ML20056G942

Forwards Minutes from 930812 Public Meeting Held to Discuss Westinghouse Owners Group Plans to Recalculate Fluence Values for Surveillance Capsules in Plants

ML20056G942
Person / Time
Issue date:	08/14/1993
From:	Mayfield M NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:	Shao L NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
Shared Package
ML20056G931	List: ML20056G930 ML20056G942
References
NUDOCS 9309070346
Download: ML20056G942 (31)

Text

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P August 14, 1993 l

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MEMORANDUM FOR:

L. C. Shao, Director Division of Engineering, RES FROM:

M. E. Mayfield, Section Leader Fracture and Irradiation Section Materials Engineering Branch

SUBJECT:

MINUTES OF NRC/NUMARC/WOG MEETING ON FLUENCE CALCULATIONS FOR SURVEILLANCE CAPSULES Enclosed are the minutes from a public meeting held on August 12, 1993, to discuss Westinghouse Owners' Group plans to recalculate fluence values for the surveillance capsules in Westinghouse plants. This effort is in response to the NRC's request for assistance in determining the fluence values for these capsules using state-of-the-art methods. The results of the effort are intended for use by the NRC in evaluating the need for changes to the embrittlement trend curves in Regulatory Guide 1.99, Revision 2.

/5 Michael Mayfield, Section Leader Fracture and Irradiation Section i

Materials Engineering Branch

Enclosure:

As stated cc:

K. Cozens, NUMARC J. Strosnider, NRR DISTRIBUTION:

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MINUTES OF MEETING BETWEEN NRC STAFF AND NUMARC/ WESTINGHOUSE OWNERS' GROUP ON l

RECALCULATION OF SURVEILLANCE CAPSULE NEUTRON FLUENCE On Thursday, August 12, 1993, members uf the NRC staff met with representatives from NUMARC, the Westinghouse Owners' Group, and members of the public. The meeting was held in the NRC's offices in Rockville, MD, and started at 1:00 p.m.

An attendance list is included as Attachment 1.

Brief introductory remarks were given by M. Mayfield (NRC/RES) and by K. Cozens (NUMARC) describing the purpose of the meeting.

Presentations were made by F. Lau and E. P. Lippincott (Westinghouse) describing the work being proposed by the Westinghouse Owners' Group to update the fluence calculations for surveillance capsules for Westinghouse plants.

This proposed work is a direct response to the staff's request for assistance in updating'these fluence values in support of a research project to evaluate the embrittlement trend curves in Regulatory Guide 1.99, Revision 2.

A copy of the handout materials is included as Attachment 2.

Additionally, Dr. Lippincott provided a written " Outline of Surveillance Capsule Fluence Evaluation," which is included as Attachment 3.

At the conclusion of the Westinghouse presentation, three questions were posed to the staff.

Specifically, the questions and staff responses were:

(1)

"Does the staff agree that what the WOG is doing and the way the calculations / comparisons are being performed is acceptable to them?"

The staff responded that, based solely on the material presented, the work appeared to provide the updated information requested during the February 1993 meeting.

The staff noted there were some questions about exactly how uncertainties would be handled, but that these did not appear to be significant problems. Based on the presentation, the staff did not identify any problems that would limit use of the results in subsequent staff evaluations of Regulatory Guide 1.99.

(2)

"Does the staff believe that these results will be usable in the RG 1.99 trend curve reevaluation?"

The staff responded that, within the limits of the presentation and the staff's interpretation of what would be provided, the results would appear to be usable in the staff's reevaluation of Regulatory Guide 1.99, Revision 2.

Assuming that the results were consistent with the staff's interpretation, the staff indicated that the results would be incorporated into the Power Reactor Embrittlement Data Base, and would then be considered in the embrittlement trend curve reevaluation.

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(3)

"Since this is a generic effort performed at the staff's request, how can the fluence results be submitted for use without specific review fees?"

The staff responded that since the results were being provided in response to a request for assistance from the Office of Nuclear Regulatory Research (RES), the Westinghouse report would be considered just like any other publicly available reference material, and review fees would not be appropriate. However, if the results were submitted l

to the NRC as part of licensing actions, the review fee issue would be considered by NRR through the normal submittal process.

It was noted by the staff that the results of the fluence calculations should be mailed to M. Mayfield directly. Mr. Mayfield would be responsible for assuring that the non-proprietary information would be submitted to the Public Document Room.

Mr. Mayfield would also be responsible for assuring that any proprietary information was handled in accordance with NRC practice and policies.

In response to a request from NUMARC, Mr. Mayfield identified Dr. F. Kam of ORNL as the appropriate technical contact if Westinghouse had any specific technical questions. Additionally, A. Taboada of the Materials Engineering Branch, RES, would be the staff contact, and any materials provided to Dr. Kam should also be provided to Mr. Taboada who will be responsible for assuring that the materials are placed in the Public Document Room.

The meeting ended at approximately 3:00 p.m.

No licensing or licensee-specific issues were discussed during the meeting.

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Mct ATTENDANCE LIST FOR MEETING ON WOG PLANS FOR SURVEILLANCE CAPSULE FLUENCE RE-EVALUATION AUGUST 12, 1993 ROCKVILLE, MD NAME AFFILIATION / ADDRESS PHONE NUMBER v/gcy Sri 7osuiam

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MEETING ON i

WOG PLANS FOR SURVEILLANCE CAPSULE FLUENCE RE-EVALUATION AUGUST I2, 1993 ROCKVILLE, MD NAME AFFILIATION / ADDRESS PHONE NUMBER the l$1sbWwz FK tt 5 -r<c,c,styiq r

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Recalculation of Surveillance Capsule Neutron FL' c.ce a

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One White Flint Offices Room 16 B11 1:00 pm August 12, 1993 i

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1 Agenda i

Meeting Objectives WOG Program Overview

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Planned Methodology j

Questions Discussion Schedule Action Items l

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WOG Objectives J

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Ensure that technical aspects of program, e.g., what and how, are i

suitable to all parties.

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Establish mechanism whereaby NRC will be able to accept and i

utilize results for trend curve development.

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WOG Program Overview Purpose Benefits

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4 Surveillance Capsule Neutron Fluence Update

Purpose:

Establish Improved Data Base of Neutron Fluence for Westinghouse

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Surveillance Capsules for use in Trend Curve Analysis Benefits:

Eliminate Excessive Scatter in the Fluence Data Provide dpa and Fluence Rate Data Define Data Uncertainties Provide Consistent and Traceable Data

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4 Planned Methodology For WOG Program Use current accepted practice Emphasis on defining uncertainties Document in WCAP for easy access

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ANALYSIS OF SURVEILLANCE CAPSUL'E FLUENCE l

l INPUT ACTION IfSI MEASURED -

ACTIVITY REFERENCE MEASURED

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LOCATION _

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SPATIAL OF THE CAPSULE GRADIEtRS _

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IRRADIATION HISTORY y

SENSOR DETERMINE SENSOR PERFORM CHECK COMPOSITIONS _

REACTION RATES FOR CONSISTENCY

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CORRECTIONS REFERENCED TO FULL WITH REACTION FOR COMPETING POWER OPERATION DATA BASE REACTIONS AND IMPURITIES _

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SENSOR ESTABLISH FLUENCE AND CROSS-SECTIONS TO THE CALCULA JION TEST SPECIMENS

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Input Data Needed for Surveillance Capsule Analysis Reactor Thermal Power History (by month) for Capsule Irradiation Reactor Geometry including Surveillance Capsule Locations Surveillance Capsule Geometry including Dosimetry Locations Measured Dosimetry Activities Dosimetry Material Specifications Fuel Power Distributions (Cycle burnups by assembly and axial power shapes)

Dosimetry Cross Sections from Evaluated Nuclear Data File Reactor Material Cross Sections from Evaluated Nuclear Data File

T Inputs to Analysis Provided by

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Neutron Transport Calculation Results used for Fluence Determination from Measurements Multigroup Neutron Spectrum at : Capsule Center Gradients in Capsule Relative Flux in Capsule at Full Power Result used to Check Fluence Derived from Measurement Average Absolute Flux w

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/e Consistency of Analysis All Capsules will be Analyzed Using Same Method Measured Data will be Compared to Ensure Consistency Fluence Values derived from Measurements will be compared with Calculated Fluence Values Comparisons will be made both with Capsules from Same Plant and from Similar Units 1

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a Uncertainty Analysis Uncertainty Components of Derived Fluence:

Reaction Rate Determination Radioactivity Measurement Uncertainty Power History Uncertainty Capsule Spatial Gradients Capsule Neutron Spectrum Dosimetry Cross Sections Uncertainty Components of Specimen Fluence:

Capsule Spatial Gradients (5-16%)

Derived Capsule Fluence Uncertainty (5-10%)

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.I NEUTRON SENSOR LOCATIONS WITHIN INTERNAL SURVEILLANCE CAPSULES i

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+10% Bias (Positive bias means reported fluence at center of capsule is higher than the specimen fluence at Charpy notch)

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yy 17 Schedule 07/1/93 Start data collection 08/12/93 Meet with NRC 08/31/93 NRC or NUMARC issue meeting minutes documenting and concurring with WOG approach and committing to use results 09/01/93 Surveillance Capsule Reanalysis begins 11/15/93 Preliminary results for review 12/31/93 Issue report to NRC

/

o Does the staff agree that what the WOG is doing and the way the calculations / comparisons are being performed is acceptable to them?

o Does the staff believe that these results will be usable in the RG 1.99 trend curve reevaluation?

o Since this is a generic effort performed at the staff's request, how can the fluence results be submitted for use without specific review fees?

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l\\{}ac.k m k Outline of Surveillance Capsule Fluence Evaluation E. P. Lippincott, Westinghouse In order to provide a consistent set of surveillance capsule neutron fluence data for Westinghouse plants, a program has been initiated to reevaluate all the capsules using a consistent methodology. This analysis will provide a set of capsule Cuences for use in the NRC l

embrittlement correlations (currently underway). The update is necessary because of the improvements in cross sections and in calculational methodology that have occurred since the I

early capsules were analyzed. Many of these early capsules were used as the basis for Reg.

Guide 1.99, Rev. 2. It is anticipated that the improved fluence consistency will improve the accuracy of the correlations and reduce the data scatter, which will ultimately result in a reduction in margin requirements. The reanalysis will not address calculation of vessel fluences and will therefore not have any direct impact on operating requirements except through the development of the new embrittlement correlation by the NRC. The description of the proposed methodology for the capsule analysis is given in this paper.

l The surveillance capsule fluences will be based on the dosimetry measurements in each capsule using a calculated neutron spectrum to relate neutron flux to dosimetry reaction rates. The fluence result will be checked by comparison with a plant specinc calculated fluence based on neutron transport. Fluence uncertainties will be derived based on measurement uncertainties, uncertainties in the fluence derivation process, and uncertainties due to the spatial extent of the l

capsule (see discussion below). A summary diagram of the Duence evaluation process is shown in the figure. A table of input data requirements is also given. Elements of the fluence evaluation process are discussed below.

8 Evaluation of dosimetry reaction rates Dosimeter activities (disintegrations per second) are measured in the laboratory and corrections made to give values at the center of the capsule (top box in the figure). These correction factors are obtained from a neutron transport calculation which adequately models the surveillance capsule. For Westinghouse capsules, all dosimetry was located close enough to the core axial i

midplane so that axial corrections are not significant. Next (second box in the Ogure), reaction l

rates are derived from the measured dosimetry decay rates using the reactor power history (thermal megawatts generated by month) for decay corrections and to adjust to full power.

Corrections to the pcmer history to obtain relative reaction rates at the capsule location vs time are necessary based en plant specific fuel management. These corrections are obtained by a cycle by cycle analysis based nn neutron transport. Measurements on individual dosimeters are corrected for impurity contributions, photo 5ssion, and burn-in, burn-out as appropriate.

Corrected measurement results are expressed as reactions per atom per second. All parameters used to derive the reaction rates will be tabulated.

capsole Analysis 7/12/93 1

8 Fluence derivation The fluence rate will be derived from the measurements using a calculated spectrum and evaluated nuclear data file (ENDF) dosimetry cross sections distributed by national data centers.

l This will be done using a least-squares computer code to properly weight the data according to l

uncertainty and to produce an estimate of the uncertainty in the final result. Measurement uncertainties will be based on estimated random and systematic uncertainties as confirmed by data consistency. Uncertainties in the cross sections will be based on evaluations from the cross section Oles. Calculational uncertainties will be assumed large in order to minimize the effect of absolute calculational value on the result. The capsule Ouence will be obtained by multiplying l

the Buence rate at full power times the effective full power seconds of irradiation.

Displacements per atom (dpa) will be derived in a similar fashion.

Unless an update is available calculated spectra will be derived using ENDF/B-IV cross sections and the dosimetry results will be evaluated using ENDF/B-V. The effect of using ENDF/B-VI cross sections will be investigated and included in the uncertainty evaluations. To the extent that ENDF/B-VI cross sections are available and tested. they may be incorporated in the analysis if time permits.

To provide a check on the least squares result, a second derivation of the fluence will be made by obtaining the fluence rate as measured by each threshold dosimeter (Duence rate = reaction rate divided by the effective reaction cross section). These results for each dosimetry reaction will then be averaged to obtain the capsule Guence rate to compare with the least-squares value.

If the two values do not agree within an expected tolerance, then the data will be checked to ensure accuracy. If necessary (when inconsistencies cannot be resolved), the uncertainty values may have to be increased.

9 Fluence calculations Fluence calculations will be carried out to: 1) derive a spectrum for the surveillance capsule for use in the fluence derivation, 2) derive relative fluence at the surveillance capsule at full power for varying fuel management for the reaction rate derivation, and 3) derive an absolute Cuence value for comparison with the measurement. The absolute Quence value will provide a check on the measurement accuracy and the evaluation process and a tabulation of the comparisons will indicate the consistency of calculational and measurement data. In the event that inconsistent measurement and calculational resolts are obtained, an investigation will be made to try to explain or resolve the discrepancy or the uncertainty on the result may be increaseo.

9 Uncertainty analysis Based on the uncertainty input to the least-squares code, the uncertainty in the measured Ouence at the capsule center is derived. A second uncertainty component will take into account variations in fluence due to the radial, azimuthal, and axial displacement of the specimens from the capsule center. These will be tabulated separately and a total uncertainty given for use in the statistical correlation. Additional contributors to uncertainty will be specifically investigated Capsule Ar.alysis 7/lF93 2

4 such as the effect of a cross section change on the calculated spectrum and the effect of changes in the ENDF dosimetry cross sections. The effect of these changes is expected to be small and well within the estimated errors.

1. Discussion The result of the fluence reanalysis will be a table of capsule fluences and dpa for the Westinghouse capsules that can be supplied to the NRC for use in correlation of the embrittlement data from these capsules. Additional documentation will also be prepared to provide traceability for the results and to allow future reevaluations to be easily conducted if indicated by future data improvements. This documentation will include all the data discussed above including as a minimum the reactor power history, reaction rate measurements and correction factors, and input for the neutron transport calculations. Consideration should also be given to regularly update this data tabulatbn to include future capsules.

Following the procedures outlined above should enable other evaluators to obtain the same fluence result within the defined uncertainty. Similarly, consistent results for non-Westinghouse plants should be obtained although additional considerations may be necessary to deal with different geometry capsules. Although consistent biases will remain between different fluence evaluations due to differences in cross section sets and other assumptions, the use of a consistent and traceable procedure will provide results that are consistent within uncertainty. In addition, as the biases become better known and the causes identified, such biases can be gradually climinated.

espsule Analysis 7/12/93 3

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ANALYSIS OF SURVEILLANCE CAPSULE FLUENCE i

INPUT ACTION TEST

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MEASURED ACTIVITY REFERENCE MEASURED SENSOR SPECIFIC ACTIVITIES

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LOCATION _

TO THE CENTER SPATIAL

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OF THE CAPSULE GRADIENTS _

V

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IRRADIATION HISTORY SENSOR DETERMINE SENSOR PERFORM CHECK COMPOSITIONS _

REACTION RATES FOR CONSISTENCY

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CORRECTIONS REFERENCED TO FULL WITH REACTION FOR COMPETING POWER OPERATION DATA BASE REACTIONS AND IMPURITIES, V

DETERMINE AVERAGE PERFORM CHECK CAICULATED~

NEUTRON FLUX AT FOR CONSISTENCY SPECTRUM THE CAPSULE CENTER WITH CAPSULE AND DATA BASES

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SENSOR ESTABLISH FLUENCE AND CROSS-SECTIONS, TO THE CALCULATION TEST SPECIMENS Capsule Analys s 7/12/93 4

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Input Data Needed for Surveillance Capsule Analysis Reactor Thermal Power History (by month) for Capsule Irradiation Reactor Geometry including Surveillance Capsule Locations Surveillance Capsule Geometry including Dosimetry Locations i

Measured Dosimetry Activities j

Dosimetry Material Specifications i

Fuel Power Distributions (Cycle burnups by assembly and axial power shapes')

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Dosimetry Cross Sections from Evaluated Nuclear Data File Reactor Material Cross Sections from Evaluated Nuclear Data File I

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v-w w

List of Surveillance Capsules

{

Domestic Westinghouse Plants Plant Cupsules Capsules Analysed by Capsules Analysed by Westinghouse Others i

Beaver V Iley i VUW 3

0 Beaver Valley 2 U

l 0

i Braidwood i U

1 0

Braidwood 2 U

1 0

Byron i U

1 0

{

Byron 2 U

1 0

Catawba 1 Z

l 0

Catawba 2 Z

l O

Comanche Peak i U

l O

Cwk1 XYTU 1

3 Cook 2 YXT 0

3 i

Callaway 1 UY 2

0

}

Conn. Yankee A F D 11 2

2 Diablo Canyon i S

I O

Diablo Canyon 2 UX 2

0 Farley I XUY 3

0 i

Farley 2 UWX 3

0 Ginna VRT 3

0 HB Robinson SVT I

2 f

Indian Point 2 TYZV O

4 Indian Point 3 TYZ 3

0 Kewaunee VRP 3

0 McGuire i OX 2

0 McGuire 2 VUX 3

0 North Anna 1 VU 1

1 North Anna 2 VU 1

1 Point Beach 1 VSRT 3

1 s

6

W.

\\

i Plant Capsules Capsules Analysed by Capsules Analysed by i

Westinghouse Others Point Beach 2 VRT 2

1 Prairie is !

VPRW 4

0

+

1 Prairie is 2 VTR 3

0 l

Salem 1 TZY 3

0 I

Salem 2 TUX 3

0 San Onofre 1 DAF i

2

?

Sequoyah 1 TUX 2

1 1

Sequoyah 2 TUX 2

1 So Texas 1 U

1 0

i So Texas 2 V

1 0

i Surry 1 TVW I

2 Surry 2 XV I

I Turkey Pt 3 TSV I

2 Turkey Pt 4 TS 0

2 Trojan UXV 3

0 Virgil Summer UVX 3

0 Vogtle 1 U

1 0

Vogtle 2 U

1 0

Wolf Creek U

1 0

Zion I TUXY I

3 Zion 2 UTY I

2 Total 81 34

.