ML20135A557

B&W Owners Group Cavity Dosimetry Program

ML20135A557
Person / Time
Site:	Davis Besse, Oconee, Arkansas Nuclear, Crystal River, Rancho Seco, 05000000, Crane
Issue date:	08/31/1985
From:	Shannon King BABCOCK & WILCOX CO.
To:
Shared Package
ML20135A527	List: ML20135A525 ML20135A557
References
BAW-1875, NUDOCS 8509100193
Download: ML20135A557 (23)

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BAW 1875 August 1985 I I I I The B&W Owners Group I Cavity Dosimetry Program I I by g S. Q. King I I Prepared for l B&W Owners Group Materials Committee Arkansas Power & Light Company I Duke Power Company Florida Power Company GPU Nuclear l Sacramento Municipal Utility District Toledo Edison Company l Prepared by BABCOCK & WILCOX 'l Nuclear Power Division P. O. Box 10935 Lynchburg, Virginia 24506-0935 I Babcock & Wilcox a McDermott company

!I ll TABLE OF CONTENTS lI 1. Introduction lI l

===2. Background=== 1I 3. The Cavity Dosimetry Program 3.1 Structure of the Cavity Dosimetry Program I 3.2 Cavity Dosimetry Benchmark Experiment 3.3 Plant Specific Program 4. Conclusions 5. Definition of Terms I I I I I I I I I

I I INTRODUCTION 1 The Cavity Dosimetry Program to be performed by the B&W Owners Group responds to NRC requirements that commerical nuclear plants monitor reactor vessel neutron fluence throughout the vessel lifetimes. The Cavity Dosimetry Program will include two experiments-the Flux Perturbation Study and the Cavity Dosimetry 1 l Benchmark -to demonstrate the ef fectiveness of ex-vessel dosimetry and to provide data upon which to develop and benchmark a method by which cavity dosimetry can be used to predict reactor vessel fluence. The program will produce a demonstrated technology for application to plant-specific cavity dosimetry installation to allow continuous monitoring of actual fluence and updating E of the predicted fluence for each reactor vessel. I

2. Background

The Reactor Vessel Surveillance Program (RVSP)1 became a requirement in the early 1970's when it was determined that significant differences occurred in the neutron embrittlement sensitivity of various steels and weldments used in reactor vessel f abrications. The applicable regulations are published in Appendix H of 10.CFR.50. In addition, American Society for Testing and Materials (ASTM) Standard E-185 supports the requirements of 10.CFR.50, Appendix H, " Reactor Vessel Material Surveillance Program Requirements" through a cooperative ef fort between the standards development committee and the NRC. (1) see page 21 l I

I Today, the regulatory requirements of Appendix H in conjunction with 10 CFR 50 Appendix G, " Fracture Toughness Requirements" are recognized throughout the nuclear industry as the standards and procedures for ensuring the integrity of reactor pressure vessels subjected to neutron irradiation during service. The l B&WOG, through the Integrated Reactor Vessel Surveillance Program l l (IRVSP), has followed these procedures and regulations and l consequently has obtained an increasingly accurate knowledge l of the condition of each reactor vessel in the IRVSP. I The condition of a pressure vessel can be ascertained at any point in plant life by knowing the accumulated fluence and the relationship between fluence and material toughness degradation. During the time that the RVSP is in effect, both of these factors are known; the fluence is measured by in-vessel dosimetry and the effect of the fluence on pressure vessel steel is obtained before the actual steel receives that level of fluence. With completion of the RVSP, the damage function will have been established but monitoring of the vessel fluence will cease unless it is provided by some other method. I It is anticipated that all of the empirical data needed to determine the relationship between accumulated neutron fluence and relevant material properties for the B&W 177 FA plants will have been collected by the early 1990's. Determining this relationship is, of course, the primary objective of the RVSP; once it has I

I I been established, the RVSP may be considered to be complete. I The RVSP has a secondary objective which is the indirect measurement of the fast neutron fluence on and through the pressure vessel. Since this fluence monitoring will stop with the completion of the RVSP, and there are reasons to continue to monitor neutron fluence over the life of the vessel, the B&WOG has developed I the Cavity Dosimetry Program, designed for that purpose. I The B&WOG Cavity Dosimetry Program will develop the technology necessary for the installation of neutron dosimetry for placement in the cavity of the B&W-designed 177FA reactors to monitor fluence indefinitely. At present, it is planned that this dosimetry will consist of Solid State Track Recorders (SSTR's), installed in special holders in the reactor cavity. This cavity-installed SSTR dosimetry has a number of advantages over capsule-contained Radiometric Material (RM) dosimetry: (a) The relative ease of installation and removal of the dosimeters allows for frequent fluence measurements. (b) Effect on refueling outage time is minimized; the dosimetry installation and removal operations are not carried out over the fuel, i.e. the Reactor Vessel head is in place. (c) Irradiated dosimeters (SSTRs) do not become significantly activated and thus are not radioactive. Handling and shipping is routine and thus less expensive. I I

I l (d) SSTR Dosimeter data do not decay; thus prompt counting is not mandatory. Before cavity dosimetry can be used to monitor vessel fluence, however, it needs to be demonstrated in a B&W plant configuration. Experience to date, by HEDL and others, has generally been positive. Thus, the primary intent of this paper is to explain the cavity Dosimetry Benchmark Experiment, and how the results will be used to establish an accurate method to determine vessel fluence from a knowledge of cavity fluence. I The important points of the preceeding discussion are restated below. The Reactor Vessel Surveillance Program has two purposes: determination of the relationship between fluence and material damage and the monitoring of the accumulation of vessel fluence. Between the time the RVSP ends and plant EOL, another method is needed to monitor vessel fluence. The method proposed by the B&W Owners Group for the surveillance of long-term vessel fluence is the Cavity Dosimetry Program. The cavity Dosimetry Program provides the technology necessary for long term surveillance of reactor vessel fluence. I

3. The B&MOG CAVITY DOSIMETRY PROCRAM The determination of vessel fluence from measurements is a semi-empirical process.

This is because there is no known way to measure vessel fluence directly. It can b2 measured indirectly, however, by first measuring a quantity that has a determinable I

i il or known relationship to vessel fluence and then calculating .I the vessel fluence through this known relationship. The measureable quantity that is most often used in vessel fluence determination (at the present time) is the activity of a foil or wire dosimeter. The dosimeter, with precisely known constituents and mass, is irradiated for a known length of time in a surveillance capsule. The capsule is located inside the Reactor Vessel (downcomer). The neutron-sensitive elements in the dosimeter become activated through various neutron absorption events occurring at a rate that is proportional to the incident neutron flux. The activity of the dosimeter after irradiation is related to the time-integrated I b neutron flux, or fluence, to which it has been exposed. The relationship between the fluence and the activity can be complicated depending on the dosimeter material, but methods have been developed that allow the determination of capsule fluence f rom dosimeter activity. The capsule fluence is not the vessel fluence, however, I and the relation. ship between the two is complex. The 2-D trant, port code, DOT-IV, is used to simultaneously calculate the vessel flux, the capsule flux, and the saturated dosimeter activity. Using the DOT-IV results with the evaluated dosimeter e:ata and the power history of the reactor, the vessel fluence is calculated. I This process is used to determine vessel fluence using in-vessel dosimetry measurements. A very similar procedure can be used to determine vessel fluence using ex-vessel dosimetry measurements, i.e. cavity-installed I

I I dosimetry. The Cavity Dosimetry Program provides the methodology to determine the vessel fluence based on measured fluence in the cavity; this methodology, called the "Semiempirical Method", is described in detail on pages 13 and 14 and in Figure 4. For either in-vessel dosimetry or ex-vessel dosimetry, actual measurements are taken at only one location, yet knowledge of l the fluence is required at numerous locations (such as to calculate 1 i the flux gradient through the vessel wall at the maximum azimuthal location). The accuracy of the prediction of the vessel flux depends on the accuracy of the DOT-IV analysis that relates the flux in the vessel to the flux at the dosimeter location. Accurate predictions of cavity flux are more difficult than in-vess el flux predictions because the problem is geometrically more complex compounded by the effects of cavity streaming. Accurate predictions are possible however, and they will be ensured by the use of an iterative refinement of the analytical model using the results of the Cavity Dosimetry Benchmark Experiment (In-Out Experiment). I The important points of the preceeding discussion are restated below. Monitoring vessel fluence is a process involving the indirect measurement of fluence at a single point and the application of benchmarked analytical procedures to result in energy and space dependent neutron fluence at numerous locations of interest. I I I

I I The use of cavity dosimetry in the process requires that the I analytical model be developed and refined using the results of the ex-vessel benchmark. 3.1 Structure of the Cavity Dosimetry Program The Cavity Dosimetry Program consists of the development of l a method to determine the vessel fluence using the results of l cavity dosimetry measurements as outlined in Figure (1). The results of the program (semiempirical method-block 22 of Figure

1) will be the subject of a topical report that will be submitted for NRC review (block 23).

Dosimetry and mounting hardware will be designed and installed in the cavity of the Davis Besse Reactor as shown in blocks (1-6) of Figure (1). After irradiation through cycle (6) (block 7) the dosimetry will be counted by various organizations followed by an initial B&W analysis of the results (blocks 8-11). All organizations concerned will be involved in refining the data and resolving discrepancies to result in the final " measured" set of dosimeter activities, (blocks 12-14). In parallel, B&W will develop a set of " calculated" dosimeter activities using the analytical methods shown in blocks (15-18) (and in more detail in Figure (3)). The calculated results will be compared to the measured results followed by iterations on the calculations as necessary as shown in blocks (19-22), resulting in the final form of the semiempirical method, described in section 3.2. I I

I I 3.2 CAVITY DOSIMETRY BENCHMARE EXPERIMENT The Cavity Dosimetry Benchmark Experiment (In-Out Experiment) features a full scale test in a B&W-designed reactor using ex-vessel .I and in-vessel measurements. Dosimetry will be supplied by four contributors to the program as shown in Table (1). The objectives of the program are to: (1) obtain measured fluence spectra at numerous locations at full scale to develop an optimum analytical model for the prediction of vessel fluence, and (2) i i assess the calculational and dosimetric uncertainties in the determination of vessel fluence. The experiment also allows estimation of the accuracy of SSTR's by comparison to other types of dosimeters (particularly RM' s ), and measurement of I the axial neutron flux profile in the cavity. Data will also be taken on cavity gamma flux for assessment of effects of photofission events on experimental results. I Fifteen dosimetry holders will be installed in the cavity of the host reactor as shown in Figure (2). Each holder has provision for five dosimeters (or sets of dosimeters) for a total capacity of 75 dosimeters (sets). At least two types of passive neutron detectors are planned to be used: RM's, and SSTR' s. In addition, the installation of Paired Uranium Detectors is planned. The threshold energies for the RM's cover the range from.5 eV to 2.5 MeV. Additionally, about twelve passive gamma detectors will be installed (LiF chips). I I

I I TABLE (1) CAVITY DOSIMETRY 1 I DOSIMETRY LOCATION OF TYPE OF I I CONTRIBUTOR DOSIMETER DOSIMETER a} COMMENTS b) OR OWNER PACKAGE 9 6 -mid p1 RM,SSTR,LiF Azimuthal n+Y fluence l3 i5 Distribution 0 Babcock & 11 -mid pl Azimutha1 dist 0 lE Wilcox 11.5 -mid pl PUD Fluence using alt. !5 Owners Group method (B&WOG) -up. act. fuel LiF Axial n + Y fluence Dist. I -nozzle PUD Axial n+Y fluence Dist. alt, method -seal plate LiF Axial n +Y fluence Dist. I 0 26.5 -mid pl RM, LIF Azimutha1 n +Y fluence dist. 0 42.5 -mid pl RM,SSTR,LiF Azimutha1 n +Y I fluence dist. 0 6 -mid pl RM,SSTR,LiF Independent measurments I 0 Hanford 11.5 -mid pl at same locations Engineering -up act fuel as most of the Dev. -nozzle B&WOG dosimetry 0 Lab. (HEDL) 42.5 -mid pl I 0 University 11 -mid pl RM Independent measurement 0 of Arkansas -over cavity Stainless at 11 (U of A) height Steel Chain axial dist. at 110 National All locations LiF measure gamma fluence 0 Bureau of 11.5 -mid pl PUD alt. meas. method Standards -nozzle PUD for neutron fluence I (NBS) Notes I (a) RM: Radiometric Monitor SSTR: Solid State Track Recorder LiF: Lithium Flouride Detector I PUD: Paired Uranium Detector (b) Overall purposes are to develop a 3-D distribution of cavity fluence, determine accuracy of SSTR's, and measure neutron and gamma spectra. I

I In-vessel dosimetry will include a normal, unirradiated surveillance capsule, installed in the spare surveillance capsule holder l

tube at 11.5 degrees. The capsule will contain six neutron-sensitive I elements ( RM' s ) covering incident neutron threshold energies from 0.5 eV to 2.5 MeV. I The azimuthal locations of the dosimeter strings were chosen to avoid areas of possibly large flux gradients where it is difficult to predict the neutron flux analytically. Three strings 0 have been located in the region of maximum flux (60 11, and 0 11.50) and one is in the region of minimum flux (42.5 ). The 110 string will include a beaded stainless steel chain for measurement of the axial flux gradient. Additionally, a string 0 at 26.5 is being installed. I The 75 sets of ex-vessel dosimetry and the in-vessel surveillance capsule will be irradiated for the duration of a single cycle. Following shutdown, the dosimetry will be removed and analyzed using gamma spectroscopy or fission track counting (as appropriate). The resulting data will provide measured fluence near the inside and outside vessel surfaces which will be used to ben chma rk an analytically-determined fluence distribution in the vessel. Data comparisons will be made to establish that the results are appropriately consistent. The comparisons will be made in such a way that the experimental uncertainty due to counting and analysis will be minimized, i.e. some of the dosimetry will I I o

I I be independently counted by each of the contributors in a round-robin approach. Each of the four contributing organizations, Babcock & Wilcox, Hanford Engineering Develognent Laboratory, University of Arkansas, and National Bureau of Standards, has the capability I l to analyze Radioactive Monitors and Paired Uranium Detectors. I Each supplier will be involved in the analysis of not only his own dosimetry but also the dosimetry of some (or all) of the other contributors. The B&W Owners Group will plan, coordinate, and supervise the round-robin effort. The internal comparisons that will be made include: (1) RM ' s to RM' s, (2) activation RMs to fission RMs, (3) SSTRs to RMs, (4) PUDS to RMs, (5) PUDS to SSTRs, (6) SS chain to nearby RMs, and (7) cycle 5 capsule RM results to In-Vessel RMs. I The effects of axial neutron streaming on the prediction of cavity fluence at the mid plane will be investigated by comparing benchmarked DOT-IV results for the axial plane (r,z) with benchmarked DOT-IV results for the azimuthal plane (r, ) at the mid plane. I The evaluated results of the IN-OUT Experiment will be used to iteratively develop a final analytical procedure to be used in prediction of vessel fluence from cavity dosimetry measurements (the Semiempirical Method), as shown in Figure (3). I The effect of various DOT-IV input parameters on the flux will be investigated and optimized in the model development process, I I

I I including energy group structure, angular quadrature, cross-sections, and interval size. This process will help to determine which of these parameters are important in calculation of fast flux in the cavity. The final form of the Semiempirical Method will depend on the results of the iterative procedure. It will include a calculational procedure similar to the flow chart in Figure l (4), but may also contain a set of instructions or recommendations l for the user as well as correction factors or analytical procedures for application to input or output or both. It is recognized that cavity configurations differ, but the effect of these differences will be accounted for by continued refinement of the Semiempirical Method on a case by case basis in the Plant Specific Program. Results and methodology will be be submitted as a Topical Report for NRC approval. I The Semiempirical Method is the process by which the vessel fluence can be determined using the results of analytical procedures that have been normalized to measured data. The method will provide the space-and energy-dependent neutron fluence in the pressure vessel using the two-dimensional transport code DOT-IV. The Semiempirical Method shown in Figure (4) will now be described in detail. The reader should refer to Figure (4) in reading the following. The time span over which the final vessel fluence (Block 18) is integrated must be the same time span covered by the Power History (Block 6) and the series of power distributions (Block 1). The time weighted average neutron source as a function l I I

r_ I . I of energy and position (Block 3) is calculated as follows. ! W First the time-weighted average power distribution is determined and then applied to the normalized fission spectrum. This quantity is then multiplied by the power density and other factors to yield the distributed source in neutrons per cubic centimeter per second, which is input directly to DOT-IV. The cross sections (Block 5) are macroscopic energy dependent material cross sections produced from microscopic data using standard techniques. The DOT-IV analysis will incorporate cylindrical geometry, an accurate scattering algorithm, and 4 8 scattering angles *. I The final DOT run will calculate the energy-dependent neutron flux as a function of position in the vessel and at the dosimetry location. Using the power history and the flux, the fast fluence is calculated at all locations of interest. The fluence (or reaction rate) at the dosimeter location is compared to the calculated fluence (Block 16) and a normalization factor is computed. This normalization factor along with any correction factors from the experiments will be applied to the vessel fluence (Block 19) to obtain the final normalized vessel fluence. I The Semiempirical Method, that has just been described, is applicable for all B&W 177 reactors. Some of the constants or correction

• Present plan is to use P S8 for initial runs.

3 I I

I l

I factors (Block 14 Figure 4) vill be different for different

!W reactors, due primarily to differing cavity configurations, l but the basic process is the same in each case. 3.3 Plant Snecific Program A knowledge of the accumulated vessel fluence as a function I of time is required in order to predict the vessel material degradation from some given time. The accuracy of the prediction is influenced by the n umber of points on the fluence vs. time curve. It therefore is desirable to maximize the numbe r of " measured" (fluence, time) pairs. Additionally, it is prudent to reserve a few specially prepared dosimeters for recording integrated vessel fluence over the 40 year life and preferably beyond that. These dedicated dosimeters would be important to plant life extension and because they would provide the only actual measurement of fluence for long irradiation times. Both of these objectives, frequent short term (1-2 years) fluence determinations and a few long term (4 years or more) fluence recordings could be accomplished with the employment of a plant specific Cavity Dosimetry Program. The detailed plan for installation and removal of dosimeters would be addressed individually by the licensee at a time consistent with the results of the IN-OUT experiment. l I l

I I

4. CONCLUSIONS I

The Cavity Dosimetry Program will provide the means to determine the accumulated vessel fluence based on measured data for the operating life of each plant. In combination with the results of the Integrated RVSP, the Cavity Dosimetry Program will provide the means to (1) determine the condition of each reactor vessel at any time and (2) project remaining vessel life. In short, the program makes it possible to continuously monitor the vessel fluence. I I I I I lI i

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{ I FIGURE 1 CAVITY DOSIMETRY i PHYSICAL HARDWARE DESIGN HARDWARE INSTALLATIONj ( CONSTRAINTS FOR DAVIS-BESSE PROCUREMENT OF 00SIMETRY 3 IN-CAVITY E AND/OR AND HOLDERS EQUIPMENT FABRICATION JL REQ R NTS R R ~ e REFINEMENT OF CONSISTENT SET FINAL RESULTS FINAL REVIEW OF DOSlMETER COMPARISON A BY BT,W WITH ALL + ACT. AND INTEG. H> PARTICIPANTS

3 FISSION RATES i

h POWER HISTORY DATA @ v N I I I f DOT-IV CALCULATED l' NUCLEAR DATA ANALYSES R-0 ACTIVITIES AND 'AND RZ FOR + INTEG FISSION DAVIS-BESSE (SEE FIG 3) 15 RATES 'g, 8 i au j l y a = D0T-IV EFFECTS OF CAVITY ANALY RZ FOR CONFIGURATION ON = AXlAL STREAMING CONFIGURATION AND MID-PLANE PLANT SPECIFIC = f FLUX CAVITY CONFIGURATION r. l

20 GRAM SSTRs TO HEDL AND POSSIBLY

• OTHERS I f FOLLOWING RMs TO EACH INITIAL RFSULTS IRRADIAT10N CONTRIBUTCR TO ANALYSIS BY M

OVER CYCLE 6 BE ANAL. IN A 1 BOW WITH NBS.HEDL, A I 'ROUND R02iN. AND U OF A EFFORT PUDS AND LIF ,. CHIPS TO NBS ]{ RESULTS YES SEMIEM RICAL FOR NRC PPR VAL ACCEPTABLE ANALYSIS L = METHODS (SEE INCLUDING TECH. FIGURE 4) FOR P/S APPLI NO I I I f ITERATIVE ANALYSIS Also AvailaMe On Aperture Card TI APERTURE CARD U 950t/cd/T35L lI 26.5 42.5 l 11.5 Y^Z Y --* Z 11o 6a Y/Z W -- Z W%X Y oo ( Seal Plate i g Eleva ion rt 1 .?. ,,[kh\\, f ~ . /: i. Reactor Nozzle l '<~. :.,J;ly Elevation N :,7 ' k,:.k \\ l ' 4..:, l i ,.I L BW 3 i I ..8 / Upper Active o HEDL \\ HEDL HEDLN 5 BW i / h / b B HEDL 3_ BW N g 1/4 Plate ,, / Elevation U i [ f BW HEDL l } / \\Locationof U of Arkansas BW Midcore Permanent Chain i,' Elevation Dosimetry y 3/16x6x2 t Corr. Plate I I l C-3 1 C ? I D i i ( lW.,: ;. I I ll / q. .t s i .g(:'.1, Typical Bracket h / Under Insulation Ring Angle h p,:, 1 Y\\.q,a' g (- l ^ .t 4,,

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FIGURE 2 REACTOR CAVITY 00SIMETRY ELEVATIONS _-

l FIGURE 3 DEVELOPMENT OF SEMIEMPIRICAL METHOD I USING RESULTS OF THE IN-0UT EXPERIMENT I CROSS SOURCE SECTIONS DISTRIBUTION I I Y j EX R NT 4 SEE FIG. (!) R0 AND RZ II y I CAVITY FLUX DATA PROCES$1NG AND AND ANALYSIS ACTIVITIES I ,r I CAPSULE ^ I ' CAVITY FLUX I AND ACTIVITIES U L ITERATIVE PROCEDURE CAPSULE ~ TO DEVELOP FINAL MODEL ? ACTIVITIES ,r NO RESULT OF COMPARISON PERTURBATION ? I EXPERIMENT YES I SEMIEMPIRICAL I METHOD SEE FIG. (4) I - u

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SEti! EMP IRICA (a s i X-Y POWER l i DISTRIBUTIONS CROSS FOR ALL PERIODS GEONETRY SECTIONS SINCE LAST ~' FLUENCE ANALYSIS I I f I f I f h h O3 0 7 F POWER (.;; SORRELL TIME AVERAGE DOT 4 CODE i HISTORY + + CODE DISTRIBUTED 2 RUNS NEUTRON SOURCE (BOOTSTRAPPED) r-I f FLUX AT DOSIMETER r LOCATION O Fi REMOVE DOSIMETERS r, + r, SHIP FOR Lii ANALYSIS U n U 1y [I f1 [.J i

METHOD FLUX IN ? PRESSURE l VESSEL h 'l CALCULATED FLUENCE lt ^ FLUENCE AT NORMAllZATION = PRESSURE ANALYSIS + DOSIMETER FACTOR VESSEL - LOCAT10N N O Araueje On A n tture cara HEDL OR Oil MEASURED + FLUENCE AT EqulV. DOSIMETER Trj ' '^ " APERTURE A CARD CORRECTION N FACTORS OR A!IRYS I S + FINAL PROCEDURES FROM + VESSEL FLUENCE CAVITY DOSIMETRY PROGRAM $$25Wowng-02 ~ lI l8 nefinition of Terms The following definitions are provided for clarity: Cavity Dosimetry Benchmark Experiment -- the experiment currently being set up to measure accumulated neutrons and gamma f l u e n c e lg at numerous locations outside the Davis Besse reactor vessel 3 and at one location inside the vessel. the program shown in detail in Cavity Dosimetry Program Figure (1). G12 code -- computer program which calculates macroscopic material cross sections from microscopic cross sections and number densities. another term for Cavity Dosimetry In-out Experiment Benchmark experiment. IRVSP -- Integrated Reactor Vessel Surveillance Program Idf -- Lithium Flouride gamma detector Perturbation Experiment experiment currently underway to demonstrate accuracy of analytical techniques. Plant Specific Cavity Desimetry -- the dosimetry which will be installed in the cavity of each reactor. PJ1D -- Paired Uranium Detector I RM -- Radiometric Material (activation or fission type neutron dosimeter). RVSE -- Reactor Vessel Surveillance Program SORREL code -- calculates time weighted average power distribution, models core liner, writes 8 and distributed source on tape for I use by DOT-IV. SSTR -- Solid State Track Recorder-neutron dosimeter I I I ' I -}}