional 16
Nuclear Society I shielding calculational benchmarks proposed by the American In these 2 Benchmark Committee (Ref. 34) may be used for methods qualification.
source are 3 benchmarks, the geometry, materials, space, and energy-dependent problems pro 4 fixed by the problem specification. The calculation of these of the trans 5 vides a detailed test of the cross-sections and various aspects convergence 6 port calculations, such as the spatial mesh, quadraturi, and 7 criteria.
data must 8 The methods used to calculate the benchmarks and plant-specific fluence 9 be consistent (to the extent possible) with those used in the vessel and 10 calculations. That is, the same cross-sections, transport techniques, appli 11 transport code parameters that are to be used in the reactor licensing and 12 cation must be employed in the calculation of the benchmark measurements 13 reference calculations.
14 Differences between measurements and calculations should be consistent calculations.
15 with the combined uncertainty estimates for the measurements and in 16 The calculated and measured reaction rates (using the methods described measurements to 17 Regulatory Positions 1.1 through 1.3) typically agree with the dosimetry.
18 within about 20% for surveillance capsules and 30% for cavity and the 19 Deviations outside these uncertainty limits must be investigated is 20 calculations or measurements modified when the cause of the deviation 21 identified (Refs. 35-42).
22 The comparisons of the calculations to the benchmarks and plant-specific 23 data should be used to estimate the calculational bias and uncertainties.
24 When a bias is applied to the calculation to determine a best-estimate 25 fluence, the justification and basis for the bias must be identified.
26 1.4.3 Estimate of Fluence Calculational Uncertainty an 27 The overall fluence calculation uncertainty must be determined by (Regulatory 28 appropriate combination of (1) the analytic uncertainty analysis comparisons to 29 Position 1.4.1) and (2) the uncertainty estimate based on the requirements are 30 benchmarks (Regulatory Position 1.4.2). The fluence accuracy of 20%
31 generally application specific; however, a vessel fluence uncertainty if the 32 (1 sigma) is acceptable for RTPTs determination. For PTS applications, the calcula 33 benchmark comparisons indicate an uncertainty greater than -20%,
within 34 tional model must be adjusted or a bias applied to bring the agreement calculated 35 this range. The uncertainty determined may be combined with the on the vessel 36 best-estimate fluence to provide a high-probability upper bound 17
_1_ __J 1 fluence. For example, assuming the uncertainty is distributed normally, the 2 95% probability upper tolerance fluence limit is the best-estimate value plus 3 1.65 sigma.
4 The fluence calculational methods of this section are summarized in 5 Section A of Table 1, Summary of Regulatory Positions on Calculation and 6 Dosimetry.
7 2. NEUTRON FLUENCE MEASUREMENT METHODS 8 Dosimetry measurements provide an independent estimate of the neutron 9 fluence to confirm neutron transport calculations. This fluence is obtained 10 from the response of passive integral detectors placed in surveillance cap 11 sules and, more recently, in the ex-vessel cavity. Procedures for performing 12 these measurements to obtain a complete analysis, a reliable uncertainty 13 assessment, and proper documentation are described in this section. Standard 14 neutron field validation and sites for placing updated dosimetry are also 15 described.
16 The fluence measurement provisions of this section are summarized in 17 Section B of Table 1, Summary of Regulatory Positions on Calculation and 18 Dosimetry.
19 2.1 Measurement Procedures 20 Measurement methods in power reactor dosimetry include passive integral 21 detectors, which are typically activation detectors, and solid state track 22 recorders. The most frequently used detectors respond to neutrons with ener 23 gies above a characteristic reaction threshold. These detectors should be 24 selected with substantial nonoverlapping energy regions (i.e., with well 25 separated thresholds) to provide coarse spectrum information as well as an 26 estimate of the neutron fluence.
27 2.1.1 Specification and Application of Dosimeters 28 Neutron dosimetry for pressure vessel surveillance may consist of as 29 built packages of threshold dosimeters placed in surveillance capsules during 30 reactor construction. The selected dosimeter set must provide adequate spec 31 trum coverage. A common set of fast neutron integral detectors that may 32 be employed in these packages is listed in Table 2. Taken together with a 18
I low-energy detector such as cobalt (to measure the thermal neutron fluence and 2 determine interference from low-energy activations), the set provides satis 3 factory neutron energy spectrum coverage for pressure vessel dosimetry.
4 Alternative detector sets that are used should provide equivalent spectrum 5 coverage. Detector selection criteria and related recommendations in ASTM 6 E844 (Ref. 43) and E1005 (Ref. 44) should be followed (see Table 2).
7 Application of activation detectors involves measurement elements that 8 must be carefully controlled and documented to establish reasonable uncer 9 tainty estimates. Where applicable, procedures in ASTM Standards E181 (Ref.
10 45) and E1005 (Ref. 44) and methods devoted to individual radiometric sensors 11 must be used as indicated in Table 2. Specific regulatory positions asso 12 ciated with the dosimetry measurements are indicated in the following.
13 2.1.1.1 Isotopic Composition. The dosimeter materials should be pure 14 enough to ensure there is no significant error in the response of the dosi 15 meter from extraneous activities. Cursory specifications of materials regard 16 ing impurities are often unreliable. Specifically, fissile residuals in Np 17 237 and U-238 and minute amounts of cobalt in copper (fractional parts per 18 million) should be determined by mass spectrography or radioactivation 19 analysis.
20 2.1.1.2 Encapsulation. The detector capsule's designs must taken into 21 account possible activation interference and neutron spectrum perturbation.
22 Thermal neutron shields that eliminate interference from thermal neutron 23 reactions in some detectors must be designed to accommodate radiation heating 24 and should be placed apart from low-energy detectors (see ASTM Standard E844, 25 Ref. 43).
26 2.1.1.3 Isotopic Mass. Stoichiometry and isotopic analysis should be 27 well documented for dosimeters that are not of pure natural elements.
28 2.1.1.4 Location. The location of individual dosimeters must be 29 determined accurately and recorded, because fluence gradients in out-of-core 30 positions are generally severe. Also, the surroundings of a dosimeter (e.g.,
31 adjacent dosimeters or material interface) can influence detector response.
32 In the pressure vessel cavity, establishing azimuthal position can be as 33 important as the radial location. Specially designed mounting arrangements, 19
I including vertical gradient wires, should be used for cavity dosimetry.
2 Comprehensive and accurate detector location information should be maintained.
3 2.1.1.5 Solid-State Track Recorders. In addition to activation detec 4 tors, integral detectors employing fission reactions make use of Solid State 5 Track Recorders (SSTRs). These sensors directly record fission fragments from 6 a thin fissionable deposit (Ref. 52). Advantages of these detectors are wide 7 sensitivity ranges, a permanent measurement record, and convenient application 8 of fission reaction dosimetry in remote and hostile environments. Because the 9 application is new and employs fissionable deposits with masses in the nano 10 gram to picogram range, details of the measurements should be well documented, 11 and standard neutron field calibration prior to application should be per 12 formed. ASTM Standard E854 (Ref. 53) provides additional information 13 concerning the use of SSTRs and may be applied.
14 2.1.2 Detector Response Measurements 15 In order to allow comparison of the dosimetry measurements with the 16 neutron transport calculation, two detector response parameters must be 17 reported: the measured reaction probability (disintegrations per nucleus) and 18 the measured average reaction rate (disintegrations per second, per nucleus).
19 Corrections should be included for time-history, detector response perturba 20 tions, interfering reactions, and, when applicable, burnup and photofission.
21 In order to allow an accurate treatment of the time-history effects, the half 22 lives of the dosimeter activation products should be considered (Ref. 44).
23 Photofission corrections can vary considerably (from 2 to 15%) depending upon 24 the location of the surveillance capsule and type of reactor. Fission yields 25 should be those specified in the relevant ASTM standards. In situ neutron 26 field perturbations (e.g., by the surveillance capsule and detector encapsula 27 tion) must be accounted for if they are not an integral part of the neutron 28 transport calculation.
29 When reporting measurements, these corrections should be described along 30 with any other effects that have a significant impact on the measurements.
31 This is especially important because pressure vessel surveillance dosimetry 32 often involves comparison of measurements carried out by different 33 organizations over long periods of time.
34 The "specific activity at end-of-irradiation" (given in disintegra 35 tions per second, per nucleus, and divided by the fission yield where 20
I appropriate) is an additional detector response parameter that should be 2 reported. This directly measured quantity has the advantage of not involving 3 irradiation-related corrections that are somewhat uncertain and are often 4 subject to re-evaluation.
5 2.1.3 Uncertainty Estimates 6 There must be at least one uncertainty table that states the component 7 uncertainties contributing to each detector response. The entries in the 8 table should be stated as standard deviations, upper bounds, or appropriate 9 fractions of the correction. Each entry should be described, along with the 10 method by which the entries are combined to obtain a total uncertainty. This 11 accounting of uncertainties provides a reliable estimate of the measurement 12 accuracy and helps establish the extent to which the analysis of the 13 measurement is complete.
14 2.2 Validation in Standard and Reference Neutron Fields 15 To ensure long-term measurement consistency and confirm measurement 16 uncertainties, dosimetry measurements must be performed periodically in one or 17 more well-characterized neutron fields. The Materials Dosimetry Reference 18 Facility (MDRF) and fission neutron sources may be used for this purpose 19 (Ref. 54). Neutron field referencing may be used as a detector response 20 calibration.
21 The validation is accomplished by exposing each type of detector to a 22 certified neutron fluence in the reference neutron field and by determining 23 the fluence using the measurement method to be validated. A calculated 24 spectrum-averaged cross-section, generally specified along with the certified 25 neutron fluence, must be used to derive the measurement value. If measured 26 and certified neutron fluence agree, the detector measurement method, includ 27 ing the detector cross-section, is validated. If the fluences disagree, this 28 calculation-to-experiment (C/E) ratio represents a bias associated with the 29 detector response measurement or the detector cross-section or both. In this 30 case, the detector measurement methods and input parameters should be re 31 examined in order to eliminate the bias. If after re-examination the bias is 32 still present, the bias may be used directly as a detector calibration factor.
33 Alternatively, when the bias factor is not too large it may be taken as an 21
1 indicator of the overall accuracy of the dosimetry measurement system for the 2 detector involved.
3 The neutron field referencing procedure should be reported in terms of 4 C/E ratios for the individual detectors and should include an uncertainty 5 table. The standard neutron field referencing may be used, as appropriate, to 6 simplify the uncertainty table called for in Regulatory Position 2.1.3 by 7 reducing or eliminating many uncertainties in the activity measurement, 8 nuclear decay parameters, and detector cross-sections.
9 Aside from validating the measurement method, standard and reference 10 neutron fields may be used for quality assurance of critical features of the 11 detector response. Examples are activation interference by impurities, proper 12 determination of fission product activities, and mass assay of fissionable 13 deposits for track recorders.
14 2.3 Fluence Determination from Detector Measurements 15 A fast neutron fluence (E > 1 MeV) should be obtained for each detector 16 as the quotient of the measured reaction probability and a calculated 17 spectrum-averaged cross-section (truncated at I MeV) from the neutron trans 18 port calculation. These measured fluences and a suitably weighted average 19 fluence must be reported along with an uncertainty that is an appropriate 20 combination of the entries in the uncertainty table (described in Regulatory 21 Position 2.1.3) combined with whatever cross-section uncertainty is available 22 from the calculation. An alternative to deriving a neutron fluence from the 23 detector responses is to directly compare measured reaction probabilities or 24 average reaction rates with results from the neutron transport calculation.
25 Whatever alternative is taken, the C/E ratios 7 for these quantities must 26 be reported along with the measurement uncertainties. When the C/E ratios 27 differ from unity by less than the assigned uncertainties, they validate the 28 calculation to within the accuracy of the measurements. If the C/E ratios 29 differ from unity by more than -20% for in-vessel dosimetry or -30% for cavity 30 dosimetry, the measurement and calculation must be reexamined (see Regulatory 31 Position 1.4.2). If an individual detector is declared suspect, it may with 32 71t should be noted that these C/E ratios differ from those discussed in 33 34 the previous section in that they refer to a comparison of in-vessel fluences rather than calibration fluences.
ýj 22
I justification be disregarded or given reduced weight.- This procedure must be 2 documented.
3 2.4 Sites for Updated Dosimetry 4 As-built dosimetry in surveillance capsules cannot easily be updated.
5 Furthermore, the dosimetry irradiated with metallurgical specimens is only 6 available at infrequent intervals. However, additional and upgraded dosimetry 7 is important for understanding and following vessel exposures, especially for 8 low-leakage core modifications. The ex-vessel cavity may be used as an alter 9 native site for installing additional improved dosimetry. Recent pressure 10 vessel benchmark experiments (Refs. 3, 55, 56) have demonstrated that the ex 11 vessel dosimetry can provide useful exposure information within the pressure 12 vessel wall (Refs. 32, 57) and, when placed at appropriate circumferential 13 locations, is a good monitor of the effectiveness of low-leakage core 14 strategies.
15 2.5 Experimental Benchmarks for Validating Calculations 16 A series of pressure vessel benchmark experiments were performed in the 17 1980s (Ref. 58) to test and validate particular aspects of neutron transport 18 calculations and improve pressure vessel surveillance dosimetry measurements 19 (Refs. 3, 27, 32, 59-61). Specific designs were incorporated into the experi 20 mental facilities to address calculational problems. Fluence transport cal 21 culations were carried out by several laboratories, and dosimetry measurements 22 using different techniques were compared to provide experimental results with 23 well-known and documented uncertainties. These benchmarks may be used to 24 define calculational uncertainties and to test nuclear data, transport 25 calculation methods, and the simulation of three-dimensional geometry.
26 The benchmark experiments were carried out in three types of 27 configurations:
28 2.5.1 Pressure Vessel Simulator Mockup Experiments 29 Pressure vessel simulator mockups of the vessel in the vicinity of test 30 reactors (Ref. 32) are unique in that they provide benchmarked dosimetry 31 measurements at the inner surface of the vessel and at locations within the 32 vessel wall (e.g., T/4 and 3T/4). These benchmarks are further characterized 23
1 by relatively simple geometries with generally less uncertainty in region 2 compositions, temperatures, and source distributions than in operating power 3 reactor benchmarks.
4 2.5.2 Power Reactor Mockup Experiments 5 Reactor benchmarks have been used to evaluate pin-wise power distri 6 butions in peripheral fuel assemblies, to investigate three-dimensional 7 effects caused by partial-length shield assemblies, and to validate the 8 modeling of heterogeneities caused by neutron pads attached to the core 9 barrel (Refs. 27, 59).
10 2.5.3 Cavity Mockup Experiments 11 Benchmark measurements involving simulated reactor cavities are described 12 in References 3 and 32. Measurements performed in power reactor cavities are 13 described in References I and 2.
14 3. REPORTING 15 This guide does not specifically contain reporting requirements.
16 However, when fluence determinations are required by the regulations, the 17 licensee's documentation describing the determination of pressure vessel 18 fluence must provide a complete description of the methods used to calculate 19 and measure the neutron fluences. The specific regulatory positions on 20 reporting are given in this section and are summarized in Table 1.
21 3.1 Neutron Fluences and Uncertainties 22 3.1.1 Fluence Methods 23 The methods used to calculate the integral and multi-group fluences and 24 fluence rates and associated methods qualification should be reported. A 25 discussion of any deviations from the procedures provided in this regulatory 26 guide should be included. The source of the cross-section data, the numerical 27 methods (e.g., quadrature, mesh, and convergence criteria), and the treatment 28 of special effects (e.g., fuel burnup, axial effects, and pin power 29 distributions) should be described in detail.
24
1 3.1.2 Multiqroup Fluences 2 The calculated absolute multigroup neutron fluence and fluence rates and 3 their uncertainties at the peak wall (or other limiting) location arns at the 4 T/4 and 3T/4 positions within the pressure vessel should be reported. The 5 multigroup energy boundaries should be included. The value and basis for any 6 adjustments made to the calculated vessel fluences based on comparisons to 7 measurement should be described in detail.
8 3.1.3 Integral Fluences 9 The calculated [E > 1 MeV] and [E > 0.1 MeV] integral fluences and 10 fluence rates, together with the uncertainties at the vessel inner wall loca 11 tions, should be reported. The measured [E > 1 MeV] integral neutron fluences 12 and uncertainties for each measurement location should be reported. The 13 results of the measurement validation should also be described. If thermal 14 neutron fluence rate measurements have been performed, these should be 15 reported together with the uncertainty and thermal shield material.
16 3.2 Specific Activities and Average Reaction Rates 17 The specific activities at the end-of-irradiation and the measured 18 average reaction rates should be reported, together with the associated uncer 19 tainty tables and the power-time history. For each dosimeter, provide the 20 reaction type, dosimeter material and form (wire, foil, etc.), weight-percent 21 and isotopic-percent of the target material, fission yields, and half-lives.
22 The corrections for the detector response perturbations, interfering reac 23 tions, and photofission should also be described. The calculated reaction 24 rates corresponding to these measurements should be reported, together with 25 the C/E ratios and a composite uncertainty derived from the measurement and 26 calculation uncertainties.
27 D. IMPLEMENTATION 28 The purpose of this section is to provide information to applicants and 29 licensees regarding the NRC staff's plans for using this regulatory guide.
30 This draft regulatory guide has been released to encourage public 31 participation in its development. Except in those cases in which an applicant 32 proposes an acceptable alternative method for complying with specified 25
1 portions of the Commission's regulations, the methods to be described in the 2 active guide reflecting public comments will be used in the evaluation of 3 applications for new licenses and for evaluating compliance with 10 CFR 50.60 4 and 50.61.
26
1 TABLE 1.
SUMMARY
OF REGULATORY POSITIONS ON CALCULATION AND DOSIMETRY 2 Regulatory 3 Position 4 FLUENCE CALCULATION METHODS 5 Fluence Determination. Absolute fluence calculations, rather 1.3 6 than extrapolated fluence measurements, should be used for the 7 fluence determination.
8 Modeling Data. The calculation modeling (geometry, materials, 1.1.1 9 etc.) should be based on documented and verified plant-specific 10 data.
11 Nuclear Data. The latest version of the Evaluated Nuclear Data 1.1.2 12 File (ENDF/B) should be used for determining nuclear material 13 reaction cross-sections. Cross-section sets based on earlier 14 or equivalent nuclear data sets that have been thoroughly 15 benchmarked are also acceptable. When the recommended cross 16 section data change, the effect of these changes on the 17 licensee-specific methodology should be evaluated and the 18 fluence estimates updated.
19 Cross-Section Angular Representation. A PN angular 1.1.2 20 decomposition of the scattering cross-sections (at a minimum) 21 should be employed.
22 Cross-Section Group Collapsing. The adequacy of the collapsed 1.1.2 23 job library should be demonstrated. This may be accomplished 24 by comparing calculations for a representative configuration 25 performed with both the master library and the job library.
26 Neutron Source. The core neutron source should account for 1.2 27 local fuel isotopics and, where appropriate, moderator density.
28 The neutron source normalization and energy dependence must 29 account for the fuel exposure dependence of the fission 30 spectra, the number of neutrons produced per fission, and the 31 energy released per fission.
32 End-of-Life Predictions. Predictions of the vessel end-of-life 1.2 33 fluence should be made with a best-estimate or conservative 34 generic power distribution. If a best estimate is used, the 35 power distribution should be updated if changes in core load 36 ings, surveillance measurements, or other information indicate 37 a significant change in projected fluence values.
38 Spatial Representation. Neutron transport calculations should 1.3.1 39 incorporate a detailed radial and azimuthal spatial mesh of -2 40 intervals per inch radially and 40-80 intervals per octant.
41 The discrete ordinates calculations must employ (at a minimum) 42 an S. quadrature.
27
TABLE 1.
(Continued)
Regulatory Position I Multiple Transport Calculations. If the calculation is 1.3.1 2 performed using two or more "bootstrap" calculations, the 3 adequacy of the overlap regions should be demonstrated.
4 Capsule Modeling. The capsule fluence is extremely sensitive 1.3.1 5 to the geometrical representation of the capsule geometry and 6 internal water region, and the adequacy of the capsule 7 representation and mesh should be demonstrated.
8 Spectral Effects on RT In order to account for the neutron 1.3.1 9 spectrum dependence ofNVTNDT, when the fluence is extrapolated 10 from the surveillance capsule or from the inside of the pres 11 sure vessel to the T/4 and 3T/4 vessel locations, a spectral 12 lead factor must be applied for the calculation of ARTNDT.
13 Cavity Calculations. The adequacy of the S angular quadrature 1.3.3 14 used in cavity transport calculations must Se demonstrated.
15 Methods Qualification. The calculational methodology should be 1.4.2, 2.5 16 qualified by both (1) comparisons to measurement and calcula 17 tional benchmarks and (2) an analytic uncertainty analysis.
18 The methods used to calculate the benchmarks should be con 19 sistent with the methods used to calculate the vessel fluence.
20 The overall calculational uncertainty should be determined by 21 an appropriate combination of the analytic uncertainty analysis 22 and the uncertainty estimate based on the comparisons to the 23 benchmarks.
24 Fluence Calculational Uncertainty. The vessel fluence 1.,
25 £1 sigma) calculational uncertainty must be demonstrated to be 1.3.2, 26 <20% for RTPTs determination. If the benchmark comparisons 1.4.3 27 indicate an uncertainty greater than -20%, the calculational 28 model must be adjusted or a bias applied to bring the agreement 29 within this range. For non-PTS applications, the accuracy 30 should be determined using the approach described in Regulatory 31 Position 1.4, and an uncertainty allowance should be included 32 in the fluence estimate as appropriate in the specific 33 application.
34 FLUENCE MEASUREMENT METHODS 35 Spectrum Coverage. The set of dosimeters must provide adequate 2.1.1 36 spectrum coverage.
37 Isotopic Composition. Use of dosimeter materials should 2.1.1 38 address material purity, total and isotopic mass assay, 39 perturbations by encapsulations and thermal shields, and 40 accurate dosimeter positioning.
28
TABLE 1.
(Continued)
Regulatory Position 1 Corrections. Dosimeter response measurements should account 2.1.2 2 for fluence rate variations, isotopic burnup effects, detector 3 perturbations, self shielding, reaction interferences, and 4 photo fission.
5 Response Uncertainty. An uncertainty analysis should be 2.1.3 6 performed for the response of each dosimeter.
7 Validation. Detector response calibrations should be carried 2.2 8 out periodically in a standard neutron field.
9 Calculation-To-Experiment Ratios. The C/E ratios, the standard 2.3 10 deviation and bias between calculation and measurement should 11 be determined.
12 Fast Neutron Fluence. The [E > 1 MeV] fast neutron fluence for 2.3 13 each measurement location should be determined using calculated 14 spectrum-averaged cross-sections and individual detector 15 measurements.
16 REPORTING PROVISIONS 17 Neutron Fluence and Uncertainties 18 Details of the absolute fluence calculations and associated 3.1 19 methods qualification should be reported, and the justification 20 and description for any deviations from the provisions of this 21 guide should be provided.
22 Calculated multigroup neutron fluences, fluence rates, and 3.1 23 their uncertainties should be reported.
24 Calculated integral fluences and fluence rates for E > 1 MeV 3.1 25 and E > 0.1 MeV and their uncertainties should be reported.
26 Measured integral E > 1 MeV fluences and uncertainties for each 3.1 27 measurement location should be reported.
28 The value and basis of any bias or model adjustment made to 3.1 29 improve the calculation-to-measurement agreement should be 30 reported.
31 The results of the standard field validation of the measurement 3.1 32 method should be reported.
29
TABLE 1.
(Continued)
Regulatory Position 1 Specific Activities and AveraQe Reaction Rates 2 Measured specific activities at the end of irradiation and 3.2 3 derived average reaction rates with uncertainties should be 4 reported.
5 The corresponding calculated reaction rates and C/E ratios with 3.2 6 uncertainties should be reported.
7 All corrections and adjustments to the measured quantities and 3.2 8 their justification should be reported.
30
1 TABLE 2. THRESHOLD DETECTORS RECOMMENDED FOR PRESSURE VESSEL DOSIMETRY 2 Nominal Applicable 3 Threshold (MeV) ASTM Standards 4 Neptunium-237 0.6 E705 (Ref. 46) 5 Uranium-238 1.5 E704 (Ref. 47) 6 Nickel-58 2.1 E264 (Ref. 48) 7 Iron-54 2.3 E263 (Ref. 49) 8 Titanium-46 3.8 E526 (Ref. 50) 9 Copper-63 4.7 E523 (Ref. 51) 31
1 References 2 1. R.E. Maerker et al., "Applications of the LEPRICON Unfolding Procedures 3 to the Arkansas Nuclear One-Unit 1 Reactor," Nuclear Science and 4 Engineering, Vol. 93(2), pp. 137-170, June 1986.
5 2. R.E. Maerker, "LEPRICON Analysis of Pressure Vessel Surveillance 6 Dosimetry Inserted into H.B. Robinson-2 During Cycle 9," Nuclear Science 7 and Engineering, Vol. 96(4), pp. 263-289, August 1987.
8 3. R.E. Maerker, "Analysis of the NESDIP2 and NESDIP3 Radial Shield and 9 Cavity Experiments," NUREG/CR-4886 (Oak Ridge National Laboratory, 10 ORNL/TvM-10389), USNRC, May 1987.1 11 4. C.Y. Fu and D.M. Hetrick, "Update of ENDF/B-V Mod-3 Iron: Neutron 12 Producing Reaction Cross-Sections and Energy Angle Correlations,"
13 ORNL/TM-9964, ENDF-341, Oak Ridge National Laboratory, July 1986.2 14 5. R.W. Roussin et al., "VITAMIN-C 171 Neutron, 36 Gamma Ray Group Cross 15 Sections in CCCC Interface Format for Fusion and LMFBR Neutronics," DLC 1977.2 16 41, Oak Ridge National Laboratory, September 17 6. R.W. Roussin et al., "VITAMIN-E: 174 Neutron, 38 Gamma-Ray Multigroup 18 Cross-Section Library for Deriving Application-Dependent Working 19 Libraries for Radiation Transport Calculations," DLC-113c, Oak Ridge 20 National Laboratory, November 1987.2 21 'Copies are available for inspection or copying for a fee from the NRC 22 Public Document Room at 2120 L Street NW., Washington, DC; the PDR's mailing 23 address is Mail Stop LL-6, Washington, DC 20555; telephone (202)634-3273; 24 fax (202)634-3343. Copies may be purchased at current rates from the U.S.
25 Government Printing Office, Post Office Box 37082, Washington, DC 20013-7082 26 (telephone (202)512-2249 or (202)512-2171); or from the National Technical 27 Information Service by writing NTIS at 5285 Port Royal Road, Springfield, VA 28 22161.
29 2Available from the Radiation Shielding Information Center, Oak Ridge National 30 Laboratory, Post Office Box 2008, Oak Ridge, TN 37831-6362.
32
References (continued) 1 7. W.E. Ford, III, et al., "Modification Number One to the lOOn - 21g Cross 2 Section Library," ORNL/TM-5249 (Available as DLC-37D/EPR from RSIC), Oak 3 Ridge National Laboratory, March 1976.2 4 8. G.L. Simmons and R.W. Roussin, "SAILOR-Coupled, Self-Shielded, 47 Neu 5 tron, 20 Gamma Ray, P3, Cross-Section Library for Light Water Reactors,"
6 RSIC-DLC-76, Oak Ridge National Laboratory, June 1980.2 7 9. M.L. Williams et al., "The ELXSIR Cross-Section Library for LWR Pressure 8 Vessel Irradiation Studies: Part of the LEPRICON Computer Code System,"
9 NP-3654, Electric Power Research Institute, September 1984.3 10 10. R.E. Maerker, M.L. Williams, and B.L. Broadhead, "Accounting for Changing 11 Source Distributions in Light Water Reactor Surveillance Dosimetry 12 Analysis," Nuclear Science and Engineering, Vol. 94, pp. 291-308, 1986.
13 11. R.E. Maerker, M.L. Williams, and B.L. Broadhead, "TIMEPATCH: A Module in 14 the LEPRICON Computer Code System for Evaluating Effects of Time 15 Dependent Source Distributions in PWR Surveillance Dosimetry," EPRI 16 Interim Report, December 1985.2 17 12. M. Todosow and J.F. Carew, "Evaluation of Selected Approximations Used in 18 Pressure Vessel Fluence Calculations," Transactions of the American 19 Nuclear Society, Vol. 46, p. 658, June 1984.4 20 13. W.J. Eich, "Advanced Recycle Methodology Program," Part II, Chapter 5, 21 Research Project 118-1, Electric Power Research Institute, 1976.3 22 3Available from EPRI Research Reports Center, P.O. Box 50490, Palo Alto, 23 CA 94303.
24 4Available from the American Nuclear Society, 555 N. Kensington Avenue, 25 La Grange Park, Illinois 60525.
33
References (continued) 1 14. A. Ahlin et al., "CASMO - A Fuel Assembly Burnup Program," AE-RF-76-4158 2 (Rev. Ed.), Studsvik Energiteknik AB, 1978. (Available from Studsvik of 3 America Inc., 1087 Beacon St., Newton, Mass. 02159).
4 15. A.L. Aronson et al., "Evaluation of Methods for Reducing Pressure Vessel 5 Fluence," BNL-NUREG-32876, Battelle National Laboratory, March 1983.2 6 16. G.P. Cavanaugh et al., "Reduction in Reactor Vessel Irradiation Through 7 Fuel Management," Transactions of the American Nuclear Society, Vol. 45, 8 p. 98, October 1983.4 9 17. M. Todosow et al., "Pressure Vessel Fluence Reduction Through Selective 10 Fuel Assembly Replacement," Transactions of the American Nuclear Society, 11 Vol. 45, p. 595, October 1983.4 12 18. D. Cokinos et al., "Pressure Vessel Damage Fluence Reduction by Low 13 Leakage Fuel Management," Transactions of the American Nuclear Society, 14 Vol. 45, p. 594, October 1983.4 15 19. J.T. West, "SORRELL," in "DOT 3.5-A Two-Dimensional Discrete Ordinate 16 Transport Code," in CCC-276, Oak Ridge National Laboratory, 1978.2 17 20. M.L. Williams, "DOTSOR: A Module in the LEPRICON Computer Code System 18 for Representing the Neutron Source Distribution in LWR Cores," EPRI 19 Interim Report, December 1985.2 20 21. "Standard Guide for Application of Neutron Transport Methods for Reactor 21 Vessel Surveillance, E 706 (IID)," ASTM E482-89, American Society for 22 Testing and Materials, Philadelphia, 1989.'
23 22. "DOT 3.5 - A Two-Dimensional Discrete Ordinate Transport Code," CCC-276, 24 Oak Ridge National Laboratory, 1978.2 25 5Available from the American Society for Testing and Materials, 1916 Race 26 Street, Philadelphia, PA 19103-1187.
34
References (continued) 1 23. W.A. Rhoades and R.L. Childs, "An Updated Version of the DOT 4 One-and 2 Two-Dimensional Neutron/Photon Transport Code," ORNL-5851, Oak Ridge 3 National Laboratory, July 1982.2 4 24. F.B.K. Kam et al., "Pressure Vessel Fluence Analysis and Neutron Dosi 5 metry," NUREG/CR-5049 (Oak Ridge National Laboratory, ORNL/TM-10651),
6 December 1987.1 7 25. J.F. Carew, D.K. Min, A.L. Aronson, "Spectral Effects in the Extrapola 8 tion of Pressure Vessel Surveillance Capsule Measurements," Nuclear 9 Technology, Vol. 55, No. 3, p. 565, December 1981.
10 26. M.L. Williams, P. Chowdhury, and B.L. Broadhead, DOTSYN: A Module for 11 Synthesizing Three-Dimensional Fluxes in the LEPRICON Computer Code 12 System, EPRI Interim Report, December 1985.2 13 27. R.E. Maerker, "Analysis of the VENUS-3 Experiments," NUREG/CR-5338 (Oak 14 Ridge National Laboratory, ORNL/TM-11106), August 1989.1 15 28. L. Lois and T. Collins, "Reactor Cavity Dosimetry - A Regulatory Perspec 16 tive," Transactions of the American Nuclear Society, Vol. 63, p. 431, 17 June 1991.4 18 29. N. Tsoulfanidis, "Neutron Energy Spectra in the Core and Cavity of the 19 ANO-2 PWR," NP-4238, Electric Power Research Institute, September 1985.3 20 30. J.F. Carew et al., "Pressure Vessel Fluence Benchmark Calculations," BNL 21 NUREG-34715, Brookhaven National Laboratory, February 1984.2 22 31. D.M. Cokinos et al., "Benchmarking of Pressure Vessel Fluence Calcula 23 tions," Transactions of the American Nuclear Society, Vol. 46, p. 636, 24 June 1984.'
35
References (continued) 1 32. W.N. McElroy, Editor, "LWR Pressure Vessel Surveillance Dosimetry 2 Improvement Program: PCA Experiments and Blind Test," NUREG/CR-1861, 3 (Hanford Engineering Development Laboratory, HEDL-TME 80-87), July 1981.1 4 33. "LWR Pressure Vessel Damage Fluence Standard Problems," NUREG/CR-6115 5 (BNL NUREG-52395), Brookhaven National Laboratory (to be published).'
6 34. J. Celnik, "Shielding Benchmarks," Transactions of the American Nuclear 7 Society, Vol. 33, p. 662, November 1979.4 8 35. B.L. Broadhead et al., "LEPRICON Adjustment Module: A Generalized Linear 9 Least Squares Data Analysis Program with Application to PWR Surveillance 10 Dosimetry," EPRI Interim Report, March 1985.2 11 36. R.E. Maerker, B.L. Broadhead, and J.J. Wagschal, "Theory of a New 12 Unfolding Procedure in Pressurized Water Reactor Pressure Vessel Dosi 13 metry and Development of an Associated Benchmark Data Base," Nuclear 14 Science and Engineering, Vol. 91(4), pp. 369-392, December 1985.
15 37. W.N. McElroy et al., "A Computer-Automated Iterative Method for Neutron 16 Flux Spectral Determination by Foil Activation," AFWL-TR-67-41, Vol. I, II). 2 17 1967 (available as RSIC Program No. CCC-112/SAND 18 38. C.A. Oster et al., "A Modified Monte Carlo Program for SAND-II with 19 Solution Weighing and Error Analysis," Hanford Engineering Development 20 Laboratory, HEDL-TME 76-60, 1976.6 21 39. C.R. Green, J.A. Halbleib, and J.V. Walker, "A Technique for Unfolding 22 Neutron Spectra from Activation Measurements," Sandia Corporation, SC 2
23 RR-67-746, 1967 (available as RSIC Program No. CCC-108/SPECTRA).
24 6Available from the National Technical Information Service, 5285 Port Royal Road, 25 Springfield, VA 22161; phone (703):487-4650..
36
References (continued) 1 40. F.G. Perey, "Least-Squares Dosimetry Unfolding: The Program STAY'SL,"
2 ORNL/TM-6062, Oak Ridge National Laboratory, 1977 (available as RSIC 3 Program No. PSR-113). 2 4 41. F. Schmittroth, "FERRET Data Analysis Code," HEDL-TME 79-40, 1979 5 (available as RSIC Program No. PSR-145). 2 6 42. F.W. Stallmann, "LSL-M2: A Computer Program for Least-Squares Logarith 7 mic Adjustment of Neutron Spectra," NUREG/CR-4349 (Oak Ridge National 8 Laboratory, ORNL/TM-9933), March 1986 (available as RSIC Program No.
9 PSR-233).'"'
10 43. "Standard Guide for Sensor Design and Irradiation for Reactor Surveil 11 lance, E 706 (IIC)," ASTM E 844-86, American Society for Testing and 12 Materials, Philadelphia, 1986.s 13 44. "Standard Test Method for Application and Analysis of Radiometric 14 Monitors for Reactor Vessel Surveillance, E 706 (IIIA)," ASTM E 1005-84, 15 American Society for Testing and Materials, Philadelphia, 1984.S 16 45. "Standard General Methods for Detector Calibration and Analysis of Radio 17 nuclides," ASTM E 181-82, American Society for Testing and Materials, 18 Philadelphia, 1983.5 19 46. "Standard Test Method for Measuring Reaction Rates by Radioactivation 20 of Neptunium-237," ASTM E 705-90, American Society for Testing and 5
21 Materials, Philadelphia, 1991.
22 47. "Standard Test Method for Measuring Reaction Rates by Radioactivation of 23 Uranium-238," ASTM E 704-90, American Society for Testing and Materials, 24 Philadelphia, 1991.5 25 48. "Standard Test Method for Measuring Fast-Neutron Reaction Rates by 26 Radioactivation of Nickel," ASTME 264-87, American Society for Testing 27 and Materials, Philadelphia, 1987. -
37
References (continued) 1 49. "Standard Test Method for Measuring Fast-Neutron Reaction Rates by 2 Radioactivation of Iron," ASTM E 263-88, American Society for Testing and 3 Materials, Philadelphia, 1988.5 4 50. "Standard Test Method for Measuring Fast-Neutron Reaction Rates by 5 Radioactivation of Titanium," ASTM E 526-87, American Society for Testing 6 and Materials, Philadelphia, 1987.5 7 51. "Standard Test Method for Measuring Fast-Neutron Reaction Rates by 8 Radioactivation of Copper," ASTM E 523-87, American Society for Testing 9 and Materials, Philadelphia, 1987.5 10 52. R. Gold et al.,."Neutron Dosimetry with Solid-State Track Recorders in 11 the Three-Mile Island Unit-2 Reactor Cavity," Nuclear Tracks, Vol. 103, 12 p. 447, 1985.
13 53. "Standard Test Method for Application and Analysis of Solid State Track 14 Recorder (SSTR) Monitors for Reactor Surveillance, E 706 (IIIB)," ASTM 15 E 854-90, American Society for Testing and Materials, Philadelphia, 16 1990.5 17 54. J.A. Grundl and C.M. Eisenhauer, Compendium of Benchmark Neutron Fields 18 for Reactor Dosimetry, NBSIR 85-3151, National Bureau of Standards, 19 Gaithersburg, MD, January 1986.7 20 55. E.P. Lippincott et al., Westinghouse-Nuclear Technology Division, 21 Pittsburgh, PA, "Evaluation of Surveillance Capsule and Reactor Cavity 22 Dosimetry from H.B. Robinson Unit-2, Cycle 9," USNRC Report NUREG/CR-4576 23 (WCAP-11104), February 1987. (Copies are available for inspection or 24 copying for a fee from the NRC Public Document Room at 2120 L Street NW.,
25 Washington, DC; the PDR's mailing address is Mail Stop LL-6, Washington, 26 DC 20555; telephone (202) 634-3273; fax (202) 634-3343.)
27 7Available from the National Institute of Standards and Technology, NIST 28 Publication Productions, Room A635, Gaithersburg, MD 20899.
38
References (continued) 1 56. F.H. Ruddy et al., "Solid-State Track Recorder Neutron Dosimetry in 2 Light-Water Reactor Pressure Vessel Surveillance Mockups in Reactor 3 Dosimetry," Proceedings of the 5th ASTM-EURATOM Symposium on Reactor 4 Dosimetry, Geesthacht, Federal Republic of Germany, Sept. 24-28, 1984, 5 EUR 9869, Commission of the European Communities, D. Reidel Publishing 6 Co., 1985.
7 57. W.N. McElroy, R. Gold, E.D. McGarry, Eds., "LWR-Pressure Vessel 8 Surveillance Dosimetry Improvement Program; PSF Startup Experiments,"
9 NUREG/CR-3320, Vol. 2 (WHC-EP-0204/R5), July 1992.1 10 58. "Standard Master Matrix for Light-Water Reactor Pressure Vessel 11 Surveillance Standards, E 706 (0)," ASTM E 706-84, American Society for 12 Testing and Materials, Philadelphia, 1988.'
13 59. P. D'hondt et al., "Contributions of the Venus-Engineering Mock-Up Experi 14 ments to the LWR-PV Surveillance," Proceedings of the 7th ASTM-Euratom 15 Symposium on Reactor Dosimetry, Strasbourg, France, August 1990.2 16 60. R.E. Maerker and B.A. Worley, "Activity and Fluence Calculations for the 17 Startup and Two-Year Irradiation Experiments Performed at the Poolside 18 Facility," NUREG/CR-3886 (Oak Ridge National Laboratory, ORNL/TM-4265),
19 October 1984.'
20 61. W.N. McElroy, Editor, "LWR Pressure Vessel Surveillance Dosimetry 21 Improvement Program. PSF Experiments Summary and Blind Test Results,"
22 NUREG/CR-3320, Vol. 1, Hanford Engineering Development Laboratory, HEDL 23 TME 86-8), July 1986.1 39
1 REGULATORY ANALYSIS 2 1. STATEMENT OF THE PROBLEM 3 The U.S. Nuclear Regulatory Commission (NRC) has promulgated regulations 4 to ensure the structural integrity of the reactor pressure vessel for light 5 water power reactors. Specific fracture toughness requirements for normal 6 operation and for anticipated operational occurrences for power reactors are 7 set forth in Appendix G, "Fracture Toughness Requirements," of 10 CFR Part 50, 8 "Domestic Licensing of Production and Utilization Facilities." Additionally, 9 in response to concerns over potential pressurized thermal shock (PTS) events 10 in pressurized water reactors (PWRs), the NRC issued 10 CFR 50.61, "Fracture 11 Toughness Requirements for Protection Against Pressurized Thermal Shock 12 Events."
13 To satisfy the requirements of both Appendix G and 10 CFR 50.61, methods 14 for determining the fast neutron fluence (E > MeV) are necessary to estimate 15 the fracture toughness of the pressure vessel materials. Appendix H, "Reactor 16 Vessel Material Surveillance Program Requirements," of 10 CFR Part 50 requires 17 the installation of surveillance capsules, including material test specimens 18 and flux dosimeters, to provide data on material damage correlations as a 19 function of fluence.
20 The neutron fluence is attenuated by several decades between the core and 21 vessel. This attenuation results in a strong sensitivity of the calculated 22 vessel fluence to the physical description of the core and vessel internals 23 and the numerical calculation of the neutron transport, and it makes an accu 24 rate determination of the pressure vessel fluence difficult. As a result, a 25 wide range of methods of varying reliability and accuracy have been used to 26 determine the reactor vessel fluence. Consequently, comparisons of measured 27 and calculated fluences have shown varying degrees of agreement, and in some 28 cases conservatisms have been required in licensing analyses to accommodate 29 the observed calculation-to-measurement differences.
30 Over the past decade, substantial improvements have been made in both 31 the calculation and measurement of the pressure vessel fluence. These 32 improvements have stemmed from both NRC and industry programs. These improve 33 ments include the development and improvement of computer codes and calcula 34 tional models, revision of basic cross-section data, improved measurement R/A-I
1 techniques, and the systematic qualification of the fluence methods by 2 comparison to NRC-sponsored benchmark experiments.
3 These calculation and measurement improvements provide increased accuracy 4 in the fluence determinations that are an essential part of meeting the 5 requirements of 10 CFR Part '0,Appendices G and H, and 10 CFR 50.61. This is 6 especially important for 91nts seeking to renew their operating licenses.
7 The wide variatiop in flýence calculation methods has resulted in lengthy 8 plant-specific reviews and made it difficult to ensure that the actual fluence 9 is adequately bounded by the calculational methods. The calculation and dosi 10 metry guide would provide standardized methods and procedures that would allow 11 these reviews to be greatly simplified, and would improve confidence in the 12 calculated fluence values.
13 2. OBJECTIVE 14 The objective of this guide is to provide state-of-the-art calculation 15 and measurement procedures that are acceptable to the NRC staff for determin 16 ing pressure vessel fluence. These procedures would yield a more accurate and 17 reliable vessel fluence determination than is generally employed at the pre 18 sent time. The improved accuracy and realistic assessment of the uncertainty 19 in the calculation would provide assurance that the fluence value is appro 20 priate for use in evaluating compliance with the requirements of 10 CFR 21 Part 50, Appendices G and H, and 10 CFR 50.61.
22 3. ALTERNATIVES 23 The alternatives to issuing the vessel fluence calculation and dosimetry 24 regulatory guide are as follows:
25 3.1 Branch Technical Position 26 The pressure vessel fluence methods provided by the regulatory guide 27 could be included in a branch technical position. However, this is not con 28 sidered an acceptable alternative since the branch position does not provide 29 as wide a consensus, especially from the public and the ACRS, which the vessel 30 fluence analysis requires.
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1 3.2 NUREG-Series Report 2 The vessel fluence procedures could be published in a NUREG-series 3 report. However, these reports also do not receive the required consensus 4 from industry and the public, and they are not appropriate for providing 5 regulatory guidance.
6 3.3 Discussions with Licensees 7 The detailed fluence calculational and measurement methods that are 8 considered acceptable to the NRC staff could be provided to the licensees 9 through individual reviews and discussions. This alternative is basically the 10 same as the current practice and is equivalent to taking no action. Individ 11 ual licensee discussions are extremely time-consuming for both the NRC staff 12 and licensee, they lead to highly individual analyses and reviews, and they do 13 not result in an established standard.
14 4. COSTS AND BENEFITS 15 4.1 Benefits 16 The methods described in this guide may be used for all fluence 17 determinations used in vessel fracture toughness evaluations, including the 18 determination of the fluence used in calculating the pressure vessel mate 19 rials' values of RTNDT for use in 10 CFR Part 50, Appendix G, and the values 20 of RTPTS in accordance with 10 CFR 50.61. The regulatory guide would improve 21 the accuracy and reliability of these evaluations and ensure consistency with 22 the present regulatory position on vessel fluence uncertainty by incorporating 23 state-of-the-art methods and procedures for determining the fluence and the 24 fluence uncertainty. The guide would also ensure the completeness of licensee 25 vessel fluence submittals and improve the efficiency of staff reviews.
26 While it is recognized that the improved fluence determination may be 27 used by licensees to relax conservatisms in the present fluence calculations, 28 it is expected the improved fluence estimates will result in enhanced safety.
29 In this regard it is noted that, for a pressure vessel near the PTS screening 30 criteria of 10 CFR 50.61, a 25% reduction in calculated end-of-license R/A-3
1 fluence, which is typical of existing uncertainties, will reduce the 2 calculated vessel failure frequency by approximately a factor of three.
3 4.2 NRC Costs 4 The NRC costs for reviewing fluence-related submittals would be reduced 5 substantially by the issuance of this guide. For estimating the costs, it is 6 assumed that of the -60 PWRs, half have a RTPTs within -40°F of the PTS 7 screening criterion (or other temperature limit) at the end of license and 8 will require a detailed review. Assuming each submittal requires a staff week 9 and only half of the PWR owners submit revised fluence analyses, the total NRC 10 cost is approximately 15 staff weeks. If the licensees used the methods given 11 in the guide, this cost could be reduced to approximately 2 staff weeks.
12 4.3 Licensee Costs 13 Increased costs to the licensee would result from changes in the fluence 14 calculation and measurement procedures. The calculational costs would be one 15 time costs and have been estimated in Table 1. The licensee costs resulting 16 from the changes in the measurement procedures have been estimated in Table 2.
17 5. DECISION RATIONALE 18 It is recommended that the proposed regulatory guide be issued because 19 (1) a wide consensus of the NRC, ACRS, industry, and the public would be pro 20 vided, (2) a methodology standard would be established, and (3) inefficient 21 use of the NRC staff, vendor and licensee resources would be eliminated.
22 The alternatives identified above for providing acceptable fluence 23 methods to the licensees do not provide the level of review and wide consensus 24 required for the vessel fluence determination used to ensure adequate pressure 25 vessel fracture toughness. In addition, while these approaches may result in 26 the same or slightly increased cost to licensees, they result in a highly 27 inefficient use of the NRC staff resources. The alternatives to a regulatory 28 guide are therefore judged to be unacceptable.
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1 TABLE 1. ADDITIONAL LICENSEE CALCULATION COSTS 2 Calculation Tasks Staff Weeks 3 (a) Modifications to calculational models +2 4 (b) Additional calculation benchmarking +8 5 and qualification 6 (c) Calculation uncertainty analysis +3 7 (d) Calculation documentation and reporting +2 8 (e) Reduced Licensing Activities* -4 9 Total Additional Licensee Cost +11 staff weeks 10 TABLE 2. ADDITIONAL LICENSEE MEASUREMENT COSTS 11 Measurement Task Staff Weeks 12 (a) Additional Quality Control +1 13 (b) Dosimeter Response Corrections +1 14 (c) Periodic Detector Calibration +2 15 (d) Response Uncertainty Analysis +2 16 (e) Additional Measurement +1 17 Documentation and Reporting 18 Total Additional Licensee Cost + 7 staff weeks 19 *This estimate reflects the reduced licensee costs from not attending meetings 20 with the NRC staff and responding to questions, as a result of following the 21 procedures in the regulatory guide.
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Federal Recycling Program FIRST CLASS MAIL UNITED STATES POSTAGE AND FEES PAID NUCLEAR REGULATORY COMMISSION USNRC PERMIT NO. G-67 WASHINGTON, D.C. 20555-0001 OFFICIAL BUSINESS PENALTY FOR PRIVATE USE, $300