ML18038A379

Final Rept, Nine Mile Point Unit 1 Surveillance Capsule Program, for Niagara Mohawk Power Corp

ML18038A379
Person / Time
Site:	Nine Mile Point
Issue date:	01/31/1991
From:	Michelle Manahan Battelle Memorial Institute
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ML17058A960	List: ML17058A959 ML18038A379
References
NMEL-90001, NUDOCS 9109120248
Download: ML18038A379 (639)

Text

NMEL-90001 FINALREPORT Nine Mile Point Unit 1 Surveillance Capsule qxgxvQ@.'cd(>~xi>~XPN% gx",'gQ%~%'c?gg%:,iQ 3 j<)i;,.:.pi;"jx@gi~Q'y:~%@g4<~;Kgj"j44jg0 524Ã<<. N Program To Niagara Mohawk Power Corporation January, 1991

'".:Bzmelllle Putting 7echnology To Work y q10gygP0>P8 910830 pDR ADOC)% 0 pDR p

Report Number NMEL-90001 Final Report entitled Nine Mile Point Unit 1 Surveillance Capsule Program to Niagara Mohawk Power Corporation January 4, 1991 by Dr. M. P. Manahan Battelle Penn State University 505 King Avenue Nuclear Engineering Dept.

Columbus, Ohio 43201 231 Sackett Building University Park, PA 16802

Battelle does not engage in research for advertising, sales promotion, or endorsement 'of our including raising investment capital or clients'nterests recommending investment decisions, or other publicity purposes, or for any use in litigation.

Battelle endeavors at all times to produce work of the highest quality, consistent with our contract commit-ments. However, because of the research and/or experi-mental nature of this work the client undertakes the sole responsibility for the consequences of any use, misuse, or inability to use, any information, appara-tus, process or result obtained from Battelle, and Battelle, its employees, officers, or Trustees have no legal liability for the accuracy, adequacy, or efficacy thereo'f.

TABLE OF CONTENTS Pacae

1.0 INTRODUCTION

. . . . . . . . . . ~ . . . . . . . . . . . 1 1.1 Historical Perspective . . . . . . . . . . . . . . 2 F 1.1 Revised Surveillance Program Description . 6 1.1.2 Materials Mix-Up Concern . ~ . . ~ . . ~ . . 10 1.2 Detailed=Work Scope . . . . ~ . . . ~ . ~ . . ~ . . 15 1.2.1 Re-evaluate Data and Write New Surveillance Program Description . . . . . . . . . . . . 15

..1.2.2 Determine Initial RT~~ of Beltline Materials 16

"-1:2:3 Encapsulate Archive Dosimetry and Temperature Monitor Materials . . . . . ~ ~ 17 1.3 Applicable Documents 17 MATERIAL MIX-UP ANALYSIS 22 2.1 Chemistry Analysis 22 2.2 Tensile Data Analysis 35 2.3 Charpy Data Analysis 40 2.4 Hardness Data Analysis 48 2.5 Conclusions Concerning Materials Mix-up 48 2.6 Surveillance Capsules A'nd C'aterials 49 2.7 Limiting Plate Adjustment Methodology 56 3.0 BELTLINE MATERIAL INITIAL RT~~ DETERMINATION 59 4.0 ARCHIVE DOSIMETRY AND THERMAL MONITOR ENCAPSULATION 61 5.0

SUMMARY

AND CONCLUSIONS 66 5.1 Benefits to Niagara Mohawk 67 5.1.1 PLEX Surveillance 67 5.1.2 Material Mix-Up Resolution 68 5.1.3 Limiting Beltline Material 69 5.1.4 RT~~ of Beltline Materials ~ ~ ~ ~ ~ 69 5.1.5 Significant Economic Benefits in Hydro-test 70 5.2 Future Direction 71 REFERENCES 73

.vl APPENDICES Pacae APPENDIX A DESCRIPTION OF CAPSULES A'ND C' . . . . A-2 A.l Dosimetry A-3 A.2 Temperature Monitors A-5 A.2.1 Melt Wire Temperature Monitors A-5 A.2.2 Solid State Track Recorder-Temperature Monitors A-6 A.2.2.1 Preparation of Quartz Glass SSTR-TMs A-7 A.2.2.2 Preparation of Mica SSTR-TMs A.2.2 ' SSTR-TM Fission Fragment Irradiation A-8 A.2.2.4 Fission Fragment Irradiation of Quartz Glass SSTR-TMs A-9 A.3 Charpy V-Notch Specimens A-9 A.4 Miniature and Conventional Tensile Specimens A-13 A.5 Capsule Design and Layout A-13 A.6 Archive Materials A-13 APPENDIX B APPENDIX APPENDIX C

D APPENDIX E APPENDIX F AS-BUILT DOSIMETRY DESCRIPTION FOR AS-BUILT C'-1 PHOTOGRAPHS FOR CAPSULES PHOTOGRAPHS OF ADVANCED DOSIMETRY FOR CAPSULES A'ND PHOTOGRAPHS OF MELTWIRE TEMPERATURE CHEMICAL ANALYSIS DATA FOR NINE MILE POINT UNlT 1 A'ND C-1 CAPSULES A'ND C'. C-1 MONITORS'-1 F-1 o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

F.l Plate Data F-2 F.2 Weld Data F-11 F.3 Supplementary Base Metal Chemical Analysis F-14 3.v

APPENDICES continued Pacae APPENDIX G TENSILE DATA G-1 G.1 Baseline Tensile Data G-3 G.2 Irradiated Surveillance Specimen Data G-6 G.3 Archive Plate G-8-3 Data G-9 G.4 Miniature Base Metal Specimen Data G-18 APPENDIX H CHARPY DATA H-1 H.l Unirradiated Data ~ ~ . ~ . . . . . . . H-3 H.l.l 1964 Unirradiated Baseline

'Charpy Data H-4 H.1.2 Unirradiated Archive Plate G-8-3 Data H-17 H.l.3 1990 (current study) Unirradiated Archive Plate G-8-3 Data (T-L Orientation) H-21 H.2 Irradiated Surveillance Specimen Data . H-33 H.2.1 1984 Irradiated 300 Degree Charpy 0 D ata . H-34 H.2.2 1985 Irradiated 30 Degree Charpy D ata . H-39 H.2.3 1990 (current study) Irradiated Base from HAZ for the 300 Degree Capsule H-42 APPENDIX I HARDNESS DATA APPENDIX J DROP WEIGHT DATA APPENDIX K FLUX AND FLUENCE DATA K-1 K.1 30 Degree Capsule Data K-2 K.2 300 Degree Capsule Data K-5 APPENDIX L NINE MILE POINT UNIT 1 RTq~~ DETERMINATION

U I

LIST OF FIGURES Pacae FIGURE 1-1 NINE MILE POINT CORE MIDPLANE SHOWING THE LOCATION OF THE 30 DEGREEi 120 DEGREEi AND 300-DEGREE SURVEILLANCE CAPSULES . . . . . . . 4 FIGURE 1-2 DRAWING SHOWING THE PLACEMENT OF THE SIX BELTLINE PLATES FOR NMP-1 . . . . . . . . . . 14 FIGURE 2-1 COMPARISON OF CHEMICAL ANALYSIS OF COPPER . . 27 FIGURE 2-2 COMPARISON OF CHEMICAL ANALYSIS OF NICKEL . . 28 FIGURE 2-3 COMPARISON OF CHEMICAL ANALYSIS OF PHOSPHORUS . 29 FIGURE 2-4 COMPARISON OF CHEMICAL ANALYSIS OF MANGANESE . 30 FIGURE 2-5 COMPARISON OF CHEMICAL ANALYSZS OF MOLYBDENUM . 31 FIGURE 2-6 COMPARISON OF CHEMICAL ANALYSIS OF CHROME . . 32 IGURE 2-7 NMP-1 NOZZLE DROP OUT LOCATIONS 34 FIGURE 2-8 EFFECT OF IRRADIATION ON PLATE G-8-1 AFTER A FAST FLUENCE OF 3.6 x 10" n/cm FIGURE 2-9 EFFECT OF IRRADIATION ON PLATE G-8-1 AFTER A FAST FLUENCE OF 4.78 x 10 FIGURE 2-10 OF IRRADIATION ON PLATE G-8-3 AFTER A FAST FLUENCE OF 4.78 x 10" n/cm'5 n/cm'FFECT 47

LIST OF TABLES Pacae TABLE 1-1 SURVEILLANCE CAPSULE MECHANICAL BEHAVIOR SPECIMEN ZNVENTORY FOR NINE MILE POINT UNIT 1 o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3 TABLE 1-2 CURRENT SURVEILLANCE CAPSULES AFTER REINSERTION FOR NINE MILE POINT UNIT 1 . . . 5 TABLE 1-3 NINE MILE POINT UNXT 1 LEAD FACTORS AND REVISED WITHDRAWAL SCHEDULE . . ~ . . . . . 9 TABLE 1-4 NMP-1 BELTLINE SURVEILLANCE PROGRAM TRACEABILITY 19 TABLE 2-1

SUMMARY

OF CHEMICAL ANALYSIS TEST MATRIX . . . 24 TABLE 2-2 CHEMICAL ANALYSIS SPECIMEN IDENTIFICATIONS . . 25 TABLE 2-3 LUKEN'S MEASURED CHEMISTRY OF BELTLINE PLATES FOR NMP-1 26 TABLE 2-4 BASE METAL ROOM TEMPERATURE TENSILE PROPERTIES 38 ABLE 2-5 CHANGE IN YIELD STRENGTH ANALYSIS . . . . . . 39 TABLE 2-6

SUMMARY

OF 1964 BASELINE CHARPY IMPACT INDICES (L-T ORIENTATION) . . . . . . . . . . 42 TABLE 2-7

SUMMARY

OF CHARPY IMPACT PROPERTIES FOR UNIRRADIATED BASE METAL PLATE G-8-3/G-8-4 . . 43 TABLE 2-8 '

SUMMARY

OF CHARPY IMPACT PROPERTIES FOR IRRADIATED MATERIALS FROM THE NiNE MILE POINT UNIT 1 REACTOR . . . . . . . . . . . . ~ . . . 44 TABLE 2-9 SURVEILLANCE CAPSULE B MATERIAL COMPOSITIONS . 50 TABLE 2-10 SURVEILLANCE CAPSULE A'ATERIAL COMPOSITIONS . 51 TABLE 2-11 SURVEILLANCE CAPSULE C'ATERIAL COMPOSITIONS . 53 TABLE 2-12 SURVEILLANCE DATA ADJUSTMENT FACTORS 57 TABLE 3-1 BELTLINE MATERIAL RT~~ DATA FOR NINE MILE POINT UNIT 1 . . . . . . . . . . . . . . . . . 58 TABLE 4-1 ARCHIVE RM AND HAFM NEUTRON DOSIMETRY MATERIALS e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 62

0 LIST OF TABLES continued TABLE 4-2 ARCHIVE SSTR NEUTRON DOSIMETRY MATERIALS . . . 63 TABLE 4-3 ARCHIVE MW-TM MATERIALS . . . . . . . . . . . 64 TABLE 4-4 ARCHIVE SSTR-TM MATERIALS . . . . . . . . . . 65

0 INTRODUCTION The primary purpose of the analyses reported herein is to resolve questions raised concerning a mix-up of the base metal materials used in the Nine Mile Point Unit 1 (NMP-1) surveillance program. As described in Section 2.0, it,has been conclusively demonstrated that a material mix-up did occur during fabrication of the surveillance capsule specimens. Conclusions concerning the actual materials used in the surveillance program are presented in Section 2.5. Since some of these materials were included in the re-encapsulation capsules A'nd C', a re-evaluation of the specimen inventories for the re-encapsulation capsules and the remaining B capsule was conducted. These esults are pxesented in Section 2.6. Since the surveillance program is based on materials which are not limiting from an embrittlement perspective, it was necessary to develop an adjustment methodology so that the surveillance data can be used in P-T analysis. This methodology is described in Section 2.7.

In addition to the material mix-up analysis, additional Charpy tests were conducted to determine the RT>>T of plate G-8-3.

These results are provided in Section 3.0. In Section 4.0, the capsule A'nd C'rchive dosimetry and thermal monitor storage plan is presented. Finally, summary and conclusions are presented in Section 5.0.

In response to the material mix-up, this report is intended to serve as a new baseline pressure vessel materials report.

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any of the earlier documents are in error and should only be used in the future with caution. This report contains all of the data and information needed to evaluate future surveillance capsules and prepare Pressure-Temperature (P-T) operating curves.

1.1 Historical Pers ective Three surveillance capsules were installed in the NMP-1 reactor in 1969 prior to initial operation. Two of the capsules have been removed to date. References [ST84] and [MA85a] contain the results of tests performed on the contents of these capsules.

The number and type of mechanical behavior specimens, as well as the capsule identification and location within the reactor vessel, are summarized in Table 1-1. Figure 1-1 shows the location of the surveillance capsules. Prior to the material mix-up analysis, the base metal Charpy 30 ft-lb index shift (BT3Q) of 114'F for plate G-8-3 was thought to be larger than the shift predicted by Regulatory Guide 1.99 (Revision 2) [RG1.99(2)]

by a statistically significant amount. Since plant life extension is being considered, Niagara Mohawk decided to reinsert two capsules (A'nd C'). The prime is used to designate the new capsule in the same azimuthal location as the original capsules.

The radial location of the new capsules is slightly closer to the core than the original capsules to increase the neutron flux.

The mechanical behavior specimens which were included in the new capsules are summarized in Table 1-2.

A TABLE 1-1 SURVEILLANCE CAPSULE MECHANICAL BEHAVIOR SPECIMEN INVENTORY FOR NiNE MILE POINT UNIT 1 Azimuthal Capsule Mechanical Behavior Location Date Removed Exposure S ecimens Capsule (Degrees) From Vessel (efpy) Charpy Tensile 30 1979 5.8 12 Base 3 Base 12 Weld 2 Weld 12 HAZ 3 HAZ 120 Not Removed 10 Base 3 Base (In Vessel) 8 Weld 3 Weld 9 HAZ 2 HAZ 9 APED 2 APED~~)

300 1982 7. 98 8 Base 2 Base 8 Weld 2 Weld 8 HAZ 2 HAZ (1) Inventory confirmed by observation at Battelle during disassembly.

(2) Six Charpy base metal specimens and one tensile base metal specimen tested prior to reencapsulation. Six Charpy specimens reconstituted in 1985 shortly after testing.

(3) Inventory based on capsule loading drawing supplied to Battelle by Niagara Mohawk.

(4) Full contents plus four reconstituted Charpy base metal specimens and four reconstituted Charpy weld metal specimens tested prior to reencapsulation.

(5) These specimens are either correlation monitors or specimens from another plant.

Core Pe rl p her y Stainless Steel Shroud 00 30 Degree Capsule Reactor Pressure Vessel Stainless Steel Liner 300 Degree Capsule I

I I

I 270 90 t20 Degree Capsule t80 NOT A SCALE DRAWING FIGURE 1-1 NINE MILE POINT CORE MIDPLANE SHOWING THE LOCATION OF THE 30 DEGREE r 1 2 0 DEGREE r AND 3 0 0 DEGREE SURVEILLANCE CAP SULES

TABLE 1-2 CURRENT SURVEILLANCE CAPSULES AFTER REINSERTION FOR NINE MILE POINT UNIT 1 Ca sule Capsule Tensile Contents"'harpy A' Base 0 2 Base 0 6 Base R 2 Base M 12 Weld 0 2 Weld 0 12 HAZ -0 3 HAZ -0 C'2 10 Base 8 Weld 9 HAZ 9 APED Base R 12 Weld R 12 Base U 3 Base 2 Weld 3 HAZ 2 APED 4

4 Base M Weld M 3 Base U (1) 0 = Original irradiated (untested) specimen.

M = Miniature tensile specimens machined from broken ends of tested specimen.

R = Reconstituted Charpy specimen.

U = Unirradiated specimen.

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1.1.1

~ ~ REVISED SURVEILLANCE PROGRAM DESCRIPTION A detailed description of the original surveillance program was given in reference [ST84]. Three surveillance capsules, each containing Charpy and tensile specimens, as well as dosimetry wires [iron (Fe), copper (Cu), and nickel (Ni)], were installed in the NMP-1 pressure vessel prior to initial startup. As of the date of this report, two capsules have been removed as described earlier [ST84, MA85a] .

The full contents from the C capsule, 24 Charpy specimens and 6 tensile specimens, were tested at Battelle in 1983 to determine tensile properties and reactor vessel base metal, weld metal, and heat affected zone (HAZ) Charpy impact nil-ductility transition temperatures (NDTT) [ST84] . Six Charpy base metal

~

specimens from the A capsule were also tested at Battelle to

~

confirm the 30-ft-lb shift of the base metal observed in the C capsule specimens [MA85a]. In addition, one base metal tensile specimen from the A capsule was tested in order to further benchmark the correlation between change in yield strength and shifts in the 30 ft-lb temperature.

MPR Associates, Inc. reviewed the NMP-1 data, and recommended to Niagara Mohawk that two capsules be reinserted during the 1986 refueling outage [NE85]. Battelle provided neutron transport data to MPR for calculation of the desired radial position of the reinserted capsules [MA85b]. A description of the revised surveillance program is provided in

~ <<4>>, V ~ <<p'is QJl ~+ I l4, .g I, 0

reference [MA85c] and is summarized below.

Charpy and tensile base metal, weld metal and HAZ specimens from the A capsule were returned to the reactor vessel in a new capsule which was installed at the A location (30 degrees azimuth), during the 1986 refueling outage. The six Charpy base metal specimens from the A capsule which were tested, were reconstituted prior to reinsertion. One tensile base metal specimen from the original A capsule was tested and machined into two miniature specimens and inserted in the new capsule. In addition, five temperature monitors were installed in the reinsertion capsule to determine the maximum specimen temperatures experienced during reactor operation. The radial position of the reinsertion capsule relative to the reactor vessel wall was modified to increase the neutron flux at the capsule location by a factor of 1.97 over the maximum flux at the reactor vessel shell at the 1/4 T location. This lead factor will make up for the time the specimens were out of the reactor and allow the specimens to reach the anticipated end of life fluence at the time the 1/4 T position reaches 24 EFPY.

It is recognized that there may be a neutron flux effect which could result in different mechanical behavior for the reinserted samples as compared with regions of the vessel irradiated at a lower flux. However, the practical constraint of achieving timely data requires a higher flux at the reinsertion capsules than at the vessel. It is anticipated that future research will either confirm or negate the damage rate

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hypothesis. If necessary, a mechanistic damage model should provide adequate correlation between the surveillance data and the vessel.

Twelve reconstituted full-sized Charpy base metal and weld metal specimens and miniature tensile specimens machined from the broken tensile base metal and weld metal specimens from the C capsule were reinserted in the NMP-1 reactor vessel in a second new capsule at the C location (300 degrees azimuth). In addition, twelve unirradiated Charpy base metal specimens and three unirradiated tensile base metal specimens were installed in

'the reinserti'on capsule. The unirradiated specimens were machined from the NMP-1 reactor vessel archive plate material (G-8-3) which was used in the original weld and HAZ surveillance specimen fabrication. No HAZ specimens were included in this capsule since the observed shift for the HAZ specimens tested to date is negligible, and the HAZ shift is expected to be bracketed by the base metal and weld metal shifts. The C'einsertion capsule also contains five temperature monitors. The same lead factor was used for the C'apsule as was used for the A'apsule.

The revised withdrawal schedule and the exposure for each mechanical behavior specimen set is summarized in Table 1-3. At one-half of original design vessel life (16 efpy), the original B capsule will be withdrawn and the specimens tested. These data will provide information on the current tensile properties and Charpy shift of the reactor vessel base metal, weld metal, and

TABLE 1-3 NINE MILE POINT UNIT 1 LEAD FACTORS AND'"REVISED WITHDRAWAL SCHEDULE Capsule Exposure Factor to at Withdrawal Position

~Ca C'ead sule 1 4 T

1. 97 EFPY 24 0.99 16 1.97 32,

mz.

At three-quarters of original vessel design life (24 efpy),

the reinsertion A'apsule will be withdrawn. Due to the 1.97 lead factor on this capsule, the exposure of the A'apsule specimens will be approximately equal to the vessel 1/4 T at end of life. Results of mechanical behavior tests for the A'apsule contents will provide information on end-of-life tensile properties and Charpy shift of reactor vessel base metal, weld metal, and HAZ.

At end of vessel life (32 efpy), the reinsertion C'apsule will be withdrawn. Due to the 1.97 lead factor on this capsule, the exposure of the previously irradiated and unirradiated specimens will be equal to 155% and 133% of originalC'apsule esign exposure. Results of mechanical behavior tests on the unirradiated C'apsule contents will provide information on the tensile properties and the Charpy shift of the reactor vessel base metal for a plant life extension of 133 percent. Results of tests on the previously irradiated C'apsule contents will provide information on tensile properties and Charpy shift of reactor vessel base metal and weld metal for a life extension of 155 percent.

A detailed description of the reinserted capsules is provided in Appendices A through E.

1.1.2 Materials Mix-U Concern The Niagara Mohawk Power Corporation surveillance program 10

I or NMP-1 is described in reports issued by the General Electric Company [HI69, APED]. Further details and mechanical properties are contained in references [H065a, CE64, CE65, LE64, ST64, H065b, LU85, BU85, ST84, MA87, MA85a, MA85b,]. Reference [H065a]

specified that "all base metal shall be taken from two plates (G-8-3 and G-8-4); and all weld and heat affected metal shall be taken from the weld metal between the plates (Code G-8-3 and Code G-8-4).'" The surveillance test results reported in references

[ST84] and [MA85a] were reported based on this understanding. As shown in Section 2.0, the base metal Charpy specimens were

'ctually 'fabricated from plate G-8-1. This, of course, has a significant effect on the measured Charpy shift and necessitates recalculation of many of the results reported in [ST84] and MA85a]. In addition, it was necessary to analyze the surveillance capsule (capsules A', B, and C') inventories and clearly identify the materials present in these capsules.

Reference [H065a] also lists the orientation of the Charpy and tensile specimens within the plate. It is assumed that the orientation of the specimens cut from plate G-8-1 is identical to that specified in reference [H065a]. The specification states that base metal specimens were taken from flat slabs cut parallel to both the plate surfaces at a depth of one-quarter and three-quarter plate thickness. The Charpy and tensile base metal specimens were machined with their long axes parallel to the plate rolling direction and the Charpy specimen notches were cut perpendicular to the plate surface. Both Charpy and tensile base

~ I 'A Cy h~~

etal specimens were designated longitudinal specimens.

The weld metal for the NMP-1 pressure vessel was welded in accordance with the Combustion Engineering Welding Specification SAA-33-A(3) and NA-33-A(7) using the submerged arc process

[LE64]. The Charpy weld metal specimens were machined in a direction transverse to the weld direction; thus, only the central notched section of the specimen would necessarily be composed of weld-deposited metal. Charpy specimens were taken throughout the weld section to a depth of 1-1/16 inch from the weld root. The Charpy weld metal specimens'ong axes were therefore'arallel to the plate surface, and the notches were cut perpendicular to the plate surface. The tensile weld metal specimens were composed entirely of weld metal and were obtained y machining the specimens parallel to the weld length and parallel to the plate surface.

The Charpy HAZ metal specimens were machined in a direction transverse to the weld length and parallel to the plate surface.

The axes of the notches were then cut perpendicular to the plate surface, with the notch located at the intersection of the base metal and weld deposit. The tensile HAZ metal specimens were machined transverse to the weld length and parallel to the plate surface. The joint between the base metal and weld deposit was located at the center of the tensile specimen gage length.

Analysis of the chemistry data reported in [NA87] indicates that the base metal Charpy specimens were fabricated from plate G-8-1 and the weld and HAZ base metal came from welded G-8-3 12

lates. In addition, Y. Soong of Niagara Mohawk analyzed the reference [CE65] drawing and produced the drawings shown in Figures 1-2 and 2-7. A total of five nozzle cuts were made in the beltline region: two from plate G-8-1, two from plate G-8-3 (one of these was at the G-8-3/G-8-4 interface), and one from plate G-8-4. Therefore, the surveillance material had to come from these dropouts. In order to confirm the hypothesis that the base metal Charpys are from G-8-1, chemistry and mechanical behavior tests and analyses were performed. The detailed scope of work is described in Section 1.2.

13

4 J Ceo

.2 Detailed Work Sco e A detailed description of the work undertaken in the present study is given below.

1.2.1 Re-evaluate Data and Write New Surveillance Pro ram As a result of the findings described above, much of the Reference [ST84, NA85a, and MA87] data are in error and the current P-T curves are overly conservative. Also, many of the surveillance program reference documents are in error or are misleading. To attempt to rectify this situation, a new surveillance program description has been written and included herein. This report supercedes all previous documents and ontains all of the information necessary to conduct the program in the future. Zn addition, several analyses were conducted to correct and update existing data. These include:

(1) Fit all unirradiated beltline Charpy data using the Weibull model.

(2) Clearly indicate which materials are G-8-1 and G-8-3. For specimens which are currently being irradiated, those for which there is some uncertainty regarding the material composition have been identified.

(3) Correlate tensile, Charpy, hardness, and chemistry data to demonstrate that base metal specimens were cut from Plate G-8-1 and weld and HAZ specimens were cut from G-8-3.

(4) Develop a correlation factor which relates data on Plates G-8-3 and G-8-1 with Plate G-307-4.

15

(5) Note several cautions for the future:

(a) Not all of the in-service capsule material compositions are known.

(b) The GE machined Charpy notches for the re-encapulation program may be out of specification. Several of the unirradiated Charpy specimens which were machined at GE and tested at Battelle were remachined prior to testing. The tensile specimens may exhibit bending during testing. There is experimental evidence that the GE machined specimens were not within ASTM specifications. During the test, evidence of bending which results in significant uncertainty in the yield strength determination was observed.

(6) Test reconstituted Charpy specimens from 300'apsule HAZ specimens to establish the Plate G-8-3 bTpp.

Reconstitution and testing of 6 HAZ base metal halves (Plate G-8-3) was performed. In addition, chemical analyses were performed on the base metal taken from a HAZ specimen to verify that the material is G-8-3.

These Charpy tests provide surveillance data on Plate G-8-3.

Perform tensile and hardness measurements on plates G-8-3 and G-8-1. Unirradiated tensile tests were rerun with more accurately machined specimens cut from plate G-8-3. In addition, miniature tensile specimens machined from the base portion of weld Charpy specimens were tested. These data were needed to develop the tensile correlation. Hardness measurements were made on both plate materials for the same purpose as the tensile data.

1.2.2 Determine Initial RT of Beltline

~ Materials The initial RT>>~ for the beltline plates and welds were not determined in accordance with current ASME requirements. The current rationale for using +10 F is weak. Therefore, it was necessary to obtain data and perform analyses to firmly establish 16

he initial RTNqq.

~

~ In the case of plate G-8-3, unirradiated Charpy and drop weight tests were performed to establish the RT~~

in strict conformance with the ASME code. A total of 18 unirradiated Charpy specimens oriented normal to the rolling direction were tested in the current study. For the remainder of the beltline materials, an analytical technique was used to conservatively calculate the RT>>~. The results are provided in Section 3.0 and Appendix L.

1.2.3 Enca sulate Archive Dosimetr and Tem erature Monitor Materials As discussed in Reference [MA87], the archive dosimetry and temperature monitor materials should be carefully stored until apsules A'nd C're pulled, tested, and analyzed. Ne recommend storage in evacuated quartz tubes. This work was done to ensure adequate protection from humidity and the encapsulated materials should be stored in a temperature controlled environment.

1.3 A licable Documents As mentioned earlier, this report serves as the new baseline beltline pressure vessel materials document and supersedes all previous reports'll of the relevant data needed to conduct the surveillance program have been extracted from earlier reports and included in this report. In cases where the material composition I

is different than previously thought, the correct composition is indicated and the change noted. The data tables in the appendices 17

ir J

have been correctly changed and footnoted to indicate these changes. Examination of any of"the

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earlier references must be done with caution. In order to provide traceability, a comprehensive listing of'the references which pertain to the beltline surveillance program are given in Table 1-4. Comments pertaining to the material mix-up have been provided as appropriate.

18

TABLE 1-4 NMP-1 BELTLINE VEILLANCE PROGRAM TRACEABILITY Affected By Material Reference Title Comments

[BU85] Fast Neutron Axial Pressure No Vessel Flux Analysis Vessel Calculations for Nine Mile Point Unit No. 1

[CE65] Combustion Engineering No Believed to be the latest Drawing, Inspection as-built drawing. Shows Revision No. 1, Material the beltline plate and Identification for Niagara weld positions and the Mohawk RV, Drawing E 231- nozzle cutouts.

582-1, 9/20/65

[HO65a] Surveillance Test Program Yes Specified base Charpys to for Niagara Mohawk Reactor be cut from plates G-8-3 Vessel and G-8-4. Specimens actually taken from G-8-1.

[HO65b] Fabrication Test Program Unirradiated Charpy and for Niagara Mohawk 213" tensile data on plate BWR General Electric Co. G-38.

[LE64] Welding Material No Weld Wire Qualification Qualification to Data Requirements of NAV ships 250-1500-1

[LU64] Lukens Test Certificates, No Lukens Test Certificates From Lukens Steel Company to Combustion Engineering 19

TABLE 1-4 NMP-1 BELTLINE SURVE NCE PROGRAM TRACEABILITY continued Affected By Material Reference Title Mixu Comments

[LU85] Copper Content of Reactor No Data on Cu content of Vessel Plates beltline pl'ates not reported in [LU64]

[MA85a] Examination and Evaluation Yes 30 Degree Capsule of the Nine Mile Point-Unit Analysis 1 30 Degree Azimuthal Surveillance Capsule

[MA85b] Radial Flux Profiles for No Radial flux profile data Nine Mile Point Unit 1

[MA87 ] Surveillance Capsules Yes Re-Encapsulation Report C'or Nine Mile PointA'nd Unit 1

[NMFS] NMPU-1 Final Safety Yes FSAR must be updated Analysis Report

[NMTS] Minimum Reactor Vessel Yes P-T curves must be Temperature for updated Pressurization

[ST64] Mechanical Test Report from No Unirradiated Charpy and W.A. Stone Metallurgical tensile data on beltline R & D Department plates 20

TABLE 1-4 NMP-1 BELTLINE SURVE NCE PROGRAM TRACEABILITY continued Affected By Material Reference Title Comments

[ST84] 300 Degree Capsule Yes 300 Degree Capsule Examination, Testing, and Analysis Evaluation of Irradiated Pressure Vessel Surveillance Specimens From the Nine Mile Point Nuclear Power Station 21

1l Q4S4 IW p 0 MATERIAL MIX-UP ANALYSIS This section contains the data supporting the conclusion that a material mix-up in the NMP-1 surveillance program did occur. A conclusive demonstration was made using the reference [MA87]

chemistry data. These data are reviewed in Section F 1. In addition, the mechanical behavior trends were examined to confirm the chemistry data, and these findings are discussed in Sections 2.2 2.4. Conclusions concerning the mix-up, the impact on the material inventories, and a discussion of the limiting plate adjustment methodology are given in Sections 2.5-2.7.

2.1 Chemistr Anal sis As described in reference [MA87], the method of Inductively Coupled Argon Plasma (ICAP) Spectrometry was the primary chemical analysis method used. In addition to the ICAP measurements, the atomic absorption (AA) method was performed to determine the Cu and Ni content of the samples as an independent check. The test matrix is summarized in Table 2-1.

As stated in [MA87], the primary goals of the chemical analysis task were: to determine whether the differences in the reference [ST84] and [MA85a] measurements are due to the XRF technique; to determine whether the unirradiated plate chemistry matches the Lukens data; and to determine whether there are any observable chemistry differences between the Charpy base metal 22

pecimens fabricated from the nozzle dropout (30-degree and 300-degree base metal specimens) and the specimens fabricated from other regions in the G-8-3 and G-8-4 plates (HAZ specimens and unirradiated plate).

23

TABLE 2-1.

SUMMARY

OF MICAL ANALYSIS TEST MATRIX Number of Measurements 300'apsule Unirradiated 30'apsule 300'apsule HAZ Specimen Material Base Material Base Material Base Material Method BCL WMAL WMAL WMAL WMAL ICPS (') 3 3 1 dup~ ~

1 dup 1 dup. 1 dup. 1 dup.

AA 1 1 (Cu, Ni)

Standard 1 1 Reference Material (NBS)

Colormetric (P)

Leco Combustion (C)

Combustion Titrimetric (S)

Gravimetric (Si)

(1) BCL measured Cu, Ni, P, Mo, Cr, Mn WMAL measured Cu, Ni, P, Mo, Cr, Mn, Co, V, Ti (2) dup = duplicate analysis 24

sI <<If)jap" l$

I w" 4~ii

At least three ICAP measurements per sample were made so that the uncertainty could be quantified. The WNAL measu'rements examined all of the elements on the Lukens record. In addition, one AA backup measurement per specimen type and one backup colormetric measurement for P content were made.

Four specimen categories were analyzed and the specimen identifications are summarized in Table 2-2. Chemistry data from the Lukens test certificates [LU64] and the Lukens measured Cu content from reference [LU85) is shown in Table 2-3. The results of the chemical analyses are provided in Appendix F and in "Figures-'2-1 'through 2-6.

TABLE 2-2. CHEMICAL ANALYSIS SPECIMEN IDENTIFICATIONS Base From Base Base From Base Base From Plate G-8-3 30 Degree,.Capsule 300 Degree Capsule HAZ Unirradiated ElA E42 JlL D25 Elc E7E JAM D21 EBK ElM JAE D01 E2U E1U JlT E31 E3T J1P 25

TABLE 2-3. LUKEN'S MEASURED CHEMISTRY OF BELTLINE PLATES FOR NMP-1 Plate CHEMICAL COMPOSITION wt %

Heat No. Identification Cu Ni P Mn Mo P2074 G-307-3 0.20 0.48 0.018 1.45 0.45 P2076 G-307-4"'/ 0.27 0.53 0.019 1.23 0.52 G-307-5 P2091 G-307-10 0.22 0.51 0. 018 1.43 0 '0 P2112 G-8-1 0.23 0.51 0.021 1.34 0 '5 P2130 G-8-3/ 0.18 0.56 0.012 1.16 0.47 G-8-4 (1) Limiting plate from a radiation damage perspective.

26

COPPER ICAP AA ICAP AA 0.200 O

K I-G lCAP AA ICAP AA 0.150 lCAP AA (WMAL~ (W MAL~ (

Base from base Base from base Base from HAZ Vnirradiated Unirradiated test test 0.100 30 degree 300 degree 300 degree archive plate archive plate certificate certificate G-8-3/G-8-4 6-8-1 WMAL Westinghouse Waltz Mill Analytical Laboratories BCD Battelle Columbus Division FIGURE 2-1. COMPARISON OF CHEMICAL ANALYSIS OF COPPER 27

NICK L 0.700 ICAP AA ICAP AA ICAP AA 0.600 I-z ICAP AA V ICAP AA K

G 0 0.500 Lukens Lukens 0.400 (WMAL) {WMAL) (WMAL) {WMAL) (BCD) test test Base from base Base from base Base from HAZ Unirradiated Unirradiated certificate certificate 30 degree 300 degree 300 degree archive plate archive plate 6-8-3/6-8-4 6-8-0 WMAL Westinghouse Waltz Mill Analytical Laboratories BCD Battelle Columbus Division FIGURE 2-2. COMPARISON OF CHEMICAL ANALYSIS OF NICKEL 28

PHOSPHORUS 0.040 I-z 0.030 O

(9 0.020 0.010 0 (WMAL) (WMAl ) (WMAL) (WMAL) Lukens Lukens Base from base Base from base Base from HAZ Unirradiated test test 30 degree 300 degree 300 degree archive plate certificate certificate G-8-3/G-8-4 G-8-1 WMAL Westinghouse Waltz Mill Analytical Laboratories BCD Battelle Columbus Division FIGURE 2-3. COMPARISON OF CHEMICAL ANALYSIS OF PHOSPHORUS 29

MANGANESE oo 1.400 0

O K

(9 1.200 1.000 (WMAL) (WMAL) (WMAL) (WMAL) (BCD)

Basefrombase Basefrombase BasefromHAZUnirradiated Unirradiated test test 30 degree 300 degree 300 degree archive plate archive plate certificate certificate G-8-3/G-8-4 G-8-1 WMAL Westinghouse Waltz Mill Analytical Laboratories BCD Battelle Columbus Division FIGURE 2-4. COMPARISON OF CHEMICAL ANALYSIS OF MANGANESE 30

'I MOLYBDENUM 0.500 0

l-z oK 0.450 I-z (9

0.400 (WMAL) (WMAL) (WMAL) (WMAL) (BCD) Lukens Lukens Basefrombase BasefrombaseBasefromHAZ Unirradiated Unirradiated test test 30 degree 300 degree 300 degree archive plate archive plate certificate certificate 6-8-3/G-8-4 G-8-1 WMAL Westinghouse Waltz Mill Analytical Laboratories BCD Battelle Columbus Division FIGURE 2-5. COMPARISON OF CHEMICAL ANALYSIS OF MOLYBDENUM 31

0.130 0.120 I-Z 0.310 0.100 U

0.090 0.080 (W MAL) (WMAL) (WMAL) (WMAL) (BCD) 0.070 Unirradiated Base from base Base from base Base from HAZ Unirradiated 30 degree 300 degree 300 degree archive plate archive plate WMAL Westinghouse Waltz Mill Analytical Laboratories BCD Battelle Columbus Division FIGURE 2-6. COMPARISON OF CHEMICAL ANALYSIS OF CHROME 32

t As discussed in Section. 1.1, nozzle drop out material was available from plates G-8-1, G-8-3, and G-8-4. Plates -G-8-3 and G-8-4 are from the same heat, have identical composition, and therefore behave the same from a radiation damage viewpoint.

,Examination of Figures 2-1 through 2-6 indicates that the chemistry of the base metal Charpy specimens closely matches that of the plate G-8-1 Luken's ladel analysis, whereas the chemistry of the base material from the Charpy HAZ specimens 'closely matches the.unirradiated archive G-8-3 plates and the Lukens G 3 ladel analysis. However, this observation is not consistent with reference "[H065a] which specified that all of the base metal material should be prepared from plates G-8-3 and G-8-4. A plausible explanation is that between the time the [H065a]

pecification was written and the time the specimens were fabricated, a decision was made to fabricate the base metal Charpy specimens using the G-8-1 material because the Cu content is higher in the G-8-1 material and closer to the highest Cu plate in the beltline region (G-307-4). Also, as shown in Figure 2-7, one of the G-8-3 nozzle dropouts fell on the axial weld line. lt is possible that concern over including weld and HAZ material in the base metal Charpy specimens may have influenced the decision to machine the base metal specimens from plate G-8-1.

33

I) k

~~ ~ 4 t~ lt

~ ~

~ % ~

C4

~ ~

<<P 4' '

2 Tensile Data Anal sis The irradiated and unirradiated baseline tensile data are provided in Appendix G. Also miniature tensile specimens were machined from the base metal portion of the weld specimens.

These data are provided in Appendix G as well.

Prior to the preparation of the [MA87] report, GE machined and sent tensile specimens to Battelle to measure the archive plate G-8-3 unirradiated properties (T03, T22, T23) and for use in the C'apsule (T01, T02, T21). There is experimental evidence that the GE machined specimens were not within ASTM specifications. During the test, evidence of bending was observed which would result in significant uncertainty in the yield strength determination. As a result, these tests were iscarded (T01, T22, T23) were rerun using specimens machined at Battelle (TN1, TN2, TN3). These data are presented in Appendix G 3.

~ Care should be exercised when specimens T01, T02, and T21 from capsule C're tested since these specimens were machined in the same batch at G.E. as the specimens which exhibited bending during testing.

An analysis was performed to determine the material from 1

which the surveillance program tensile specimens were machined.

Unfortunately, there is no irradiated base metal tensile specimen material available at present for chemical analysis. Therefore, the composition of the original base metal tensile specimens is unknown and must be determined in the future when the surveillance capsules are withdrawn. The only tensile specimens 35

vailable for testing at present are the HAZ tensile specimens JUD and JTU. The base metal portion of JUD was analyzed and the results are given in Appendix F. As suspected, the chemistry of this material matches that of plate G-8-3. Therefore, it is likely that all of the HAZ tensile specimens were prepared using plate G-8-3 material. The weld specimens are composed entirely of weld metal.

A summary of the room temperature tensile behavior data is given in Table 2-4. Several valuable insights can be gained by examining these data:

~ 'he "TN-1 data (plate G-8-3 tested in 1990) is consistent with the 1964 CE data for the same plate.

~ The plate G-8-1 and G-8-3 unirradiated tensile properties are essentially identical. As discussed in reference [OD85], the uncertainty in the average yield strength change (boy) for the LWR data base is about 3 3.5 ksi, with individual uncertainties as high as 10 ksi. Therefore, for the same heat of material, we would expect that values with + 3-5 ksi would indicate similar material behavior.

~ Based on chemical analysis, it is believed that all HAZ and weld CVNs were fabricated using plate G-8-3.

Therefore, specimen EJD is most likely irradiated G-8-3 material.

'I

~ Assuming EJD represents irradiated G-8-3 material, and JJA was prepared from G-8-1 material, we would expect the change in yield strength to be larger for the JJA specimen. These assumptions are consistent with the

)

data given in Table 2-4. A comparison of the bT30 using the reference [OD85] yield strength correlation and the RG1.99(2) model is shown in Table 2-5.

~ Based on these observations, a plausible interpretation of the tensile data is that the HAZ and weld (entirely

- weld metal) tensile specimens were fabricated using plate G-8-3 and the base specimens were fabricated using plate G-8-1 material. This interpretation is consistent with the way in which the Charpy specimens were fabricated. This hypothesis must be verified in the future by chemical analysis when the capsules are withdrawn.

37

TABLE 2-4. BASE METAL ROOM PERATURE TENSILE PROPERTIES Unirradiated RT Pro erties Irradiated RT Pro erties Ultimate Ultimate Yield Tensile Reduction Yield Tensile Reduction Plate Strength Strength In Area Strength Strength In Area Identification ~ksi ~ksi ~ksi G-8-1 66. 6 87.5 66 '

(CE 1964)

G-8-3 65.0 86.2 65.4 (CE 1964)

TN-1 67.4 (upper) 87.2 62.5 (Battelle 1990)

(plate G-8-3) 65.4 (lower)

JJAt" 79.2 99.7 65.7 (300'apsule 7.98 EFPY)

JDE 76. 1 96.8 66.1 (30'apsule 5.8 EFPY)

E JD<'> 71.1 93. 3 65.1 (Base from weld 300'apsule CVN)

(1) The composition of JJA and JDE are unknown and must be determined in the future when the surveillance capsule is withdrawn. It is currently thought that these specimens were machined from G-8-1 material.

(2) The were composition of EJD was not measured but fabricated using plate G-8-3.

it is believed that all HAZ and weld CVNs 38

TABLE 2-5. CHANGE IN YIELD STRENGTH ANALYSIS Irradiated Yield RGl. 99 (2)

Specimen Identification Material Identification Com Cu osition"',Cry Ni ksi Strength Model Model JJA 0.23 0 '1 12.6 92 44 EJD G-8-3"'.18 0.56 6.1 41 37 (1) Lukens ladel analysis.

(2) These are the postulated materials used to fabricate the specimens. The actual composition must be confirmed by chemical analysis.

(3) Both specimens are from the 300'apsule with a fluence of 4.78 x 10" n/cm'.

39

if,HIft pjpp

.3 Cha Data Anal sis Appendix H contains all of the Charpy data generated to date. The unirradiated 1964 data for the beltline plates (L-T orientation) was fit using the SAM McFRAC code [MC89]. The plots are given in Appendix H.1 and the indices are summarized in Table 2-6. Additional tests were conducted on plate G-8-3 using specimens with both the L-T and T-L orientation. The L-T to T-L orientation change results in an average upward shift of 24'F of the Charpy curve at the 30 ft-lb level. These data are summarized in Table 2-7.

One concern raised by the data presented in Table 2-7 is the relatively low USE exhibited by plate G-8-3 tested in the T-L rientation. ~ Analyses should be conducted in the future to determine the radiation damage effects on the

~

USE decrement.

The irradiated base metal data from the 300 and 30 degree capsules and the weld and HAZ data from the 300 degree capsule were fit using the McFRAC code, and the results are presented in Appendix H.2. As described earlier, it is believed that the base metal specimens were machined from plate G-8-1. In order to further confirm this hypothesis, six specimens were reconstituted using base metal from the 300 degree capsule HAZ Charpy specimens. These data are also presented in Appendix H.2. It is believed that these specimens were prepared using plate G-8-3 material.

The irradiated material Charpy indices and 30 ft-lb shifts 40

re summarized in Table 2-8. Plots showing the effect of irradiation on the Charpy shift are shown in Figures 2-8 through 2-10. Flux and fluence data are provided in Appendix K. These data confirm the hypothesis concerning the material mix-up. As shown in Table 2-8, the measured shift for the G-8-1 material is higher at the higher fluence. Also, the G-8-3 shift is consistent with the lower Cu content. The measured BT~~s for the G-8-3 and G-8-1 materials are within the RG1.99(2) two sigma confidence band.

Overall, the Charpy data confirm the chemistry data trends

"'which show that the surveillance capsule base metal specimens came from plate G-8-1 and the HAZ and weld specimens were fabricated using G-8-3 base metal.

41

It, TABLE 2-6.

SUMMARY

OF 1964 BASELINE CHARPY IMPACT INDICES (L-T Orientation) 30 ft-lb 50 ft-lb Plate Transition Temperature Transition Temperature Upper Shelf Energy Identification ~F ~F ~ft-1b G-307-10 -3.9 33.9 99.0<"

G-307-4 -0.5 54.9 81.5 G-307-3 -14.0 33 ' 103.2 G-8-3/G-8-4 -26.5 14.4 99.5 G-8-1 7.9 49.9 86. 7~'>>

(1) Insufficient data to determine the temperature dependence of the upper shelf. The USE was taken to be the average of the highest temperature test. These data were judged to be upper shelf based on the performance of other materials.

42

TABLE 2-7.

SUMMARY

OF CHARPY IMPACT PROPERTIES FOR UNIRRADIATED BASE METAL PLATE G-8-3/G-8-4 Orientation

'-T '(1987)

Remarks

'Archive (1987) 30 ft-lb Transition

'F) 50 ft-lb Transition Temperature Temperature (F)

-21 Upper Shelf Energy (ft-lb) 108 L-T (1964) C.E. Data (1964) -26.5 14.4 99.5 T-L (1990) Archive (1990) -0. 2 46.5 68.3 43

4 ABLE 2-8.

SUMMARY

OF CHARPY IMPACT PROP S FOR IRRADIATED MATERIALS FROM THE NI MILE POINT 1 REACTOR 30 ft-lb 50 ft-lb Upper Transition Transition Shelf Fluence Temperature Temperature Energy Material (E>1. 0 mev) (F) RG 1. 99 (2) (F) (ft-lb)

(n/cm2) SHIFT G-8-1 7.9 49.9 86. 7(2)

G-8-1(>> 4. 78 z 10~7 87.2 132.8 94 6' (300'ase)

Change 4.78 x 10" 79.3 44.0 82.9 (2)

G-8-1(') 3. 60 x 10" 1 00 (30'ase)

Change 3 . 60 x 1 0>> 55.1 37.6 50.1 G-8-3 -26.5 14.4 99.5 G-8-3(') 4.78 z 10" -15.3 22.0 -100.0(')

(300'apsule)

Change 4.78 x 10" 11.2 37.2 7.6 (1) Base material from the 300 and 30 degree capsules believed to be G-8-1 material based on chemistry and mechanical property trends. Base from weld or HAZ believed to be G-8-3 material.

(2) The uncertainty in the pre- and post-irradiation USE data is high. Therefore, no conclusion can be drawn at this time concerning the USE decrement.

(3) Insufficient data to determine the USE.

44

NINE MILE POINT UNIT 1 PLATE Q-8-1 SHIFT AT 5,6X10+a1?(H/Clio+2) 120 UNIRRADIATED 100 I EXP. DATA I- IRRADIATED LL 80 EXP. DATA WEIBULL FIT Q TRANSITION 60 HYPERBOLIC

~ (+w TANGENT FIT 40 WEIBULL FIT Q A TRANSITION 20 HYPERBOLIC TANGENT FIT

-100 -50 0 50 'I 00 'I 50 200 250 TEST TEMPERATURE (F)

Figure 2-8 Effect of Irradiation on Plate G-8-1 After a Fast Fluence of 3.6 x 10'/cm 45

NINE MILE POINT UNIT PLATE C-8-1 SHIFT AT 4.8X10+a17(N/CMae2)

I EXP. DATA 120 IRRADIATED Cl EXP. DATA 100 I

I- WEIBULL FIT LL 80 TRANSITION CI WEIBULL FIT 60 UPPER SHELF HYPERBOLIC 40 TANGENT FIT V

20 QfEIBULL FIT TRANSITION 0 HYPERBOLIC

-50 0 50 100 150 200 250 300 350 TANGENT FIT TEST TEMPERATURE (F)

Figure 2-9 Effect of Irradiation on Plate G-8-1 After a Fast Fluence of 4.78 x 10" n/cm'6

NINE MILE PQINT UNIT 1 e 'NIRRADIATED PLATE Q-S-S SHIFT AT 4.8X100017(N/Cllao2)

EXP. DATA 120 IRRADIATED EXP. DATA 100 I

I- WEIBULL FIT LL 80 TRANSITION gj e WEIBULL FIT 60 UPPER SHELF HYPERBOLIC 40 TANGENT FIT V 0 g.'g 20 oaooooooooooooooo Q(EIBUU FIT TRANSITION 0 .---- HYPERBOLIC

-100 -50 0 50 100 'I50 200 250 TANGENT FIT TEST TEMPERATURE (F)

Figure 2-10 Effect of irradiation on Plate G-8-3 After a Fast Fluence of 4.78 x 10" n/cm'7

"i ill 4

2. 4 HARDNESS DATA ANALYSIS N

Rockwell B and C hardness tests were performed on the unirradiated plate G-8-3 material and Charpy specimens from the 300 degree capsule. These data are provided in Appendix I. The Rockwell C tests were performed in, addition to the Rockwell B tests because the data fell high on r the Rockwell B scale. Future tests on more highly irradiated material may require the use of the Rockwell C scale. The broken Charpy specimens were indented after fracture testing. Specimens which exhibited little plastic deformation during Charpy testing were chosen so that the specimen was properly supported in the hardness test machine.

The specimens were tested on the surface containing the notch and on the surfaces normal to the notch. A slight orientation effect was observed. Overall, the orientation effect s not significant. The G-8-3 unirradiated material exhibited an average Rockwell C hardness of 10 ' and Rockwell B hardness of 89.9. The G-8-3 specimens irradiated to 4.78 x 10'/cm had an average Rockwell C hardness of 13.5 compared to an average of

,. 15.0 for the G-8-1 material, and an average Rockwell B hardness, of 91.8 for the G-8-3 material compared to 92.5 for the G-8-1 material. Therefore, the hardness data confirms the material mix-up hypothesis.

2.5 CONCLUSION

S CONCERNING MATERIAL MIX-UP Based on a careful examination of all of the NMP-1 surveillance data, the following conclusions have been made:

~ A material mix-up did occur in the NMP-1 surveillance program. This conclusion is based on definitive 48

chemical analysis data and further substantiated by examining the mechanical behavior trends.

The base metal Charpy specimens were prepared from plate G-8-1 material and the HAZ and weld specimens were prepared using G-8-3 material.

The HAZ tensile specimens were prepared using G-8-3 plate material. It is likely that the weld (entirely weld) tensile specimens were produced using the G-8-3 material. Based on the mechanical behavior trend, it is likely that the base metal tensile specimens were prepared from G-8-1 material. Additional analyses will be needed in the future to confirm the tensile materials.

~ When the capsules are pulled in the future, chemical analyses should be done on the following materials to confirm the base metal composition:

base metal from Charpy weld for the 30'nd 300'apsules, base, base from weld, and base from HAZ for Charpys in the 120'apsule, all tensile base metal specimens.

2.6 SURVEILLANCE CAPSULES A'nd C'ATERIALS Based on the conclusions drawn in Section 2.5, the current surveillance capsule inventories were reassessed and the material compositions indicated. These data are given in Tables 2-9 through 2-11 ~ As stated earlier, some specimen compositions must be confirmed in the future.

49

0 TABLE 2-9. SURVEILLANCE CAPSULE B MATERIAL COMPOSITIONS~i Specimen Specimen Base Metal

" Identification ~Te Plate Material All Base Metal Specimens Charpy G-8-1 Tensile G-8-1 All Weld Specimens Charpy G-8-3 Tensile n/a All HAZ Specimens Charpy G-8-3 Tensile G-8-3 These materials should be confirmed by chemical analysis I'1) when the capsule is pulled.

50

TABLE 2-10. SURVEILLANCE CAPSULE A'ATERIAL COMPOSITIONS Base Metal S ecimen Identification S ecimen T e Plate Material E71A Reconstituted G-8-1 Charpy Base E12 Charpy Base G-8-1 E31A Reconstituted G-8-1 Charpy Base E2E Charpy Base G-8-1 E2T Charpy Base G-8-1 E2UA Reconstituted G-8-1 Charpy Base E17 Charpy Base G-8-1 E1AA Reconstituted G-8-1 Charpy Base E2Y Charpy Base G-8-1 E1CA Reconstituted G-8-1 Charpy Base E1D Charpy Base G-8-1 EBKA Reconstituted G-8-1 Charpy Base ED1 Charpy Weld G-8-3 ED2 Charpy Weld G-8-3 ED3 Charpy Weld G-8-3 ED4 Charpy Weld G-8-3 ED5 Charpy Weld G-8-3 ED6 Charpy Weld G-8-3 ED7 Charpy Weld G-8-3 EDA Charpy Weld G-8-3 EDB Chazpy Weld G-8-3 EDC Charpy Weld G-8-3 51

E TABLE 2-10. SURVEILLANCE CAPSULE A'ATERIAL COMPOSITIONS (Continued)

Base Metal S ecimen Identification S ecimen T e Plate Material EDD Charpy Weld G-8-3 EDE Charpy Weld G-8-3 Charpy HAZ G-8-3 J12 Charpy HAZ G-8-3 J13 Charpy HAZ G-8-3 J14 Charpy HAZ G-8-3 J15 Charpy HAZ G-8-3 Charpy HAZ G-8-3 J17 Charpy HAZ G-8-3 J1A Charpy HAZ G-8-3 J1B Charpy HAZ G-8-3 J1C Charpy HAZ G-8-3 J1D Charpy HAZ G-8-3 J1E Charpy HAZ G-8-3 JD1 Tensile Base G-8-1 JD2 Tensile Base G-8-1 Tensile Base G-8-1 Tensile Base G-8-1 JLK Tensile Weld n/a JL2 Tensile Weld n/a JTA Tensile HAZ G-8-3 JUL Tensile HAZ G-8-3 Tensile HAZ G-8-3 (1) The base metal tensile and base from Charpy weld should be confirmed by chemical analysis when the capsule is pulled.

52

TABLE 2-11. SURVEILLANCE CAPSULE C'ATERIAL COMPOSITIONS Base Metal S ecimen Identification S ecimen T e Plate Material NC01 Charpy Base G-8-3 NC21 Charpy Base G-8-3 NC02 Charpy Base G-8-3 NC22 Charpy Base G-8-3 NC03 Charpy Base G-8-3 NC23 Charpy Base G-8-3 NC04 Charpy Base G-8-3 NC24 Charpy Base G-8-3 NC05 Charpy Base G-8-3 NC25 Charpy Base G-8-3 NC06 Charpy Base G-8-3 NC26 Charpy Base G-8-3 El JA Reconstituted G-8-1 Charpy Base El JB Reconstituted G-8-1 Charpy Base E1KA Reconstituted G-8-1 Charpy Base E1KB Reconstituted G-8-1 Charpy Base EA5A Reconstituted G-8-1 Charpy Base EA5B Reconstituted G-8-1 Charpy Base E42A Reconstituted G-8-1 Charpy Base E1MA Reconstituted G-8-1 Charpy Base 53

"J TABLE 2-11. 'SURVEILLANCE CAPSULE C'ATERIAL COMPOSITIONS (Continued)

Base Metal S ecimen Identification S ecimen T e Plate Material E1UA Reconstituted G-8-1 Charpy Base E3TA Reconstituted G-8-1 Charpy Base E7EA Reconstituted G-8-1 Charpy Base J2CB Reconstituted G-8-3 Charpy Base EDKA Reconstituted n/a Weld

,, EDLA Reconstituted n/a Weld EDMA Reconstituted n/a Weld E JTA Reconstituted n/a Weld JAEA Reconstituted n/a Weld Reconstituted n/a Weld Reconstituted n/a Weld Reconstituted n/a Weld J1MA Reconstituted n/a Weld J1PA Reconstituted n/a Weld J1TA Reconstituted n/a Weld Jl JA Reconstituted n/a Weld T01 Tensile Base G-8-3 54

P l

TABLE 2-11. SURVEILLANCE CAPSULE C'ATERIAL COMPOSITIONS (Continued)

Base Metal S ecimen Identification S ecimen T e Plate Material T02 Tensile Base G-8-3 T21 Tensile Base G-8-3 Tensile Base G-8-1 Tensile Base G-8-1 Tensile Base G-8-1 Tensile Base G-8-1 Tensile Weld n/a Tensile Weld n/a Tensile Weld n/a 10 Tensile Weld n/a (1) The base metal tensile and base from Charpy weld should be confirmed by chemical analyses when the capsule is pulled.

55

7 LZMZTZNG PLATE ADJUSTMENT METHODOLOGY As described earlier, the NMP-1 base metal surveillance program consists of irradiation and testing of Charpy specimens fabricated from both G-8.-1 and G-8-3 plate materials. The limiting plate, from a radiation damage perspective, is that plate with the highest copper and nickel content. As shown in Table 2-3, the plate G-307-4 has the chemistry which results in the largest RG 1 99(2) chemistry factor. As with many plants in

~

operation today, the surveillance material is not the limiting material. Therefore, in order to be able to use the surveillance data in P-T curve calculations, we are recommending the use of a "correction factor" which adjusts for the chemistry differences.

Zn essence, two factors are developed, (one for G-8-1 material and one for G-8-3 material) and used to map the surveillance data into plate G-307-4 equivalent data. The RG 1.99(2) chemistry factor data base was used to provide the correction terms. The results are given in Table 2-12. The adjustment factor derived is as follows:

QZ 30 g(0.28 - o.3. og E) [gpG-307-4 gag)) +

[ 30

where, f= fast fluence in units of 10" (n/cm )

CF = RG1.99(2) chemistry factor bT~30 measured 30 ft-lb shift j = G-8-1 or G-8-3 material

TABLE 2-12. SURVEILLANCE DATA ADJUSTMENT FACTORS Adjustment Factor"'o Material E Obtain Plate G-307-4 uivalent Data ~ncm'easured Fluence Shift F Equivalent Plate G-307-4 Shift F G-8-1 FF (19.9) + hT3 4.78 x 10>> 79 3

~ 85.0 G-8-1 FF (19. 9) + b,T 3. 6Q x 10~7 55.1 60 '

G-8-3 FF (43.7) + bT3 4. 78 x 10" 11.2 23.7 (1) FF = f' '~ ', f= fast fluence

. 10'n/cm')

57

TABLE 3-1 Beltline Material RT>>~ Data for Nine Mile Point Unit 1 RTggg (T L) a, Plate OF OF G-8-3/G-8-4 3 ll)

G-8-1 36 G-307-3 28 G-307-4 40 G-307-10 20 RTgpg Gz Plate oF oF N5214/5G13F -50 17 86054B/4E5F -50 17 1248/4K13F -50 17 1248/4M2F -50 17 (1) Measured in accordance with ASME code.

58

~ II 'hl '

3.0 BELTLINE MATERIAL INITIAL RT T DETERMINATION An analysis was performed to establish best estimate values for the RT>>T and 6, terms used in Regulatory Guide 1.99 (Rev. 2) calculations for the Nine Mile Point Unit 1 (NMP1) beltline plate and weld materials. With the exception of plate G-8-3, sufficient data are not available to determine the RT>>T in strict conformance with the current ASME code rules. In the case of the weld metals, only three unirradiated Charpy tests were conducted at 10'F Charpy transition behavior data are available for the

~

plate materials, however, there are no drop weight data (for all of the beltline materials except for plate G-8-3/G-8-4). The methodology used is described in detail in Appendix L. The results of these calculations are presented in Table 3-1.

As described in reference [MA87], the plate G-8-3 RT>>, could not be determined in accordance with the current ASME code requirements because the Charpy specimens tested at that time had an L-T orientation. Therefore, Charpy (Appendix H) and drop weight (Appendix J) tests were performed to firmly establish the initial RT>>T for plate G-8-3. Specimens machined from plate G 3 with the T-L orientation were tested as part of the current work. These data are presented in Appendices H.1.2 and H.1.3 and the results are summarized in Table 2-7. The L-T to T-L orientation change results'n an upward shift of the Charpy curve of 21 F. This is consistent with data in the literature [EP82].

The Charpy and drop weight data are summarized in Appendices H and J, respectively. The NDT determined by drop weight tests for 59

plate G-8-3 is -25 F. Three Charpy specimens with the T-L

~

orientation exhibited 50 ft.-lbs. of absorbed energy at 57 F.

~ ~

Therefore, the RT>>~ for plate G-8-3 is -3 F.

60

.0 ARCHIVE'OSIMETRY AND THERMAL MONITOR ENCAPSULATION In surveillance capsule irradiations, it is prudent to store dosimetry and temperature monitor materials in case it is necessary to resolve conflicts in the data obtained. With these materials, calibration experiments can be repeated; and, if necessary, neutron benchmark irradiations can be conducted.

Should definitive follow-on experiments of this type be needed, the availability of these archive materials can be a crucial factor in the success of the dosimetry analysis. To this end, archive materials, identical to those used in Capsules A'nd C',

have been provided to Niagara Mohawk. Tables 4-1 through 4-4 summarize the supplied archive materials for RM neutron d osimeters, SSTR neutron dosimeters, MW-TMs and SSTR-TMs, espectively. These materials were sealed inside an evacuated quartz tube and sent to Niagara Mohawk [MA89].

61

TABLE 4-1. ARCHIVE RM AND HAFM NEUTRON DOSIMETRY MATERIALS Dosimeter Material Type PO No. Batch Vendor Form Quantity Fe 07448 26/17944 MRC 0.020D Wire 2 inches Ni SE Roll 2 Semi Element 0 '20D Wire 2 inches Cu RM, HAFM 19047 CPI 3054 Cominco Am. 0 '20D Wire - 2 inches Co/Al 44451 SC Bar 26 Sigmund Cohn 0.020D Wire 2 inches 19046 Cat 614 Reactor Exp. 0.020D Wire 2 inches Al 19045 SE Roll 1 Semi Element 0.020D Wire 2 inches NOTES Archive samples of the vanadium encapsulated fissionable RM dosimeters are not provided; those included in Capsule Sets C're to be used for this purpose. A'nd (2) Samples of the Be, Fe, and Ni HAFM dosimeters are not provided.

62

TABLE 4-2. ARCHIVE SSTR NEUTRON DOSIMETRY MATERIALS SSTR De osit

SSTR Mass TYpe ID Diameter Isotope ID Total (pg) Density (pg/cm')

Mica 219 0. 168 235-U 219 0. 826 8.81 Mica 231 0. 168 238-U 231 7. 65 81.6 Mica 199 0.168 237-Np 199 10.39 110.8 (1) Three additional mica SSTRs of 0 '68 inch diameter have been supplied as control samples to be stored along with the archive SSTR dosimeters 219, 231, and 199.

These control SSTRs are unnumbered and are not in contact with any deposit.

(2) The outer aluminum foil of the archive SSTR packages are marked. as follows:

U-235 is labeled 5; NP-237 is labeled 7; U-238 is labeled 8.

63

TABLE 4-3. ARCHIVE-MW-TM MATERIALS Melting Temperature Length Diameter Composition, Wt% (F) (in. ) (in. ) Source 80 Au, 20 Sn 536 0.25 0.030 Indium Corp. of America 90 Pb, 5 Ag, 5 Sn 558 0.25 0.090 Babcock & Wilcox 97.5 Pb, 2.5 Ag 580 0.25 0.084 Babcock & Wilcox 97.5 Pb, 1.5 Ag, 1.0 Sn 588 0.25 0.084 Babcock & Wilcox 98.8 Cd, 1.2 CU 498 0.25 0 '83 Babcock & Wilcox

. 64

TABLE 4-4. ARCHIVE SSTR-TM MATERIALS K

Material Quantity Description and Purposes India Ruby Muscovite Mica Pre-annealed and pre-etched disks 15/16 in.

in diameter and about 0.004 in. thick.

Representative material for SSTR-TM and SSTR dosimetry.

India Ruby Muscovite Mica C1 and C2 as described in Table 4-2. Needed as standards when mica SSTR-TMs are removed from the reactor.

upra II Quartz Glass 3/8 in. thick squares about 40 mils thick.

Representative materials used in SSTR-TMs ~

Supra II Quartz Glass C and 3C are 3/8 in.

squares about 40 mils thick. They are described in Table 4-2.

Needed as standards when the Supra II quartz glass is removed from the reactor.

65

J <<.~

.0

SUMMARY

AND CONCLUSIONS Based on careful examination of all data available, it has been concluded that a materials mix-up has occurred in the NMP-1 surveillance program. For the capsules which have been pulled to date, it has been conclusively demonstrated that the base metal Charpy specimens were fabricated from plate G-8-1, and not plate G-8-3 as originally specified. The base metal portion of Charpy weld and HAZ specimens is composed of plate G-8-3 material.

Future chemical analysis of capsule B materials and all base metal tensile specimens are needed to determine the composition

" of thesematerials. It is likely that the base metal tensile specimens were fabricated from plate G-8-1, and the same approach to specimen fabrication for capsule B was used as that used for apsules A and C.

As a result of these findings, this report was prepared and is intended to serve as the new baseline surveillance document for NMP-1. All of the data needed to conduct. the surveillance program is contained herein. Should future analysis require examination of earlier reports, these studies should be conducted with caution since the earlier reports are known to contain errors related to the material mixup.

In response to these findings, it is recommended that new P-T curves be prepared. The current P-T curves are overly conservative since the measured shift was determined using G-8-1 irradiated material and G-8-3 unirradiated material. The actual measured shifts, which have been corrected to account for the 66

Sgp

~"

C Ill.- s+ hl

aterial mix-up, are consistent with the [RG 1. 99 (2) ] model.

Therefore, it is recommended that Niagara Mohawk prepare new P-T curves to replace the current overly conservative curves.

5.1 Benefits to Nia ara Mohawk A substantial amount of work has been performed under the NMP-1 surveillance program since removal of the A and C capsules.

Capsule A was removed in 1979 after a vessel exposure of 5.8 effective full power years (efpy) and capsule C was removed in 1982 after a vessel exposure of 8 efpy. The culmination of this

" work is reported"herein and has resulted in the firm re-establishment of the NMP-1 surveillance program. A brief summary of the past problems, resolution of the problems, and the enefits of this work to Niagara Mohawk are presented below.

5.1.1 PLEX Surveillance The capsule C (at 300 azimuth) analysis showed a base metal BT3Q of 114 F at a fast fluence of 4.78 x 10"n/cm . In order to confirm this finding, the 30 degree capsule was analyzed and a base betal bT>> of 90 F =was measured at a fast fluence of 3.6 x 10'/cm . These data exceeded the RG 1.99(2) prediction by over 3 standard deviations.

In response to the concern raised by these findings, a surveillance capsule re-insertion program was undertaken. Two new capsules containing both irradiated and unirradiated (from archive) specimens were re-inserted. These are the first

J

,t l

apsules ever designed specifically to generate plant life extension data. These capsules are advanced BWR capsules and contain advanced dosimetry, temperature monitors, and innovative mechanical test specimens. The benefit to Niagara Mohawk will be end-of-license and life extension data'hese data can be used in plant specific damage models to yield accurate estimates of the K,~ curve shift for P-T calculations.

5.1.2 Material Mix-U Resolution Measurement and careful study of the material chemistry data "and as-'builtdrawings led to the discovery that a material mix-up had occurred in the NMP-1 surveillance program. The mechanical behavior trends were examined and found to confirm the chemistry ata. As a result, the plate G-8-3 (Cu = .18, Ni = .56) measured shift (hT>>), originally thought to be 114 F, was correctly established to be to 11 F at a fluence of 4.78 x 10 n/cm . Since the surveillance program is irradiating two plate materials (G 3 and G-8-1), the Charpy bT,~ can be determined for plate G-8-1 (Cu = .23, Ni = .51) as well. For plate G-8-1, bT~, = 79 F at a fluence of 4.78 x 10'/cm and ET3Q 55 F at a fluence of 3.6 x 10"n/cm . Therefore, the resolution of the material mix-up has resulted in three measured Charpy bT~,s instead of two, and the measured shifts are much lower than believed earlier.

68

.1.3

~ ~ Limitin Beltline Material As with many plants in operation today, the NMP-1 surveillance material is not the limiting beltline material ~

Based on the chemistry data and analysis of initial RT>>~ for the beltline materials, plate G-307-4 is the limiting material.

Therefore, in order to enable use of the surveillance data in P-T curve calculations, a correction factor which adjusts for the chemistry differences between G-307-4 and the surveillance materials (G-8-3 and G-8-1) has been developed. This approach allows Niagara Mohawk to use plant-specific data to determine the "ART>>~ 't'rend curve in accordance with the guidance provided in RG

1. 99 (2) .

~ 1' RT of Beltline Materials The RT>>~ of the beltline materials, with the exception of plate G-8-3, cannot be measured in conformance with the ASME code requirements. A value of +10 F has been assumed. Should the NRC decide to question the assumption, Niagara Mohawk would be vulnerable. However, a rigorous statistical method which is consistent with the intent of the ASME code was used to determine the RT>>~ of all of the beltline materials. Xn the case of the welds, it has been demonstrated that the +10 F assumption is overly conservative, and for the plates, the +10 F assumption is non-conservative. Although the RT>>~ for the limiting plate (G-307-4) was found to be 40 F, the analaysis enables identification of the actual limiting material Following the RG 1.99(2)

~

69

V 'j.'

~ Jl,

approach, the weld materials would have been limiting since the regulatory guide requires the use of .35% Cu and 1.0% Ni in cases where measured chemistry data are not available. This chemistry assumption leads to a chemistry factor of 272 for the weld material as apposed to 174 for the limiting plate. Therefore, weld material would have been limiting and the ARTgT for the weld would have been larger than that for the plate if the RTND~

analysis had not been done.

5.1.5 Si nificant Economic Benefits in H dro-test

'"The hydro-test temperature has been reduced significantly

(-20F) as evidenced by the new P-T curve which was developed using Charpy shift data appropriately corrected for the material mixup, [MA91]. The technical specitication defines the cold shutdown condition as the reactor coolant temperature being less than (or equal to) 212'F and the hot shutdown condition is defined by a temperature greater than 212'F. Hydro-test in the hot shutdown condition requiers longer time to warm up the water and, in addition, hot shutdown must be scheduled as a critical path activity. Hydrotesting in the hot shutdown condition requires that the ECCS system must be operational, the safety system surveillance must be complete and the primary containment must be isolated. In the hot shutdown condition, leak detection is more difficult and the inspection conditions are more severe for personnnel. On the other hand, cold shutdown requires less time because the hydro-test can be done in parallel with other 70

I'h tf~

utage activities. Saving one day or even one half day of outage time results in a significant economic benefit to NMP-1.

5.2 Future Direction Niagara Mohawk should consider several actions for the future. The general direction should be to establish a strong plant specific surveill'ance program and avoid the use of overly conservative "generic" trend curves. If the NMP-1 and Oyster Creek data were shared, there would be sufficient data available to develop a plant specific trend curve. This approach can be

'implemented within the guidelines of RG 1.99(2) since the regulatory guide allows the adjustment of the hRT>>~ analytical model in cases where two or more credible surveillance data are vailable. Therefore, it is recommended that NMPC establish a cooperative program with GPU Nuclear as soon as possible.

With regard to the surveillance capsules, the B capsule should be pulled as soon as possible and tested. Serious consideration should be given to re-inserting a capsule which contains advanced dosimetry, temperature monitors, irradiated and unirradiated mechanical property specimens, and unirradiated fracture toughness specimen blanks. Fracture toughness specimen blanks were recently placed in the Oyster Creek re-insertion capsule. It would be best if the new capsule B target fluences were compatible with the Oyster Creek fluences.

In order to be certain of future regulatory acceptance, it is recommended that the limiting plate, adjustment factor be used 71

1

'n the next P-T curve update. It is essential that this approach be accepted by the NRC since it is important to establishing a plant-specific surveillance program. It is also suggested that work be undertaken to combine the NMP-1 and Oyster Creek surveillance programs and the existing data be used to establish a plant specific trend curve. It is likely that the current overly conservative model assumptions can be relaxed and future hydrotest and P-T operating windows opened. The NRC submittals should be coordinated and similarities of the two programs noted.

A submittal which requests NRC approval for combining the data

'ases shou'ld 'be prepared in the near future. An overall plan for the combined program which optimizes the withdrawal schedule should be prepared.

72

I 6 0

~ REFERENCES

[APED] "Modified Surveillance Program for General Electric BWR Pressure Vessel Steels", General Electric Report APED-5490, April, 1967.

[ASTM261] "Standard Method for Measuring Neutron Flux, Fluence, and Spectra by Radioactiviation Techniques", ASTM Designation E261-77, Annual Book of ASTM Standards, Part 45 (1982), pp 930-941.

[ASTM263] "Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Iron", ASTM Designation E263-82, Annual Book of ASTM Standards, Part 45 (1982),

pp 951-956

[ASTM482] Guide

'Standard for Application of Neutron Transport Methods for Reactor Vessel Surveillance", ASTM Designation E482-82, Annual Book of ASTM Standards, Part 45 (1982), pp 1088-1092.

[ASTM522] "Standard Method for Calibration of Germanium Detectors for Measurement of Gamma-Ray Emission of Radionuclides", ASTM Designation E522-78, Annual Book of ASTM Standards, Part 45 (1982), pp 1139-1144.

[ASTM523] "Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Copper", ASTM Designation E523-82, Annual Book of ASTM Stnadrads, Part 45 (1982),

pp 1145-1152.

[BU85] Burns, L. S., Rogers, D. R., "Fast Neutron Axial Pressure Vessel Calculations for Nine Mile Point Unit No. 1", October 28, 1985 BUGLE 80 Cou led 47 Neutron 20 Gamma-Ra

'BUG75]

P3 Cross Section Librar for LWR Shieldin Calculations, RSIC Library DLC-75.

[CE65] Combustion Engineering Drawing, Inspection Revision No.

1, Material Identification for Niagara Mohawk RV, Drawing E 231-582-1, 9/20/65.

[CE90] "Niagara Mohawk Power Corporation Nine Mile Point Unit 1 Reactor Vessel Weld Materials", Report No. 86390-MCC-001, ABB Combustion Engineering Nuclear Power Combustion Engineering, Znc., Windsor, Connecticut, June, 1990.

73

[DOT75] RSIC Computer Code Collection, DOT 4.3 One- and Two-Dimensional Trans ort Code S stem, Radiation Shielding Information Center, Oak Ridge National Laboratory, Oak Ridge, Tennessee, November 17, 1975.

[EP82] "Nuclear Reactor Vessel Surveillance Data Base", EPRI NP-2428, June 1982

[GO79] Gold, R., "Process for Measuring Temperature with Solid State Track Recorders", U.S. Letters Patent 4,167,109, September ll, 1979.

[HI69] Higgins, J. P. and Brandt, F. A., "Mechanical Property Surveillance of General Electric BWR Vessels", General Electric Report NEDO-10115 (July, 1969).

[HO65a] Howard, A., "Surveillance Test Program for Niagara Mohawk Reactor Vessel", Combustion Engineering, Contract 164, Rev. 2, 1965.

[HO65b] 'Howard, "D.A., Fabrication Test Program for Niagara Mohawk 213" BWR General Electric Company, DE A.

Howard Combustion Engineering, December, 1965.

[LE64] Lewis, S.R., Welding Material Qualification to Requirements of NAV Ships 250-1500-1, Metallurgical R &

D, Combustion Engineering, Sept ~ 1964 Feb. 1965.

[LO84] Lowry, L. M., et al, "Examination, Testing, and Evaluation of Irradiated Pressure Vessel Surveillance Specimens from the Monticello Nuclear Generating Plant", Final Report from Battelle-Columbus to Northern States Power Company (March 15, 1984).

[LO84 ] Private Communications, John Conway of Niagara Mohawk Power- Corporation to L. M. Lowry of Battelle's Columbus Laboratories, March 24, 1984 and May 2, 1984.

[LU64] Lukens Test Certificates, from Lukens Steel Company to Combustion Engineering, May-July, 1964.

[LU85] Letter to Mr. Tom Caine from J. Fredric Longenecker, "Copper Content of Reactor Vessel Plates", October 9, 1985.

[MA85a] Manahan, M. P., Failey, M. P., and Landow, M. P.,

"Examination and Evaluation of the Nine Mile Point-Unit 1 30 Degree Azimuthal Surveillance Capsule, final report from Battelle to Niagara Mohawk Power Corporation, April 23, 1985.

74

V

[MA85b] Letter to Mr. Ray Pasternak from Dr. Michael P.

Manahan, "Radial Flux Profiles for Nine Mile Point Unit 1", June 24, 1985.

[MA85c] Letter to J. A. Zwolinski from Mr. C. V. Mangan, Niagara Mohawk Power Corporation, 300 Erie Boulevard West, Syracuse, NY 13202,

Subject:

Docket No. 50-220, DPR-63, December 10, 1985.

[MA86] Manahan, M.P. et al, United States Patent 4,567,774, "Determining Mechanical Behavior of Solid Materials Using Miniature Specimens",

[MA87 ] Manahan, M. P., "Surveillance Capsules A'nd C'or Nine Mile Point Unit 1", September 30, 1987, Draft Final Report to Niagara Mohawk.

[MA89] Letter to Mr. Yang Soong, Niagara Mohawk Power Corporation, from Dr. Michael P. Manahan,

Subject:

Proposal/Agreement No. 723-R-2326

] Manahan, M. P ~, "Pressure Temperature Operating Curves

'MA91 for Nine Mile Point Unit 1," Final Report dated January, 1991.

[MC89] "SAM McFRAC: Statistical Analysis Methodology for Mechanics of Fracture", PC Version Developed October, 1989.

[NE85] Letter to Dr. Michael P.,Manahan from J. E. Nestell, MPR Associates, Inc., 1050 Connecticut Avenue, Washington, D.C. 20036,

Subject:

Nine Mile Point Unit 1 (NMP-1) Reactor Vessel Surveillance Capsules, dated September 17, 1985. (Revised 11/27/85).

[NMFS] Nine Mile Point Unit 1, Final Safety Analysis Report

[NMTS] Nine Mile Point Unit 1, Technical Specifications, Section 3.2.2, Minimum Reactor Vessel Temperature for Pressurization, updated 1987.

[OD85] Odette, G.R., Lombrozo, P.M., and Wullaert, R.A.,

"Relationship Between Irradiation Hardening and Embrittlement of Pressure Vessel Steels," Effects of Radiation on Materials: Twelfth International Syposi um, ASTM STP 870, F. A. Garner and J. S. Perrin, Eds., American Society for Testing and Materials, Philadelphia, 1985, pp. 840-860.

75

[RO80] Roberts, J. H., Gold, R., and Ruddy, F. H., "Thermal Annealing Studies in Muscovite and in Quartz",

Proceedings of the Tenth International Conference on Soli d State Nuclear Track Detectors, Lyons, France, July 2-7, 1979, Pergamon Press, Oxford, (1980) 177.

[ST64] Stone, W.A., Mechanical Test Report (contract 164 Niagara Mohawk), from W. A. Stone Metallurgical R & D Department, Combustion Engineering, July-October, 1964.

[ST84] Stahl, D., et al, "Examination, Testing, and Evaluation of Irradiated Pressure Vessel Surveillance Specimens from the Nine Mile Point Nuclear Power Station", Final Report from Battelle-Columbus to Niagara Mohawk Power Corporation (July 18, 1984).

[ST84] Stahl, D., Manahan, M. P., Failey, M. P., Landow, M.

P., Jung, R. G., and Lowry, L. M., "300 Degree Capsule Examination, Testing and Evaluation of Irradiated Pressure Vessel Surveillance Specimens From the Nine Mile Point Nuclear Power Station", Niagara Mohawk Power Corporation, July 18, 1984.

76

APPENDICES

APPENDIX A DESCRIPTION OF CAPSULES A'ND C'-2

APPENDIX A. 1 DESCRIPTION OF CAPSULES A'ND C'his appendix of the report describes the contents and layout-of Cap-sules A'nd C'. These capsules are considered to be advanced boiling water reactor (BWR) surveillance capsules since they contain mechanical behavior, temperature monitors, and dosimetry, which are not found in the currently used BWR capsules. The capsule inventory and assembly drawings are provided in Appendix A and the as-built photographs are provided in Appendix B. Details concerning the capsule design and fabrication of the contents are provided below.

Battelle established a subcontract with Metrology Control Corpora-

, .tion (MCC) to provide the radiometric monitors (RMs), helium accumulation fluence monitors (HAFMs), solid state track recorders (SSTRs), melt wires, and solid state track recorder-temperature monitors (SSTR-TMs). MCC entered into a contract with Westinghouse Hanford Company (WHC) for U.S. Department of Energy (DOE) dosimetry services of the National Reactor Dosimetry Center RDC) at the Hanford Engineering Development Laboratory (HEDL) and the helium mass spectrometry laboratory at Rockwell International Rocketdyne Division (RI-RD) (Mc87a). All other work described in this report was performed by Battelle.

Dosimetr In order to determine neutron exposure levels throughout the reactor vessel-geometry with a minimum uncertainty, it is necessary to employ a com-bination of rigorous analytical techniques and spectral dosimetry. In parti-cular, neutron transport calculations coupled with advanced neutron monitor sets located in surveillance capsules provide an improved approach to the determination of both axial and azimuthal exposure gradients within the pres-sure vessel. It is felt that this approach is superior to the current dosi-metry practice in which only wires of Fe, Cu, and Ni are included.

The design philosophy used for the dosimetry was to rely heavily on proven, time-tested technology which is why spectral RMs were used widely.

owever, as a backup, recognizing that the technology will advance over the xt decade, we have included HAFMs and SSTRs. The SSTRs are integrating A-3

I~

dosimeters which accumulate fission tracks throughout the irradiation.

Automated track counting is under development and counting methods will undoubtedly improve over the next few years. The HAFMs are wires composed of material with an appropriate (n, a) cross section. The helium accumulates in the material with increasing fluence and can be measured using any one of a variety of destructive chemical analysis techniques.

This section of the report gives a complete description of the advanced dosimetry fabricated for MCC by the NRDC at HEDL to provide two sets of dosimetry at three axial locations per set for Capsules A'nd C'. One set is to be exposed for -14 efpy years and the second set for -24 efpy years.

This advanced dosimetry has been prepared by the NRDC at HEDL with the assis-tance of RI-RD. The advanced dosimetry services provided are delineated in Table 1 of Appendix C. Assistance was. also provided by the NRDC to MCC in the preparation of purchase order specifications for the acquisition and fabrica-tion of the outer stainless steel (SS) and inner Gadolinium (Gd) shield capsules as well as the subsequent gA of the fabricated parts.

Capsule identification (etched on the outer SS surface) is as specified by Battelle. Anticipated fluence level exposures of -14 efpy and

~ ~

-24 efpy are denoted by A and C, respectively.* Bare and Gadolinium-covered dosimeters are designated by B and G, respectively; and the top, midplane, and bottom axial locations are indicated by 1, 2, and 3, respectively. A fourth identifying letter (A or B) was used when all the dosimetry for a single location could not be loaded into'a single capsule; e.g., capsules AG1A and AG1B. As additional examples of the numbering system:. AG1B indicates an "A" fluence exposure Gadolinium Covered - Top position - and second Capsule B with HAFM dosimetry; and CB3 would denote a "C" fluence exposure - Bare-Bottom axial location. In all instances, the first letter of the identifying number is located at the bottom end of the dosimetry capsule. The materials used and the 18 individual capsules are described in Appendix C. The dosi-metry holders and contents are shown in Appendix D. Each of the 18 capsules contains one or more of the following types of dosimeters: (1) RMs, (2) HAFMs, (3) SSTRs, and (4) SSTR-TMs.

Note: The Battelle designation for A and C are A'nd C', respectively.

A-4

~'

A.2 Tem erature Monitors Two different types of passive temperature monitors (TMs) were supplied to observe irradiation temperature. A set of conventional melt wire (NW) TMs were used to cover the temperature range from 536 up to 598..F; These monitors are proven and time-tested and therefore are the primary source of temperature monitoring. A detailed description of these MW-TNs is provided below.

A second set of TMs were supplied which are based on the annealing properties of SSTRs. SSTR-TNs are a new and novel mean's of passively observ-ing irradiation temperature (Go79). A detailed description of these SSTR-TMs is provided in Section A.2.2 below. The advantage of these monitors is that

'they"are"'potentially"capable of providing an estimate of the average tempera-ture in the capsule. They have not yet been proven for reactor irradiations and the same fission track counting problems are experienced as with SSTRs.

However, these monitors were included as a backup to the melt wires since we ticipate that this technology will continue to develop over the next decade.

~ ~

A.2.1 Melt Wire Tem erature Monitors Eutectic materials with unique melting temperatures were included in Capsules A'nd C'. Composition and impurities greatly influence the melting

'temperature;'therefore, the materials used consist of purities of 99.9 percent or greater, so that the measured melting temperature is within at least +6 F of the recognized melting temperatures. The MWs were encapsulated in quartz tubing, evacuated and backfilled to approximately one atmosphere of helium pressure for optimum transfer of heat between the encapsulated NW and the outside environment. The MW-TMs included in Capsules A'nd C're listed in Table A-1 and photographs are provided in Appendix E.

The NW capsules were formed f'rom 0.24 inch outer diameter quartz tubing with 0.062 inch thick walls. To facilitate identification of these different NW temperature monitors, each MW was encapsulated in quartz tubing of a specific length as shown in Table A-l. After fabrication, the MW cap-sules were leak checked, pressure-tested and certified by con'ducting melt eriments. The certification test results indicated that all the MW ompositions melted within +2 F of the nominal value.

A-5

'U P I

TABLE A-1. MW MATERIALS AND MELTING TEMPERATURES Composition, WtX Melting Temperature, F quartz Capsule Length, in.

80 Au, 20 Sn 536 1 90 Pb, 5 Ag, 5 Sn 558 1-1/4 97.5 Pb, 2.5 Ag 580 1-1/2 97.5 Pb, 1.5 Ag, 1.0 Sn 588 1-3/4 98.8 Cd, 1.2 CU 598 2 A.2.2 Solid State Track Recorder-

.Tem erature Monitors A number of limitations of MW-TMs can be overcome by using a new passive TM. This new passive TM provides for continuous monitoring of tem-rature with assigned uncertainties. It represents a novel, patented appli-ion of SSTR techniques [Go79]. Since tracks are annealed in SSTRs at elevated temperatures, the degree of annealing can be used to measure the in-situ temperature. Of equal importance is the nature of this SSTR-TM response, since it is based on track annealing and therefore, responds to the average time-temperature history that is experienced by the SSTR. Conse-quently, normal,.temperature excursions due to reactor operations are expected to make only small contributions to track annealing and the SSTR-TM will therefore furnish an accurate measurement of the in-situ temperature.

Fission fragments produce a narrow path of radiation damage in quartz glass, mica, and other SSTR materials. When chemically etched, a hole is produced along the damage path which can be seen under a microscope; this hole is called the fission track. All or part of the damage can be annealed out by heating the SSTR. This annealing, when carried out before the etching process, drastically modifies the length and/or diameter of the track when etching occurs. The amount of annealing is a function of time and temperature

[Go79, Ro80].

'n the case of quartz glass SSTRs, the annealing effect will reduce diameter of tracks produced by normally incident fragments, and will.

A-6

eliminate some of the tracks produced by isotropic incidence (exposures made with the source in contact with the SSTR material). In the case of mica SSTRs, the track length is reduced if damage has been partially removed by annealing.

The outstanding problem in the application of SSTR-TMs in power reactors is the lack of calibration data. MCC will be initiating an extensive calibration program within the next few months. Calibration curves for SSTR-TMs are obtained by observing the annealing characteristics of SSTRs.

Sets of quartz glass and mica SSTRs are prepared in the same way the SSTR-TMs were prepared for inclusion in the NMP-1 replacement surveillance capsule.

Both isotopic and normally incident fission fragment irradiations are used.

Subsets of these SSTRs are then heated in an oven at constant temperature for a given period of time. The oven must be well regulated so that stable temperatures are attained with an accuracy of at least +4 F. Upon withdrawal from the oven, the SSTRs are etched and subsequently scanned.

Calibration data are generated by repeating this process at different oven mperatures for different time durations. Preliminary calibration data for uartz glass and for mica have already been reported [Go79, RoBOJ.

Both quartz glass and mica are used for SSTR-TMs in Capsules A'nd C'. The pre-irradiation preparation of quartz glass and mica SSTR-TMs is discussed separately in Sections A.2.2.1 and A.2.2.2.

A.2.2.1 Pre aration of uartz Glass SSTR-TMs. The type of quartz glass chosen for SSTR-TM applications in reactor surveillance should possess a very low concentration of fissionable impurities. Consequently, a high purity quartz glass, namely Supersil, has been used. Supersil quartz glass was supplied by Amersil Company, Inc., Hillside, New Jersey.

The Supersil quartz glass was cut into a disk shape, 0.178 inches in diameter by 0.06 inches thick. Both surfaces of the disk were then mechani-cally polished to remove scratches and imperfections. A final surface polish was obtained by etching in 49 percent HF at 70 F for 120 minutes.

The polished Supersil quartz glass was then cleaned by rinsing in the following solutions in the order given: (1) Palm Olive Soap Solution,

2) distilled water, (3) double distilled water, (4) Ultrex BN nitric acid,

, finally, (5) double distilled water. This cleaning process was used both A-7

,'ljl 4

Qd+(

~n

prior to irradiation with fission fragments

~ ~ ~

and just prior to encapsulation

~ ~

for reactor surveillance.

A.2.2.2 Pre aration of Mica SSTR-TMs. The type of mica chosen for SSTR-TM applications in reactor surveillance should also possess a ve~ low concentration of fissionable impurities. To this end, mica from India, sup-plied by The Perfection Mica Company, has been used. The mica was first pre-annealed in an oven for about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 1076 F. It was then pre-etched in 49 percent HF 70 F for about 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />. When placed under a microscope, resi-dual fission fragment tracks are seen. The mica with a low density of tracks was selected. It was then cleaved to thicknesses of about 0.003 inches.

Disks, 0. 178 inches in diameter, were punched from the freshly cleaved mica.

Mica'STR-TMs were cleaned by the same procedure used for the quartz glass. This cleaning process is important just prior to encapsulation, in order to rid the mica of transuranium or other actinides that may produce undesirable fission fragment tracks while being exposed to neutrons in the ctor.

A.2.2.3 SSTR-TM Fission Fra ment Irradiation. Prior to use, the SSTR-TMs must be exposed to fission fragments. A uniform source of 252Cf was

.used for the irradiation. Two types of, fission fragment exposure, isotropic and normally incident, were used. The isotropic exposure is performed by placing the 'SSTR-'TM in firm contact with the spontaneous fission fragment source. Normally incident exposures were obtained in a vacuum chamber. There was approximately 2.2 inches between the SSTR-TM and the spontaneous fission fragment source. For the isotropic exposures, a 252Cf spontaneous fission fragment source having a diameter of 0.25 inches', on a stainless steel disk 0.75 inches in diameter, was used. The source had a very thin gold coating.

On 1-29-86, this source had an activity that produced 9230 tracks per minute per square centimeter in mica placed in direct contact with it. A 252Cf source about 1/8 inch in diameter was used for exposures in the vacuum chamber. On 1-31-86, an exposure gave a track density for normal incidence of 60.1 tracks per minute per square centimeter. The statistical accuracy of oth measurements is 2.7 percent.

A-8

+1IF A.2.2.4 Fission Fra ment Irradiation of uartz Glass SSTR-TMs. One -

side of the quartz glass SSTR-TM was irradiated isotropically with fission fragments and the other side was irradiated with normally incident fission fragments. Estimated track densities for the isotropic and normal irradia-tions, respectively, are given in Table A-2. Two such quartz glass $ STR-TMs were prepared for each capsule. Isotropic exposure data from the mica SSTR-TM are given in Table A-3.

A third quartz glass SSTR-TM was prepared for each capsule as described earlier, but was not irradi'ated with fission fragments. One surface of this third quartz glass SSTR-TM is used to assess fission fragment track background induced during the reactor surveillance irradiation. The other surface was placed in firm contact with a 23BU deposit to form an SSTR neutron dosimeter for the'reactor surveillance irradiation. This SSTR neutron dosi-meter will provide fission fragment tracks that are formed throughout the entire course of the surveillance irradiation. Consequently, this type of SSTR-TM provides information on the time-temperature history of the surveil-

~ ~

nce irradiation.

~ ~

After being cleaned, the

~

SSTR-TMs were loaded into capsules AB2 and CB2 as shown in Figures A-1 and A-2.

A.3 Char V-Notch S ecimens The types-and number of Charpy specimens in the reinserted capsules were summarized in Table A-2. Niagara Mohawk decided to weld reconstitute the Charpy specimens prior to reencapsulation. The specimens can be tested in this configuration or miniature Charpy specimens could be machined in the future when the capsules are withdrawn. Battelle has developed a miniature Charpy test in which 16 miniature specimens can be machined from one conven-tional specimen.

A total of 24 Charpy V-notch (CVN) specimens (12 base, 12 weld) were reconstituted for inclusion in Capsule C'. A total of six CVNs were recon-stituted for inclusion in Capsule A'. The specimens were stamped on the ends with the appropriate F.A.B. code designation. The original F.A.B. code

'dentification was used with an additional letter "A" or "B" to designate a onstituted specimen. The specimen identifications and locations within the packets are given in the Appendix A specimen inventory drawings.

A-9

TABLE A-2. EXPOSURE DATA FOR SUPERSIL II QUARTZ GLASS+

Isotro ic Normal Incidenae SSTR Label Time (min) Tracks/cm2 Time (min) Tracks/cm2 C* 20.6 1. 2E5 1014 4.45E4 11 30 2.4E5 680 4. 1E4 12 30 2.4E5 680 4. 1E4 13 0 0 0 0 3C* 21 1.2E5 1014 4.45E4 21 30 2.4E5 680 4.1E4 22 30 2.4E5 680 4.1E4 23 0 0 0 0

+ 'll exposures made on 1-27-86, except 4-1-87.

for C and 3C, which were exposed on

• Control samples, not put into the reactor. These will be stored at or below room temperature, and etched along with the SSTR-TM when they are removed from the reactor.

TABLE A-3; ISOTROPIC EXPOSURE DATA FROM MICA SSTR-TM+

SSTR Label Exposure Time in Min. Predicted Tracks per cm2 Cl* 13.75 9.2E4 11 11 1.0E5 12 10 9.2E4 13 0 0 C2* 13.75 9.2E4

'2 21 10 10 9.2E4 9.2E4 23 0 0

+ All exposed on 1-27-86 except for C1 and C2, which were exposed on 4-1-87.

• Control samples, not placed in the reactor. These will be stored, at or below room temperature, and etched along with the SSTR-TM when they are removed from the reactor.

,II ~ 4 %18 ~

-~

f ISSION FRAGHCt4T DISTR E D XRRRD I R T'I OH

.t!OtiE VICE $ 50TROPXC NICR 12 b I SOTROP ZC t40NE U-2 39 'lt23 3 NONE MICR 13 b HONK:

NORMRL QLlRRTZ I BOTROP 3C I SOTROP I C QURRTZ I2 b OORHFiI U-2 38 423 8 QURRTZ 13 t MICR SPRCER FIGURE A-1. ARRANGEMENT OF SOLID STATE TRACK RECORDER TEMPERATURE MONITORS IN CAPSULE AB2 (A t denotes that the ID is on the top surface of the mica or quartz glass, whereas a b denotes that the ID i s on the bottom surf ace. )

FISSION F RRGHEHT SSTR TD IRRRDIRT I Ot4 NOAE Nl CR 21 t IBOTROP1C I SOTROP IC NONE U-238 +226 NONE NI CR 29 b HOME NORgQ QURRTZ 2I t I SOTROP IC ISOTROPIC QURRTZ 22 b HO~PL Ll-238 422$

NONE QVRRTZ 23 t NlCR SPRCER FIGURE A-2. ARRANGEMEHT OF SOLID STATE TRACK RECORDER TEMPERATURE MOHITORS IN CAPSULE CB2 (A t denotes that the ID is on the top surface of the mica or quartz glass, whereas a b denotes that the ID is on the bottom surface.)

A.4 Miniature and Conventional Tensile S ecimens The types and number of tensile specimens in Capsules A'nd-C'ere summarized in Table A-2. In cases where conventional tensile specimens were already tested, miniature specimens were machined from the broken halves. The Battelle proprietory specimen design is shown in Appendix A [Ma86]. A total of two miniature tensile specimens (base) were included in Capsule A'nd eight miniature tensile specimens (four base, four weld) were included in Capsule C . The miniature tensile specimens were numbered consecutively from one to 10. The conventional tensile specimens retained their F.A.B. code designations. The specimen identifications and locations are shown in the Appendix A inventory drawings.

A.5 Ca sule Desi n and La out Surveillance capsule packet inventory drawings were prepared for the o reinsertion Capsules A'nd C'nd are given in Appendix A. The inven-ories include all specimens shown, in Table A-2 and the recommended dosimeters and temperature monitors. Packet fabrication drawings were also prepared that show specimen, dosimeter, temperature monitor, and spacer locations and dimen-sions in each capsule. GE fabricated the spacers, packets, baskets, and lead tubes. The drawings indicate the specimen F.A.B. code designations and the as-built photographs show specimen by specimen agreement with the fabrication drawings.

The dosimetry was carefully oriented and spaced to avoid neutron field perturbations which would affect the data. The dosimetry was also located so that the axial flux profile can be measured. Good heat transfer is insured by tight specimen packing and a helium backfill in each packet.

A.6 Archive Materials The length of these planned irradiations, namely Capsule "A" for 14 efpy and Capsule "C" for 24 efpy, make it mandatory that archive dosimetry nd TM materials be retained by Niagara Mohawk. Experimental results from wer reactor irradiations can sometimes by inconsistent or hard to resolve.

Cf It is

~

essential, therefore, that these archive materials be available as a

~

contingency should problems arise with the experimental dosimetry results.

With these materials, calibration experiments can be repeated and, if n'eces-sary, neutron benchmark irradiations can be. conducted. Should definitive follow-on experiments of this type be needed, the availability of these archive materials can be a crucial factor in the success of the dosimetry analysis. To this end, archive materials, identical to those used in Cap-sules A'nd C'ave been provided to Niagara Mohawk. Tables A-4 through A-7 summarize the supplied archive materials for RM neutron dosimeters, SSTR neutron dosimeters, MW-TMs and SSTR-TMs, respectively.

TABLE A-4. ARCHIVE RM AND HAFM NEUTRON DOSIMETRY MATERIALS Dosimeter aterial Type PO No. Batch Vendor Form guantity Fe RM 07448 26/17944 MRC 0.020D Wire 2 inches Ni RM "1" SE Roll 2 Semi Element 0.020D Wire 2 inches Cu RM, HAFM 19047 CPI 3054 Cominco Am. 0.020D Wire 2 inches Co/Al RM 44451 SC Bar 26 Sigmund Cohn 0.020D Wire 2 inches Ti RM 19046 Cat 614 Reactor Exp. 0.020D Wire 2 inches Al HAFm 19045 SE Roll 1 Semi Element 0.020D Wire 2 inches NOTES: (1) Archive samples of the vanadium encapsulated fissionable RM dosimeters are not provided; those included in Capsule Sets A and C are to be used for this purpose.

(2) Samples of the Be, Fe, and Ni HAFM dosimeters are not provided; but are available for purchase from RI-RD.

A-14

V PI TABLE A-5. ARCHIVE SSTR NEUTRON DOSIMETRY MATERIALS SSTR De osit(2)

SSTR(1) Mass Type ID Diameter Isotope ID Total (pg) Density -(pg/cm2)

Mica 219 0.168 235-U 219 0.826 8.81 Mica 231 0.168 238-U 231 7.65 81.6 Mica 199 0.168 237-Hp 199 10.39 110.8 (1) Three additional mica SSTRs of 0.168 inch diameter have been supplied as control samples to be stored along with the archive'SSTR dosimeters 219, 231, and 199. These control SSTRs are unnumbered and are not in contact

,with any deposit.

(2) The outer aluminum foil of the archive SSTR packages are marked as follows: U-235 is labeled 5; NP-237 is labeled 7; U-238 is labeled 8.

TABLE A-6. ARCHIVE MW-TM MATERIALS Melting Temperature Length Diameter Composition, WtX (F) (in.) (in.) Source 80 Au, 20 Sn 536 0.25 0.030 Indium Corp. of America 90 Pb, 5 Ag, 5 Sn 558 0.25 0.090 Babcock & Wilcox 97.5 Pb, 2.5 Ag 580 0.25 0.084 Babcock & Wilcox 97.5 Pb, 1.5 Ag, 1.0 Sn 588 0.25 0.084 Babcock & Wilcox 98.8 Cd, 1.2 CU 498 0.25 0.083 Babcock & Wilcos A-15

TABLE A-7. ARCHIVE SSTR-TM MATERIALS Material guantity Description and Purposes India Ruby Muscovite Mica Pre-annealed and pre-etched disks 15/16 in. in diameter and about 0.004 in. thick. Representative material for SSTR-TM and SSTR dosimetry.

India Ruby Muscovite Mica Cl and C2 as described in Table 4-2.

Needed as standards when mica SSTR-TMs are removed from the reactor.

Supra II quartz Glass 3/8 in. thick squares about 40 mi ls thick. Representative materials used in SSTR-TMs.

Supra II quartz Glass C and 3C are 3/8 in. squares about 40 mils thick. They are described in Table 4-2. Needed as standards when the Supra II quartz glass is removed from the reactor.

A-16

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2. Charpy V notch shoItd flee the pressure SIGNATURE DIV DATE 505 King Avenue vessel wall. oRAwccpv/ .-"".".":Battelle Columbus. Ohio 43201.2693

3. Charpy packets. matt wire packets and /35 r//c SC. Cotombuc tabor 2 cocicc Telephone f614) 424 6424 lcnsilc 'lccbcs arc sc)kd cs follows: RAFTIRO Arvo, rc//Q c.

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~2/ i~err 3-\. Exhaust air to (educe intecnal pressure to I lorr. Back lill with welding grade Ileticm al atmosphenc CAPSULE A'SSEMBLY DRAWING FOR Attn, TASK COORO pressure and scat.

3-2. Verily leak tigttnesc using mass

~ 4 NINE MILEI POINT UNIT 1 spectrometer. fhe cllowable leakage f3A MANAGER is to s atm cc.min. (RE-ENCAPSULATION)

~ 04 4. The original dosime!ry consisted of Fe, Cu, Ni Stgg COOS tOSNT NCX OCV NO. DWCA NO. RKV.

wires. These wires ~re rcinstalled at each I location In the packets al shown lone Fe. one FROJCCT r'/.

Ati'O.

79986 442 BCD - NMP -OOI 1 O Cu, and one Ni al e ch lccation).

Z NOTFi Fnn 33rD fASR Bv oa cvcc v Attn Z

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Irradiated HAZ Reconstituted Base Irradiated Weld Tensile Tubes Melt Wires 0 0 0 SPACER v~@>~e0 g@ggg,fc Cggb E71A EO1 JD1 JLK JTA JUL (Beee) (Vreld) (KAZ) (KAZ)

J12 E12 ED2 t

598 F J13 E31A EO3 l J14 E2E ED4 1l Advanced S Dosimetry IB)f P 588 F A

J15 E2T EOS C Original E Dosimetry R J16 E2UA E06

{Fe, Cu, Ni Wires) il 580 F JD2 1 JL2 JUJ J17 E17 ED7 I (Bere) (Bete) plveld) (KAZ)

Advanced Dosimetry )IID)y BIDtr trio)r BZO)r tyco)r B cQ)r J1A E1AA EOAi 2 558 F (Bere)

J1B E2Y EDB See Note1 J1C E1CA EOC 536 F J1D E1D EOO' JTE EBKA EOE NOT A SCALE DRAWING 0

Z o

O Ri Oa -0 SIGNATURE DIV DATE 505 Kmt Avenue

1. REVIEW BOARD DATE

., gC NOTE: t. The original dosimetry consisted ol Ff, Cu, Ni wires. These wires'are reinslailed at each location DRAWN SV~ /3d / Z'6

"..,.:Battelle-Columbul teboreloINI Columbus Otvo 4320) 2693 Telephone (6 I 4) 424.6424 In the packets ss shown lone Fe, one, Cu, and onc DRATTINDAreo Ni at each location).

~/S I

2. The advanced dosimetry tubes musl lie Installed CAPSULE A MECHANICALBEHAVIOR

2. TASK COORO JfASK LEADER so that the liest character {A) In the tube Aeep.

Identification Is st the lowest elevalion as shown in the drawing.

N

~li3 SPECIMEN INVENTORY DRAWING FOR NINE QA MANA(IFR MILE POINT UNIT 1 (RE-ENCAPSULATION)

3. The capsule must be assembled exactly ss IEg~l 4 AZ Qx specliied due to neutronic snd other design considerations Zjztr!JC SIZK COOK lOKNT. NCk OIV. NO. OWO. NO.

- - 002 RKV.

4a/f 4 C 79986 I'Roeccv ATTD. 442 BCD NMP 1 NOTE: sn "r ss BvoACIAT A(ALL se CIAL Tep.

SCALE Aocr. N093B- 2900 sKKKT I of I CAT II~ I ~ I IOIIN D ~ Ie Ie

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ALL DIMENSIONS ARE IN INCHES 0

Z

+0 cs 20 -0.0)s ~ O SCALE 5: I ALL DIMENSIONS ARE IN INCHES B

~

SIGNATURE DIV DATE 505 King Avenue

1. D Stok BOARD DATE oRAwsrov~ "'.".".Battelle Columbus. Oho 43201 2693

/35 Columbus tsborr tories Telephone (614) 424.6424 P

~g~ac. MINIATURE TENSILE SPECIMEN FOR

2. TASK COORD ASK LEADER adit kOTESt 1. 0~0.150 g0.001 Diameter at cente of reduced section.

2. 0'~Actua'I -D" Diameter+ I0.002 -0 003 ) al ends of reduced section, tapering to "D" at center.

AFJ'o NINE MILE POINT UNIT i (CAPSULE A S. OA f4AMABER 3. 100% dimensional inspection require

4. Polish reduced section and radius to 2 rms, //-g /.A

"" AND C RE-ENCAPSULATION)

COtrg IDENT. ko. tSIV. NO. tSWG. NO. REV remainder as turned.

79986'42 I

4'OTE:

5. All olher dimensions are to be within s0.001 . enopc'T Aeao C BCD - NMP -003 1 sort ngtgAs'c tsv 0A orrcv clAL rro N0958-2900 sttggT I of I CAT. Is ~ sot roau o ~ Ir rs 4

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I/32 X 450 CHAM TYP A~<<~ ~

I/16 CUT 0.005 to O.OIO g+ I/32 Clearance with actual IE 10 of tensile lube I 0.255 I.D. I0.444 10. Ref.)

ALL DIMENSIONS ARE IN INCHES Source. General Electric 0rowing No. I07C3?97 DRAY/INGAPPROVED BY:

,R<b I 0 KO SIGNATURE DIV DATE I I

1. D SIGN REVIEW BOARD DATE

/35 4/iA/sb Tetephene (6t4) 424 6424 PRAFTIRC AA00

<tu!X

~u P/X/TC.

CONVENTIONAL TENSILE SPECIMEN

2. TAsK CDDRD ABK LEADER Aeep, Ml:~4 FILLER SLEEVE

3. OA MANAGER

>@8 As/& ~ 'e/2 SliII/W SIZC CDDC IDENT. ND. DIV. ND. DWO. ND.

- NMP - 004 Rcv.

NOTE: ron RracAse ev oA c>>4.v eeolecv Aeep.

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0 5000 Note

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NOTES: l. The outside dimensions oi he basket contents have a cninimum .015 clearance c((0 for disassembly. SIGNATURE D(V DATE 505 Kurt( Avenue

1. OE6lGM REVtEW BOARD DATE 2.

~

Charpy V.notch should fac the pressure vessel wall.

onAWNav ",,-Battelte Columbus. One 43201 2693 Tetentene (614) 424 6424 Columbus tcbornor les KXTE~ 3. Charpy packets, melt wire ackets and tensile tubes are sealed a s follows:

, M~'YCK 3-1. Exhaust air to reduc e internal pressure to 1 tort. 8 I k till with CAPSULE C-ASSEMBLY DRAWING FOR

2. TASK COORD. ASK LEADER welding grade Neliu at atmosphecic ACCO pressure and seal.

NINE MILE POINT UNIT 1

3. QA MANAGER

~J~

3-2. Verify leak tightness sing mass spectrometer. The ap wable leakage is lp-s atm cc/min. I'gn aa 'z (RE-ENCAPSULATION)

SCZE CODE CDENT. NO. Dtv. NO. DWP. NO. REV.

4. Four plugs foc tensile lube are 1.0 'by NOTE: FOR RctpasE av PA Pcrcv 0.434 S 0.002 diameter. >nocccz Acco.

i<("c C 79986- <<Z BCD -NMP -OO5 s

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D D Reconstituted Weld Unirradiated Base Reconstituted 8 ase Tensile Tubes Melt Wires SPACER T01 P P (ease)

L L EDKA NCOI E1JA U P U G L G EDLA NC21 E1JB U 3 7 598 F EDMA NC02 E1KA I) (esse) (Weld)

EJTA NC22 E1KB Advanced S Dosimetry P IBO ZBO CBO 588 F A 4 8 C (ease) (Weld)

JAEA NC03 EASA E C R

JAMA NC23 EASB 580 F See Note 1 J2CA NC04 E42A T02 T21 5 9 Advanced (esse) (8222) (8222) (Weld)

Dosimetry Tf 1SO 8(SO )fZOO BCOO T(COO BCSO,:

J1LA NC24 E1MA 'f 6 10 558 F (ease) (Weld)

J1MA NCOS E1UA I J1PA NC25 E3TA P P 536 F J1TA NC06 E7EA L L U U J1JA NC26 See Note 1 J2C G G NOT A SCALE DRAWING 0

z ts 0

BY:

SIGNATURE DIV DATE 6(5 King Avenue

1. Qg@gf AEVIEW BOARD DATE DRAWN av ~

>/I>//'b

"".".'"-Battelle CorumbuL Ohio 4320) 2693 NOTEt l. All reconstituted specimen were prepared from //Z5 Columbus lsborrroIN2 Telephone (6)4) 424 6424 broken weld or base metal ppecimens excepl tor 2L(~gc J2CA and J2CB..These specimens were prepared </cd from the weld and base metal portion of a broken HA2 specimen, J2C. M/gry CAPSULE C'-'MECHANICAL BEHAVIOR G

2, ~COOAO ASK LEADEA 2/ )Attn

~I

• SPECIMEN INVENTORY DRAWING FOR

~i~c y

2. The advanced dosimetry tebes must be installed ~

I/

so that the first character (C) in the tube

s. ~ANAG('n Identification is at the (owe'st elevation as shown NINE MILE POINT UNIT1(RE-ENCAPSULATION)

~SPAZ/xseb.. ~y4 in the drawing. ft-Q Cu4 IA/g sir f(tC srzc coos togrrr. No. Olv. No. owo. rKx REV.

3. The capsule must be assersbled exactly as NOTE: sent ~

..sc ov OA oN(v specitied due to neutronic nd other design considerations.

trro/ccv Attp, 7 9 9862BCDNMP0061 CCIAL Attn, 2/2 /~ N0938-2900 srtggT I of I cA'I, I ~ I I 22 rouu 0' ~ 2'22 3 I I 1

P 0

Aa~

V 4

4. 'Rg I,>>

-l1

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v I

APPENDIX B AS-BUILT PHOTOGRAPHS FOR CAPSULES A'ND C'

e h, ~

1 11 g jf I iy I

~pa"

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~-,.t P~, pt

+:aat~ '$

ca ft t'ai wm A a~ I"~ggM Ia

~Per~i It' aa tyI ta j>p li f ..A Y

CAPSULE ~.

'I I

tc.)

..CI'PSV!L st'IGURE

1. CAPSULE A'HARPY SPECIMENS

E l)

FIGURE 2. CAPSULE A'HARPY SPECIMENS B-3

1I

,I

~ g

'k

FIGURE 3. CAPSULE A'ENSILE SPECIMENS B-4

0 r

FIGURE 4. CAPSULE A'EMPERATURE MONITORS B-5

I!

FIGURE 5. CAPSULE C'HARPY SPECIMENS

%~iN 1 ~ p>

FIGURE 6. CAPSULE C'HARPY SPECIMENS B-7

Wag~ 11 FIGURE 7. CAPSULE C'ENSILE SPECIMENS B-8

, p r< bA FIGURE 8. CAPSULE C'EMPERATURE MONITORS B-9

APPENDIX C DOSIMETRY DESCRIPTION FOR CAPSULES A'ND C'

TABLE 1 ADVANCED DOSIMETRY PROVIDED BY METROLOGY CONTROL CORPORATION (MC~)

Radiometric (RM) Flux Monitors

a. 6 sets of non-fissionable Gd covered RMs; each set consists of one each of: Fe, Ni, Cu, Ti and Co/Al metal wires.

b. 6 sets of non-fissionable "bare" RMs; each set consists of one each of: Fe and Co/Al metal wires.

c. 2 sets of fissionable, vanadium encapsulated, Gd covered RMs; each set consists of one each of: U-235, U-238 and Np-237 oxide wires.

d. 2 sets, of fissionable, vanadium encapsulated, "bare" RMs; each set consists of one each of: U-235 oxide wire.

Solid State Track Recorder (SSTR) Fluence Monitors

e. 2 sets of Gd covered SSTRs; each set consists of one each of: U-235, U-238 and Np-237, (fission deposits on solid state track recorder backing material with mica SSTRs).

2 sets of "bare" SSTRs; each set consists of one each of: U-235, (fission deposit with mica SSTR).

Helium Accumulation Fluence Monitors (HAFM

/

g. 6 sets of non-fissionable, Gd covered HAFMs; each set consists of one each of Be (pieces wr apped in Al)*, Cu, Co/Al (or Al), Ni and Fe metal"wires.

h. 6 sets of non-fissionable, "bare" HAFMs; each set consists of one each of Be (pieces wrapped in Al), Cu, Co/Al (or Al), Ni and Fe metal wires.

Thermal Monitors TMs)

i. 2 sets of melt wire (MW) TMs; each MW-TM set consists of five melt wires, each of which is contained in a quartz capsule.

• Material will be in the form of one or more small pieces of Be metal, wrapped in aluminum with a nomial effective package diameter of -20 to 40 mils and length of less than 0.25".

C-2

kt 4 TABLE 1 (Cont'd)

Solid State Track Recorder (SSTR)-THs

j. 2 sets of TMs; each set consists of 3 quartz and 2 mica pre-irradiated SSTRs wrapped in Al. Each set placed in a Gd cover (with other dosimeters) in a SS dosimetry capsule.

Oosimetr Ca sule Fabrication

k. 6 bare and 12 Gadolimium lined stainless steel holders were loaded and assembled.

1. The SS capsules were back-filled with high purity (dry) Argon** prior to welding and were subsequently leak checked by immersion in near boiling water (that has been boiled to remove air bubbles). X-Rays of the loaded holders were also taken.

QA and Oocumentation

m. A QA data package including the "as-built" description and documentation of all dosimeters and encapsulation containers is provided in the Appendices.

• • Helium was not used because it could adversely (by absorption or diffusion) bias the interpretation of the HAFH monitor results.

C-3

TABLE 2 NINE MILE POINT UNIT-1 OOSIMETRY MATERIALS (Total Weight Inc uding Dosimetry, Backing, and Encapsu ation Materials)

"A" Capsule Series "C" Capsule Series Material Wt (Grams) Wt (Grams)

'238-U 0.007 0.007 237-Np 0.006 0.006 235-U 0.002 0.003 Al &.61 M.59 Be 0.020 0.013 Co 0.0002 0.0002 0.38 0.37 Fe 1.03 1.00 Gd <<7 7 <<7.7 Ni 1.59 1.56 0.16 0.15 0.06 0.06 Stainless Steel -8.7 -8.7 Quartz <<4.7 <<4. 7 C-4

TABL CAPSULE LOADING SEQUENCE "A" SERIES (-14 EFPY EXPOSURE)

(FROM BOTTOM OF SS OR Gd CAPSULE)

Loaded SS Ca sule Al Wr a ed Dosimeters Location ia. In. en t n. Tyy>e 0> a. emnth t. m Top AG1A 0.253 0.829 2.9025 RI RM 0.165 0.368 251.82 AG1B 0.254 0.834 2.9723 H1 HAFM 0.180 0.380 231.65 Fe Grad. 0.

AB1 0.256 0.830 '.7789 H2 HAFM 0.175 0.394 256.42 (R2) RM Mid AG2A 0.257 0.832 2.9552 R3 RM 0.190 0.375 287.55 AG2B 0.253 0.830 3.0509 H3 HAFM 0.175 0.358 255.85 7

8 SSTR 5

Fe Grad.

AB2 0.253 0.829 1. 9125 A ST -T H4 HAFM 0.180 0.376 250.83 5 SSTR 0.195 0.012 R4 RM Bottom AG3A 0.254 0.834 2.8947 R5 RM 0.180 0.372 225.8 AG3B 0.253 0.834 3.0017 H5 HAFM 0.175 0.355 248.5 Fe Grad.

AB3 0.256 0.834 1.7750 H6 HAFM 0.177 0.368 258.25 1 No dosimeter materials are closer than 0.10" from outer SS capsule top. Therefore the top can be parted off up to 0.10 from top. Care must be 'taken that no burrs exist nor the open end swaged as the inner capsules may not shake out.

2) It is recommended the Gd capsules actually be cracked open (i.e., use "nut cracker" on lower end of capsule) the reason for this is that several of the dosimeter packets are quite a tight fit and possibly would be damaged in the attempt to pull them out of the capsule or to cut open the bottom and push them out.

3) No Al material should be discarded without close observation as to whether it is simply spacer material of wrapped dosimeter material. Fe gradients are wrapped individually and though marked Fe could be mistaken for spacer material. Both bare spectral Co/Al and Fe/235-U are wrapped in Al and are unmarked and could therefore be mistaken for spacer material.

TABLE 4 TASK/ID: 85-WHSC-9010 Task 1/Nine Nile Point Unit 1 SS Capsule Number: AG1A RH Set ID: R1 HAFH Set ID: SSTR ID: SSTR-TH ID: Cover: Gd 81 Location:

Purchase Order Batch Haterial Sample Elemental Set ID Haterial Number Number Desert tIon ~Wt. m s ID ~tom . Comments Stain ess tee Outer AG1A Ca sule Bottom Gd II1 Gd Inner Ca sule Bottom R1 Ti S acer e

~ u Fe 26/17944 folded wire 53.436 99.999X SE Ro 2 ~ U Ni 07448 2 folded wire 47.967 99.999K

~ U Cu 19047. CPI 3054 folded wire 61.994 99.999K Al Wra ed

~ U tl- Spectral Set 19046 Cat 614 folded wire 38.266 99.917K Sc ar ut-Co/Al 44451 26 folded wire 9.259 0.506K Al Wra ed Ti S acer 251.82 0.165" OD X 0 '68" OL Al Shim S acer Gd Inner Ca sule To Al Shim S acer arge e eat Sink S acer SS Outer Ca sule 2902.48 0.253" OD X 0.830" OL To

TABLE 5 TASK/ID: 85-WHSC-9010 Task 1/Nine Mile Point Unit 1 SS Capsule Number: AG18 RH Set IO: HRFH Set IO: HI SSTR IO: SSTR-TH ID: Cover: Gd FIO LocetIon: ~To Purchase Order Batch Material Sample Elemental Set ID Material Number Goober Descr1 tIon ~Ht., m s ID ~Com . Comnents Stain ess teel Outer AG1B Ca sule Bottom Gd 810 Gd Inner Ca sule Bottom H1 Ti S acer Be RI RI-7* 3 Metal Pcs Al Wra ed 4.25 NM-BE-1

.8 ' ngle Fe Rl -RI-4* Fold Wire 62.695

~ ~sm ~ ng e rappe RI RI-11A* Fold Wire 125.384 HAFM Set EDL- ng e Cu 19047 CP I 3054 Fold Wire 33.187 99.999K 0- ng e Al 19045 SE Roll 2 Fold Wire 9.301 99.999K Fe Rx Lot 11 o

Ti S

.03 Disk

'ta acer 231.25 82.26 0.180" OD X 0.380" 92.772K OL Al Wrapped Gradient Al Shim S acer Gd Inner Ca sule To Al Shim S acer Large T el eat Sink S acer SS Outer Ca sule 2972.30 0.254" 00 X 0.834" OL To

• Batch No. given by RI for Boron analysis.

et TABLE 6 TASK/ID: 85-WHSC-9010 Task 1/Nine Mile Point Unit 1 SS Capsule Number: AB1 RH Set 10: R2 HAFH Set ID: H2 SSTR 10: SSTR-TH ID: Cover: Bare Locatton: ~To Purchase Order Batch Material Sample Elemental Set ID Material Number Number Desert t ton ~Nt. m a ID ~Com . Comments Stain ess tee Outer AB1 Ca sule Bottom H-2 Ti S acer Be RI RI-7* 3 Metal Pcs Al Wra ed 2.75 NM-BE-3

.8 ingle Fe RI RI-4* Fold Wire 64.514 Snge Rl Rl-11A* Fold Wire 124.704 E L- "Snge Al rapped Cu 19047 CP I-3054 Fold Wire 30.008 99.999K HAFM Set E - 90 . X . " S ng e Al 19045 Se Roll 2 Fold Wire 9.068 99.999K Ti S acer 256.42 0.178"OD X 0.394"OL eta R2 Fe Rx Lot 11 Disk 82.66 99.772K Al Wra ed o d Co/Al 69-89-0885 SRM 953 9.095 0.116K Al Wra ed Al S acer Shim arge T e eat Sink S acer SS Outer Ca sule 1778.90 0.256 OD X 0.830 OL To

• Batch No. given by RI for Boron analysis.

TABLE 7 TASK/ID: 85-WHSC-9010 Task 1/Nine Mile Point SS Capsule Number: AG2A RM Set ID: R3 HAFM Set ID: SSTR ID: SSTR-TM IO: Cover: Gd I2 Location: Hid Purchase Order Batch Material Sample Elemental Set in Material Number Number Oescri tion ~Mt. m a ID ~Com . Comments Stain ess tee Outer AG2A Ca sule Bottom Gd 82 Gd Inner Ca sule Bottom R-3 Fe 07448 26/17944 Ti S Folded Wire acer u t- 49.149 99.999%

Ni SE Roll 46.011 99.999%

A rapped Cu 19047 CPI 3054 60.330 99.999% RM S ectral Set 19046 Cat 614 39.602 99.917%

Co/Al 44451 SC Bar 2 9.998 0.506%

X . 90" V 235-U 88705 264C Ca sule 1.903 87.97% Series 4 1

238-U 77014 ES-Z Ca sule 7.561 87.25% Series 8 5 .

237-N 88705 24HP-U Ca sule 6.612 88.3% Series 6 Ti S acer 287.55 0.19" OD X 0.375" OL Al Shim S acer Gd Inner Ca sule To Al 'Shim S acer

TABLE 7 (CONTtD)

TASK/ID: 854HSC-9010 Task 1/Nine Hi le Point SS Capsule Number: AG2A RH Set ID: R-3 HAFH Set ID: SSTR ID: SSTR-TH ID: Cover: Gd 82 Location: Hid Purchase Order Batch Haterial Sample Elemental Set ID Hateri al Number Number Descri tion ~Wt. m e ID ~Cem . Comments Lar ge e eat Sink S acer SS Outer Ca sule 2955.22 0.257" OD X 0.832" OL To

TABLE 8 TASK/ID: 85-WHSC-9010 Task 1/Nine Mile Point SS Capsule Number: AG2B RH Set ID: HAFH Set IO: H3 SSTR ID: ~7 8 5 SSTR-TH IO: Cover: Gd d3 Locatltoo: Mid Purchase Order Batch Material Sample Elemental Set ID Material Number Number Descri tion ~Wt. m a ID ~Com . Comments Stain ess tee Outer AG28 Ca sule Bottom Gd 83 Gd Inner Ca sule Bottom H-3 Ti S acer RI RI-7* 3 Metal Pcs Al Wra ed 3.32 NM-BE-8 Be Rl-4*

.0 . 'ng e Al rapped Fe RI Fold Wire 62.351 HAFM Set

.8 ing e Ni RI RI-11A* Fold Wire 123.889 ng e Cu 19047 CPI-3054 Fold Wire 31.868 99.999K ED-9 5 . '

nge 19045 Se Roll 2 Fold Wire 8.945 ~

99.999K Ti S acer 255.85 0.175" 0 X 0.358" OL Mica Huscovite SSTR Label Down Al Wra ed TR 237-N 24HP Oe osit Label Down Table 12 203 Mica Muscovite SSTR Label Down Al Wra ed SSTR 238-U 49096 ES-Z De osit Label U Table 12 229 Mica Huscovite SSTR Label Down Al Wra ed 235-U 49096 314A Oe osit Label Down Table 12 217

• Batch No. given by RI for Boron analysis.

TABLE 8 (CONT'D)

TASK/ID: 85-WHSC-9010 Task 1/Nine Mile Point SS Capsule Number: AG28 RH Set ID: HAFH Set ID: H3 SSTR ID: 7,8,5 SSTR-TH ID: Cover: Gd f3 Location: Mid Purchase Order Batch Haterial Sampl e Elemental Set ID Material Number Number Descri tion ~ltt. m e ID ~Cem . Comments 0.16 Al tfvrappe Fe x Lot 11 Disk 81.38 99.772K Gradient Al Shim S acer Gd Inner Ca sule Al Shim S acer Large T eld eat Sink S acer SS Outer Ca sule 3050.88 0.253"OD X 0.830"OL To

0 TABLE 9 TASK/10: 85-WHSC-9010 Task 1/Nine Mile Point Unit 1 SS Capsule Number: AB2 RN Set IO: R4 HAFN Set ID: H4 SSTR ID: 5 SSTR-TN ID: AB2 Cover: Bare Looattoo.

r Purchase Order Batch Material Sample Elemental Set ID Material Number Number Oescri tioo ~Wt. m s ID ~Com . Comments Stain ess tee Outer AB2 Ca sule Bottom AB2 Mica uscovite S acer uartz Supersil Label U 13 238-U ES-Z Label U Table 12 230 uartz Supersil Label Down 12 uartz Su ersil Label U Al Wra ed

-T Mica 1uscovite Label Down 13 Monitors 238-U ES-Z Label U Table 12 233 Mica 'luscovite Label Down 12 Mica tuscovite Label U Ti S acer Be RI RI-7* 3 Pcs Metal Al Wra ed 3.09 NH-BE-2 0.30'X .87'ingle A Wrapped Fe RI RI-11A* Fold Wire 62.333 HAFH Set angl e Ni RI RI-4* Fold Wire 120.157 HEDL >ng e Cu 19047 CPI 3054 Fold Wire 30.844 99.999Ã

• Batch No. given by Rl for Boron analysis.

0 TABLE 9 (Cont'd)

TASK/ID: 85-WHSC-9010 Task 1/Nine Hile Point Unit 1 SS Capsule Number: AB2 RH Set ID: R4 HAFH Set ID: H4 SSTR ID: 5 SSTR-TH ID: AB2 Cover: Bare Location: Hid Purchase Order ( Batch Haterial Sample Elemental Set ID Haterial Number ) Number Descri tion ~Wt. m e ID ~Com . Comments (NH)ii904 )* 0. . Sin- Al Wrappere Al 19045 SE Roll 1 le Fold Wire 8.783 99.999K HAFH Set Ti S acer 250.83 0.180"OD X 0.376"OL Hi ca 1uscovite SSTR Label Down Table 12 Al Wra ed 235-U 49096 314A De osit Label U 214

.164"X0.031'ta R4 Fe x Lot 11 Disk 81.9 99.772K Al Wra ed

.035 . 90 V Series 2 235-U 88205 264C Ca sule 1.207 Fo Co/Al 69-89-0885 SRR 953 Wire 9.123 0.116K Al Wra ed Al Shim S acer arge e eat Sink S acer SS Outer Ca sule 1912.45 0.253"OD X 0.829"OL To

TABLE 10 TASK/ID: 85-WHSC-9010 Task 1/Nine Hile Point Unit 1 SS Capsule Number: AG3A RH Set ID: R5 -

HAFH Set ID: SSTR ID: SSTR-TH ID: Cover: Gd 84 Location: Bottom Purchase Order Batch Haterial Sample Elemental Set ID Haterial Number Number Oesort tion ~Nt., m s ID ~Com . Comments State ess tee Outer AG3A Ca sule Bottom Gd 83 Gd Inner Ca sule Bottom R5 Ti S acer At u Fe 07448 26/17944 Folded Wire 54.735 99.999K ut-Se Roll Folded Wire 46.351 99.999K

~ U r appe Cu 19047 CPI 3054 Folded Wire 62.060 99.999K S ectral Set .

~ \t ~ u 19046 Cat 614 Folded Wire 41.232 99.917K u

Co/Al 44451 Sc Bar 26 Folded Wire 9.933 0.506K Al Wra ed Ti S acer 225.8 0.180" OD X 0.372" OL Al Shim S acer Gd Inner Ca sule To Al Shim S acer arge e eat Sink S acer SS Outer Ca sule 2894.73 0.254" OD X 0.834" OL To

TABLE 11 TASK/ID: 85-WHSC-9010 Task 1/Nine Mile Point Unit 1 SS Capsule Number: AG3B RM Set ID: HAFH Set ID: H5 SSTR ID: SSTR-TH ID: Cover: Gd $ 14 Location: Bottom Purchase Order Batch Hateri al Sample Elemental Set ID Materi al Number Number Decent t1on ~Mt. m e ID ~Com . Comments Stain ess tee Outer AG3B Ca sule Bottom Gd 814 Gd Inner Ca sule Bottom H5 Ti S acer Be RI RI-7* 3 Pcs Metal Al Wra ed 3.55 NH-BE 12

. 0 . Single A rappe Rl RI-4* Fold Wire 63.053 HAFH Set ng e RI R I-11A* Fold Wire 129.574 L- ng e Cu 19047 CPI 3054 Fold Wire 37.833 99.999K

- 9 ng e Al 19045 Se Roll Fold Wire 9.166 99.999K Ti S acer 248.5 0.175" 0 X 0.355" OL eta A rapped Fe Rx Lot 11 Disk 81.49 99.772K Fe Gradient Al Shim S acer Gd Inner Container To Al Shim S acer arge e eat Sink S acer SS Outer Ca sule 3001.65 0.253" 00 X 0.834" 00

• Batch No. given by RI for Boron analysis.

I1 TABLE 12 TASK/ID: 85-WHSC-9010 TASK 1/Nine Hile Point Unit 1 SS Capsule Number: AB3 RH Set ID: R6 HAFH Set ID: H6 SSTR ID: SSTR-TH ID: Cover: Bare Location: Bottom Purchase Order Batch Haterial Sample Elemental Set ID Haterial Number Number Descri tion ~Mt. m s ID ~Com . Comments Stain ess tee Outer AB3 Ca sule Bottom H6 Ti S acer Be R2 RI-7* 3 Hetal Pcs Al Wra ed 2.94 NH-BE-9 ingle Fe R2 RI-11A" Fold Wire 62.815 sng e A rappe R2 RI-4* Fold Wire 126.896 HAFH Set E L- ng e Cu 19047 CPI-3054 Fold Wire 31.886 99.999K

>ng e Al 19045 SE Rol l Fold Wire 9.016 99.999K Ti S acer 258.25 0.177"00 X 0.368"OL eta Fe x Lot 11 Disk 81.32 99.772K Al Wra ed Co/Al 69-89-0885 SRH 953 Wire 8.863 0.116K Al Wra ed Al Shim S acer Lar ge e d eat Sink S acer SS Outer Ca sule 1775.05 0.256"00 X 0.834"OL To

• Batch No. given by RI for Boron analysis.

TABLE 13 SOLID STATE TRACK RECORDERS (SSTR)(")

235-U 238-U 237-N SS CAPSULE A DE IT MA DEPO I A T ID NO. (pg) ID NO. (pg) ID NO. (pg) ID AB2 214 0. 702 1 AB2B 217 2 03 2 229 91.19 3 203 2.04 4 CG2B 221 0.447 5 232 43.71 6 202 3.91 7 CB2 216 1-21 8 SOLID STATE TRACK RECORDER TEMPERATURE MONITORS (SSTR-TM)(")

238-U SS CAPSULE D OSIT DEPOSIT MA S T lO K. MATERIAL/I0 AB2 230 33.45 quartz/13 233 68.53 Mica/13 CB2 225 23.39 quartz/23 226 12.09 Mica/23

1) All deposits are made on 0.168" Ni backing and have a deposit diameter of 0.136". The estimated uncertainties associated with the SSTR mass values are within +5% (1a).

C-18

HAFM DOSIMET TS Be(1 2) Ni(2,3) Fe(2,3) Cu(2 3) Al(2,3)

SS Capsule Al Capsule Lot RI-7 Lot RI-4 Lot RI-11A Lot HEDL-3054 Lot HEDL-19045 ID ID ~m ~m m m m AG1B H-1 4.25 125.384 62.695 33.189 9.301 AB1 H-2 2.75 124.704 64.514 30.008 9.068 AG2B H-3 3.32 123.889 62.351 31.868 8.945 AB2 3.09 120.157 62.333 30.844 8.783 AG3B H-5 3.55 129.574 63.053 37.833 . 9.166 AB3 2.94 126.896 62.815 31.886 9.016 CG1B H-7 2.25 121.639 61.938 30.743 8.708 CB1 H-8 2.02 120.041 60.452 30.597 8.454 CG2B H-9 2.16 114.982 58.747 28.795 8.430 CB2 H-10 1.90 119.615 59.479 33.123 8.495 CG3B H-11 2.32 121.077 61.289 30.358 8.594 CB3 H-12 2.13 120.439 61.109 30.848 8.535 Wire Diameter .0.04" 0.03" 0.02" 0.02" B Content (ppm) 8.9 0.004 0.0046 <0.0002 0.075

1) Each Be sample consists of 3 small pieces of metal weighed and wrapped in a single Al foil package by RI.

2) gA analysis for Boron content of the various HAFM materials may be found in NUREG 3746 Vol. 1-HEDL-TME 84-20, Semi-Annual Pro ress Re ort, October 83-March 84, (November 1984) pp. RI1 through RI9,

3) All wires are single folded with total length of &.87".

TABLE 15 QUARTZ ENCAPSULATED MELT WIRE (MW) TEMPERATURE MONITORS (TM)

Melt Temp. Wire Dia. Wire Length Wire Wt. Quartz Dia. Quartz Length

~Ca sule oF (in.). (in. ~(m ) (in.) (in.

536 0.030 0.5 172.5 0.235 1.387 558 0.090 0.5 563.0 ,0.235 1.430 580 0.084 0.5 505.2 0.235 1.620 588 0.084 0.5 497.0 0.235 1.990

.598 0.083 0.5 524.0 0.235 2.067 536 0.030 0.5 174.0 0.235 1.199 558 0.090 0.5 541.6 0.235 1.312 580 0.084 0.5 509.6 0.235 1.686 588 0.084 0.5 525.2 0.235 1.878 598 0.083 0.5 502;0 0.235 2.135

TABLE 16 CAPSULE LOADING SEQUENCE "C" SERIES (-24 EFPY EXPOSURE)

(From Bottom of SS or Gd Capsule)

Al Capsule "Ial Top CG1A 0.254 0.832 2.8753 R7 RM 0.180 0.387 233.35 CG18 '.252 0.832 2.9847 H7 HAFM 0.180 0.362 255.28 Fe Grad.

CB1 0.255 0.831 1.7688 H8 MAFH 0.180 0.368 247.25 R8 RH Hid CG2A 0.255 0.836 2.9598 R9 RH 0.180 0.380 280.9 CG2B 0.254 0.830 3.9016 SSTR-TH H9 HAFH 0.180 0.357 234.1 7

8 SSTR 5

Fe Grad.

CB2 0.254 0.833 1.0617 CB2 SSTR-TH H10 HAFH 0.178 0.380 248.23 5 SSTR 0.193 0.014 R10 RH Bottom CG3A 0.255 0.833 2.8687 R11 RH 0.180 0.392 222.0 CG3B H11 HAFH 0.175 0.341 249.35 Fe Grad.

CB3 0.255 - 0.836 1.7624 H12 HAFH 0.180 0.372 245.55 R12 RH o os meter mater a s are closer than 0.10" from outer SS capsule top. Therefore, the top can be parted off up to 0.10" from top. Care must be taken that no burrs exist nor the open end swaged or the inner capsules may not shake out.

2) It is recommended the Gd capsules actually be cracked open (i.e., use "nut cracker" on lower end of capsule).

The reason for this is that several of the dosimeter pockets are quite a tight fit and possibly would be damaged in the attempt to pull them out of the capsule or to cut open the bottom and push them out.

3) No Al material should be discarded without close observation as to whether it is simply spacer material or wrapped dosimetry material. Fe gradients are wrapped individually and though marked, Fe could be mistaken for spacer material. Both bare spectral Co/Al and Fe 235-U are wrapped'n Al and are unmarked and could therefore be mistaken for spacer material.

TABLE 17 TASK/ID: 85-WHSC-9010 Task 1/Nine Mile Point Unit 1 SS Capsule Number: CG1A RN Set ID: RT NAFM Set ID: SSTR ID: SSTR-TM ID: Cover: Gd 55 LocatIon: ~To Purchase Order Batch Material Sample Elemental Set ID Material Number Number Descrd tIon ~Wt. m s ID ~Com . Comments Stain ess tee Outer CG1A Ca sule Bottom GD 85 Gd Inner Ca sule Bottom Rj Ti acer S

ut-Fe Ni 07448 so 1

sl 26/17944 Sf Roll Folded Wire

.02 "X-1.5 Folded Wire ul t- 51.216 45.727 99.999K 99.999%

~ u t A rappe Cu 19047 CPI 3054 Folded Wire 60.127 99.999K S ectral Set 2'X>> . Ultl-19046 Cat 614 Folded Wire 34.087 99.917K

. 2'-. ut-Co/Al 44451 Sc Bar 2 Folded Wire 9.374 0.506K Al Wra ed Ti S acer 233.35 0.180"OD X 0.387"OL Al S acer Shim Gd Ca sule To a

Al S acer Shim Large T eld eat Sink S acer SS Outer Ca sule 2875.25 0.254"ODX0.8330L To

0 0

TABLE 18 TASK/ID: 85-WHSC-9010 Task 1/Nine Hile Point Unit 1 SS Capsule Number: CG1B RH Set ID: HAFH Set 10: H-I SSTR IO: SSTR-TH IO: Cover: Gd F12 Locatson: ~To Purchase Order Batch Hateri al Sample Elemental Set ID Haterial Number Number Oescrs tson ~Nt. m s IO ~Com . Comments Stasn ess tee Outer CG18 Ca sule Bottom Gd 12 Gd Inner Ca sule'i Bottom H7 S acer eta rappe Be RI RI-7* in Al 2.25 NH-BE-7 sng e rappe Ni RI RI-4 Fold Wire 121.639 HAFH Set

.3 . nge Fe Rl R I-11A* Fold Wire 61. 938 ng e Cu 19047 CPI 3054 Fold Wire 30;743 99.999K DL- 0 . . " sng e Al 19045 SE Roll Fold Wire 30.597 99.999K Ti S acer 255.20 0.180"ODX 0.362"OL

.16 'X . 1'ta A rappe x Lot 11 Disk 80.59 99.772K Gradient Al Shim S acer Gd Inner Ca sule To Al Shim S acer arge e eat Sink S acer SS Outer Ca sule 2984.72 0.252"ODX0.833"OL To

• Batch No. given by RI for Boron analysis.

f9 TABLE 19 TASK/ID: 85-WHSC-9010 Task 1/Nine Mile Point Unit 1 SS Capsule Number: CB1 RH Set ID: ~RB HAFH Set ID: H-8 SSTR ID: SSTR-TH ID: Cover: Bare Locatton: ~To Purchase Order Batch Hater 1 al Sample Elemental Set IO Material Number Number ~baser) t3on ~Wt. m s ID ~tom . Comments SS Outer Ca sule H-8 Ti S acer Be RI RI-7* 3 Pcs Metal Al Wra ed 2.02 NM-BE-11

.04 0.8 Single Ni RI RI-4* Fold Wire 120;041

.8 S ng e A Wrapped Fe RI RI-11 A* Fold Wire 60.452 HAFH Set ot ng e Cu 19047 CPI-3054 Fold Wire 30.597 Lot HEOL-190 4 . 2 . in-Al 19045 SE Roll 1 le Fold Wire 8.454 Ti S acer 247.25 0.180"ODX0.368"OL

. 6 . 1 eta In ividua y Al RB Fe x Lot 11 Disk 81.00 99.772K Wra ed RH lng e nmar ed Co/Al 69-89-0885 SRH 953 Fold Wire 8.562 0.116K Al Shim S acer arge T> e eat Sink S acer Outer apsu e Ta 1768.79 0.255"ODX 0.831"OL

" Batch No. given by RI for Boron analysis.

TABLE 20 TASK/ID: 85-WHSC-9010 Task 1/Nine Mile Point Unit 1 SS Capsule Number: CG2A RM Set ID: R9 HAFM Set ID: SSTR ID: SSTR-TM ID: Cover: Gd IG Locatton: Wid Purchase Order Batch Material Sample Elemental Set ID Material Number Womber Oescrt t ton ~Wt. m s ID ~Com . Comments Stain ess teel Outer CG2A Ca sule Bottom Gd 86 Gd Inner Ca sule Bottom R-9 Ti S acer

~ U 't 1-Fe 07448 26/17944 Folded Wire 47.155 99.999K

~ U Ni SE Roll Folded Wire 44.984 99.999K s e u 't Cu 19047 CPI 3054 Folded Wire 60.013 99.999K v ~ u A rappe 19046 Cat 614 Folded Wire 33.717 99.917K S ectral Set u ti-Co/Al 44451 SC Bar 2 Folded Wire 8.583 0.506K Al Wra ed 235-U 238-U 88705 77014 264C ES-Z 0.. Ca Ca sule sule 5 V 1.775 7.552 87.97K 87.75K Series Series 3

B 237-N 88705 24HP-U Ca sule 6.555 88.3X Series 6 Ti S acer 280.9 0.180"ODX 0.380"OL Al Shim S acer Gd Inner Ca sule To

4'h

.li

TABLE 20 (Cont'd)

TASK/ID: 85-WHSC-9010 Task 1/Nine Mile Point Unit 1 SS Capsule Number: CG2A

'RH Set ID: RD HAFH Set ID SSTR ID: SSTR-TH ID: Cover: Gd 96 Location: Hid Purchase Order Batch Material Sample Elemental Set IO Material Number Number ~Deecri tion ~Mt., m e ID ~Com . Comments Al Shim S acer arge e eat Sink S acer SS Outer Ca sule 2959.85 0.255"ODX 0.836"OL To

TABLE 21 TASK/ID: 85-WHSC-9010 Task 1/Nine Mile Point SS Capsule Number: CG28 RH Set ID: HAFH Set ID: H-9 SSTR ID: ~7G.S SSTR-TM ID: Cover: Gd d7 LocatIoo: MId Purchase Order Batch Hater) al Sample Elemental Set ID Material Number Number Descri tion ~Wt. m a ID ~Com . Comments Stain ess tee Outer CG2B Ca sule Bottom Gd 87 Gd Inner Ca sule Bottom H9 Ti S acer Be RI RI-7* 3 Pcs Metal Al Wra ed 2.16 NH-BE-6 ng e Ni RI RI-4* Fold Wire 114.982

-.8 ng e A rapped Fe RI RI-11A* Fold Wire 58.747 HAFH Set 0- ng e Cu 19047 CPI 3054 Fold Wire 28.795 99.999K EO- ng e Al 19045 SE Roll 1 Fold Wire 8.430 99.999K Ti S acer 234.1 0.180"ODX 0.357"OL ee Tab e A rapped Mica Muscovite SSTR Label Down 20 SSTR 237-N 24HP De osit Label U Table 12 202 ee a e rappe Mica Huscovite SSTR Label. Down 20 SSTR 238-U 49096 2718 De osit Label U Table 12 232 ee a e A rapped Mica Muscovite SSTR Label Down 20 SSTR 235-U 49096 314A De osit Label U Table 12 221

• Batch No. given by RI for Boron analysis.

I' TABLE 21 (Conttd)

TASK/ID: 85-WHSC-9010 Task 1/Nine Hile Point Unit 1 SS Capsule Number: CG2B RH Set ID: HAFH Set ID: H-9 SSTR ID: ~7 8 5 SSTR-TH ID: Cover: GD d7 Locattoo: Mtd v

Purchase Order Batch Hateri al Sample Elemental Set ID Hater ial Number Number Descri tion ~Mt. m a ID ~Com . Comments 0.16 . etal Al Wrapped R9 Fe x Lot 11 Disk 81.21 99.772K Gradient Al Shim S acer I

Gd Inner Ca sule To O Al Shim S acer I

Large 1 e eat 00 Sink S acer SS Outer Ca sule 3061.69 0.254"ODX 0.829"OL To

~ a TABLE 22 TASK/ID: 85-WHSC-9010 Task 1/Nine Mile Point Unit 1 SS Capsule Number: CB2 RH Set ID: 'R1D WRFH Set ID: 111D SSTR ID: 5 SSTR-TH ID: CB2 Cover: Bare Location: Hid Purchase Order Batch Material Sample Elemental Set ID Material Number Dumber Descri tion ~Wt. m s ID ~Com . Comments Stain ess tee Outer CB2 Ca sule Bottom CB2 Hica uscovite S acer uartz 3 Supersil Label Down 238-U ESZ Label U Table 12 225 uartz Supersil Label Down 12 Al rapped uartz Supersil Label U SSTR-TH on tors Mica uscovite Label Down 13 238-U ESZ Label U Table 12 226 Mica uscovite Label Down 12 Mica uscovite Label U H10 Ti S acer cs eta rapped Be RI Rl 7* In Al 1.90 NM-BE-10

>ng e Fe RI RI-11A* Fold Wire 59.479

. 0 . >ng e Ni Rl RI-4* Fold Wire 119.615 Al Wra ed HEOL 05 * ;20' 'in- HAF Set Cu 19047 CPI 3054 le Fold Wire 33.123 99.999

TABLE 22 (Cont'd)

TASK/ID: 85-WHSC-9010 Task 1/Nine Hile Point Unit 1 SS Capsule Number: CB2 RN Set ID: R10 HAFN Set ID: HIO SSTR ID: 5 SSTR-TN ID: CB2 Cover: Bare Location: Nid Purchase Order Batch Haterial Sampl e Elemental Set ID Haterial Number Number Descri tion ~Wt., m a ID ~Com . Comments callw

• 0.0 . ingle Al 19045 SE Roll 1 Fold Wire 8.495 99.999K Ti S acer 248.23 0.118" ODX0.380 "OL 5 Hica SSTR SSTR Label Down Al Wra ed TR Set 235-U 49096 314A De osit Label U Table 12 216 0.16 X .031 eta R10 Fe Rx Lot 11 Disk 81.73 Al Wra ed o Pac age 235-U 88705 264C Ca sule 1.187 87.97K Series 5 o d Co/Al 69-89-0885 SRH 953 Wire 8.158 0.116K Al Wra ed Al Shim S acer arge e eat Sink S acer SS Outer Ca sule 1901.65 0.254"ODX 0.833"OL To

• Batch No. given by Rl for Boron analysis.

TABLE 23 TASK/ID: 85-WHSC-9010 Task 1/Nine-Mile Point Unit 1 SS Capsule Number: CG3A RN Set ID: R11 HAFN Set IO: SSTR ID: SSTR-TN IO: Cover: Gd IO Location: Bottom Purchase Order Batch Material Sample El cmental Set ID Material Number Number Descri tion ~NI., m e ID ~Com . Comments Stain ess tee Outer CG3A Ca sule Bottom Gd 89 Gd Inner Ca sule Bottom R11 Ti S acer u t Fe 07448 26/17944 Fold Wire 52.113 99.999K

~ U A rappe Se Roll Fold Wire 52.643 99.999K S ectral Set

~ 02' ~ 5 u tl-19047 CPI 3054 Fold Wire 60.402 99.999K Cu ut-19046 Cat 614 Fold Wire . 39.413 99.917K u tl" Co/Al 44451 SC Bar 2 Fold Wire 9.695 0.506K Al Wra ed Ti S acer 222.0 0.180"ODX0.3920L Al S acer Shim Gd Inner Ca sule To Al S acer Shim Large Ti eld eat Sink S acer SS Outer Ca sule 2868.72 0.255"ODX0.8330L To

TABLE 24 TASK/ID: 85-WHSC-9010 Task 1/Nine Mile Point Unit 1 SS Capsule Number: CG3B RM Set ID: HAFM Set ID: H11 SSTR ID: SSTR-TM ID: Cover: GD 813 Location: Bottom Purchase Order Batch Material Sample Elemental Set ID Material Number Number Oescrt t1oo ~Wt. m s ID ~Com . Comments Stain ess tee Outer CG3B Ca sule Bottom Gd 13 Gd Inner Ca sule Bottom H-11 Ti S acer Be Rl RI-7* 3 Pcs Metal Al Wra ed 2.32 NM-BE-5

'ingle Fe RI Rl-11A* Fold Wire 61.289 X . nge A rappe Ni RI RI-4 Fold Wire 121.077 HAFM Set

.8 'ing e Cu 19047 CPI-3054 Fold Wire 30.358 99.999K D- sng e Al 19045 Se Roll Fold Wire 8.594 99.999K Ti S acer 249.35 0.175"ODX0.341"OL "X .031 eta rappe Fe x Lot 110 Disk 81.82 99.772K Gradient Al Shim S acer Gd Inner Ca sule To Al Shim S acer Large T eld eat Sink S acer SS Ca sule 2989.80 0.253"ODX0.830"OL To

0 0

TABLE 25 TASK/ID: 85-WHSC-9010 Task 1/Nine Hile Point Unit 1 SS Capsule Number: CB3 RH Set ID: R12 - HAFH Set ID: H12 SSTR ID: SSTR-TH ID: Cover: Bare Location: Bottom Purchase Order Batch Haterial Sample Elemental Set ID Haterial Number number Descri tion ~Wt. m s ID ~Com . Comments Stain ess tee Outer CB3 Ca sule Bottom H-12 Ti S acer Be Rl RI-7* 3 PCS Metal Al Wra ed 2.13 NH-BE-4

&.8 'ingle Fe RI RI-11A* Fold Wire 61.109 sng e rappe Ni Rl RI-4 Fold Wire 120.439 HAFH Set

.0 0 X .87'sng e Cu 19047 CPI 3054 Fold Mire 30.848 99.999K E ng e Al 19045 SE Roll Fold Wire 8.535 99.999K Ti S acer 245.55 0.180"ODX0.372"OL eta Fe Rx Lot 11 Disk 81. 8 99.772K Al Wra ed

.020 ' . wo-Co/Al 69-89-0885 SRH 953 Fold Mire 8.766 0.116K Al Wra ed Al S acer Shim arge > e eat Sink S acer SS Outer Ca sule 1762.44 0.255"ODX0.836"OL To

• Batch No. given by RI for Boron analysis.

TABLE 26 NINE MILE POINT UNIT 1 DOSIMETRY gA INFORMATION Activation RH/HAFH Dosimeter Materials Batch/ Target Isotopic Element w/o Boron Diameter Dosimeter Form Lot No. PO No. ~lento e Abundance Vendor Vendor HEOL ~Wt m or Thickness (at~om ractson)

Al M Wire SE Roll 1 19045 27 1.000 Semi. Ele. 99.999 1.48 0.020" D Be Metal RI 7 Rockwell 9 1.000 8.9 Co/Al H Wire 'RM 953 89095 59 1.000 NBS 0.116 0.117 1.23 0.020" D H Wire SC Bar 26 44451 59 1.000 Sigmund Cohn 5 0.506 0.020" D CU H Wire CPI 3054 19047 63 0.6917(1) Comico-Am. 99.999 ~30ppb Co 0.0002'.020" D Fe H Wire RI 11A Rockwell 56 0.9172(30) 0.0046 0.030" 0 H Wire 26/17944 07448 54 0.058(1) Materials Res. 99.999 0.015 0.020" T H Disc Rx 110 54 0.058(1) Reactor Exp. 99.772 0 031st T Ni H Wire RI 4 Rockwell 58 0.6827(1) 0.004 0.040" D II II H Wire SE Roll 2 1 58 0.6827(1) Semi. Ele. 99.999 0.015 0.020" D H Wire Rx 139W 19046 46 0.080(1) Reactor Exp. 99.917 0.020" D A) Values for natural isotopic abundances are from "Isotopic Composition of the Elements 1983", Pure & Appl. Chem.", Vol. 56.

pp 676-694 (1984). Error assignments in the last digits of the values are given in parenthesis [1 .e., 0.9172 (30) is 0.9172 k 0.0030].

B) These reported Boron content values along with a discussion of the analytical techniques may be formed in NUREG/CR-3746, Vol. 1, HEDL-THE 84-20, "LWR Pressure Vessel Surveillance Dosimetry Improvement Program Semi-Annual Progress Report, October 1983-March 1984, (November 1984), pp RI-2/9, (1984).

TABLE 26 (Con't)

FISSIOHABLE RH t SSTR OOSIHETER HATERIAL Ana) tlcal Lab Results Batch Hire Anal. Element Isoto lc w/o Capsule Mal I ttesfeeter tie. tete ~Dte. Ie. PO Ho. Lab p w/o 235-U 264C Ox )de Hire 0.0186 88705 ORHL 8.68 87.97 ~0.0005 0.034 99.89 0.025 0.053 0.007 HEDL - 87.37 IOHS ~0.0037 0.0294 99.901 0.022 0.043 IIEOL 88.00 Titr .

235-U 314A Oxide 49096 ORML (0.001 0.036 99.940 0.011 0.013 SSTR 238-U ES-ZOxide Mire 0.0175 77014 ORHL 9.62 87.25 (0.0001 (0.0001 0.0012 %0.0001 99.999 0.007 LLNL - 87.61 HEOL - 87.87 <0;0028 ~0.0035 ~0.0039 4).0032 99.987 238-U 2718 Oxide 49096 ORHL ~ =0 ~ 0001 (0.0001 0.0006 (0.0001 99.999 SSTR 235 236 237 238 239 237-Hp 24HP-U Oxide Hire 0.173 88705 ORHL 8.14 88.3 ~99.999 (Only Isotope Detected) V 0.007 24HP Ox lde ORHL <0.0005 (0.0005 <<99.99 S0.003 S0.003 SSTR

0 0

APPENDIX 0 AS-BUILT PHOTOGRAPHS OF ADVANCED DOSIMETRY FOR CAPSULES A'nd C'

~l) " l " "'"'""N~l'r gj 'rll -: - '.j~ "-" 'l pl~ r "'l"m ~rior'BP~r Sr ~~~r~

r

• ~

p,rr, r 4j g~k r

r pe ~ yr

~

'pal

+Pi. r ~

g r FIGURE 1. Stainless Steel.- (SS),'-'.Encapsul ated. Dosimetry Sets.

D-2

0 0.060

~ FILLET WELD (TYP)

SS END CAP 0.100 Ti DISC (0.001 THICK) WELD HEAT SINK SPACER CRUMPLED AI FOIL 0.030 FILLER PLUG 0.030 Gd CAP aC

~

~

~

~

~

I'

~

SLIP FIT

~ ~

h~

0.830 0.510 'r'. ~t 1

fP

~ ~ ~ \

DOSIMETER SPACE

~ ~

4o' Gd LINER BODY

~

0.220 OD x 0.020 WALL 0.040 I SS CONTAINER BODY 0.060 0.250 OD x 0.010 WALL o.ooo SS END CAP NOTE: ALL DIMENSIONS ARE INCHES HEDL 8704.034.2 FIGURE 2. Stainless Steel Container and Gadolinium Liner Design.

D-3

4)

CAPSULE 'C 'SET'. .CAPSULE A SET CG1A, AG1A AGIB CB1.

-CG2A CG2B CB2 P~[j j AB2 CG3A AG3A CG38 AG3B CB3 BEAM INTENSITY OPTIMIZED BEAM INTENSITY OPTIMIZED

'TO SHOW INDIVIDUAL ~ TO SHOW SS WELDS AND DOSIMETERS IN THE.SIX INDIVIDUALDOSIMETERS IN Gd CAPSULES THE THREE BARE. CAPSULES HEDL8603-208 FIGURE 3. Radiographs of Encapsulated Dosimeters.

D-4

~

4.

,'1q ~ I+(

C ~

A~i

wc'

':.- "~~<+ta 4 +~'~+4~~g~" >>;

'>>~4'~",...'-;

'.,;,Cp4 gh~'94:4 'Ye ym FIGURE 4. Dosimetry Capsules AGlA, AG2A, AG3A, CG1A, CG2A and CG3A.

>leg 8602722-10 D-5

4 I

4 11, 4

1 ~l'Sp C

I

r

. r, Prr IV~, g,y'q4<)gf4rr'r~&C'$~, r > '~ y~~ @PE.r ~

OUTER SS CAPSULE

'+A

~rPPi'ger:

lt fr re r > ~ .

r S

~.,""Qg L

r

~~< WR-.. 444AhLeVp~">p,.L.A-.y~r'

. '-'M+",<I<'$r. ~q~'"+@'~~I~'~re!~>:<i,,':rrI TOP:; ) TI WELO HEAT SINK p: L 'kp r"'. Irr

~bMm5Ipr')AP,~, Pkg" ra,.4<I, ~ 49+@" P.gal% .'>;,L I,:. 'Wk<j%+e " 'rP$

~~P'~k 4'~'m~~~.r 4k

'A r ~ .y<~(<<r e ~~ E~A4~ rd.~~44~W~

4 X2; 'Vr ' 'r . r- ~>> ~~ kf"-~rrITp <<yA + ' ~P, *

%f g Q r

r

'.rNQ+gg~~g~~)~Q~Qt ~%

r, j/f'--. '- -

4)fggg'Pg rr ..,, A

.... 'IE,dj, .MMr@~~';'; ~4[5)',y>~Q),,..Gd I,'NNER CAPSULE

FAA+@yp : 1t. ,/

~>@g~1Yfgr.'"I 4

1NR4PPED H4l IVl SET "..,:

..'ORE

5. Dosimetry Capsules AGlB, AG38, CGlB and CG3B. Neg 8602722-7 D-6

WRAPPED Fe GRADIENT:fkga+<<p$

>P'..~.*,,","~'.~I,

~.4~<4~:.)~g>~Q.<P~

WRAPPED:237-Np SSTR Fi'IQt',aA'Chi~4>> ':,",;4"~jj."i'jjlI~gj@hlgi:.:,j::::j'i X .A ~wi','>>!,','::WRAPPED HAFM SET > ';;ij'j-,'t->',"Q'-~:,!~~~

Dosimetry Capsules AG28 and CG2B. Neg 8602722-8 D-7

4~

L lg

'I

• a 6 l ,6! i ~ ~a t ~ ~

1

a.<~> i9a

$w~<<%~p@yijg,

+p$p$ y'r,++/

~>>h%

+

i,g'P g)%%a~~

Q

<<4 g'lv,. <<+IL W<<

ew <>>

"<<44~~PPPA+,Gg~<<@!~ggyp+~;;~'.:.,,q p4.<<

$ )'4 p:~.

p),.~.

P.,".f'~O.,

SS OUTER CAPSULE "e: .

W v

'<p'. =-- ()+~~>p~<~rhf g< '~'" +Q- "~  !"--( ~<<4" '<<"'g'~p>~~l, '.'8' 4~p<~-k~~~~ $"e'~ "~4% "'4'~<<~'~:

<4 + '~"4'~~ ~ "~g ~%@A<<<<~A~>~

>>sw$ <dq<<:w+4K+$ ~+A/g<p~i<<~$

~/~~4!p +'4~" g P>~v~

~ < <<cg,

~--"':-~

$ j':jhow wQ g '~~ g>> wf 'Y 4<<>>wr ">> >PYp)<'~ ~5'-gY, + g gg

-'~-'-.'-: "< -'c H*

~<~Q~w,; ~~<>>ivg~w < pj -~K~,y Yz-~y~~'~Yc. d~, &)M<<p4<<~tr~g) K gfg~q~p~ 8 WRAPPED Co/Al RM i,',",,f h

FIGURE 7. Dosimetry Capsules ABl, AB3, CBl and CB3. Neg 8602722-9 D-8

I

~C.

C~

4 t., ~

I!

0 ct ot

.;:"'g~g"-'yF~ "."-4y3";: "<~!".","'.'j"'.~'.".'-'-'>~~'j~"

,-.'-'>-.-';"::.'> WRAPPED SSTR-'TIVI SET fM"."g."'-",p+;".'-4~v<<.'&4~<

IGURE 8. Dosimetry Capsules AB2 and CB2. Neg 8602722-12 D-9

J I-C

~, > hah $ . ~ ~ >a<~+w~~&OW@8~

FIGURE 9. Bare RM Spectra1 Set. and Gradient RM. Neg 8602722-2

~,

C

'p 1 p

~ t,

))

. '~a(i,A+g

'" ~~'y"pk~~." ",;", ' p, ~ x .~, ".~g, -"j'm~+ *4~~& '.s,j~ zq -".Vjp 'pg FIGURE 10. Bare RH Spectral Set with 235-U. Neg 8602722-6

'A

'I I lE

s s s s

,aS, a

~

~

~

s

%~a%'

s s s

FIGURE 11; Gd-Covered RM Spectral Set. Neg 8602722-4

0 t

I ~,

f) g)a II p( g!)t gg ~

"~

>(~ <

"r ly II 1

"1

FIGURE 12. Gd-Covered RM Spectral Set with Fissionable Oosimeters.

Neg 8602722-3

II

'I ~~

~ gal y

r k,)

., ~. <<j'r;;, TOP kj,

'r h,: T; BPAOBBB +'<g BOTTOM . 4",~,;-:~QPr,i5:,cbrit..<j '

h>>

'hr r

h r p

r FIGURE 13.~ KAFM Dosimetry Set.~ Neg 8602722-5

P I

1 >ll 4

WRAPPED 235-u'-

k~Ppg~': i:" A~e@~w '~

,"bA ~w'. ~-y'k ~~~<$ p~pQ>>

wg4ggjg<<~

FIGURE 14. SSTR Dosimeters. Neg 8602722-1

>.S 4,

g.,lt p4 ~ h f

yl I

k 8

8

~

Hi

+"'* ' OUTER'Al. WRAPPlNG W

I

/

C.

h(

MICA SPACER'-,

' j+ ~,;4/' .'" '

QUARTZ SSTR-TM,

~~

r4y~@ggg, ~gy P~gg,p, p$

~ 'IGURE-'15.'STR TM Temperature Monitor- Sets. Neg 8602722-13 .

APPENDIX E PHOTOGRAPHS OF MELTWIRE TEMPERATURE MONITORS

w ll 1 il l~

QUARTZ CAPSULE

~QUARTZ CAPSUI.E BACK-FILLED WITH APPROXIMATELYONE ATOMOSPHERE OF HEUUM O.Z3"4I237" SPECIFIED LENGTHS TYPICAL MELTWIRE 668 F

+c I' Qg+F z CAPSULE A MEI TWIRE SET CAPSULE C MELTWIRE SET 1N Ol NN IN I FIGURE 1. quartz Encapsulated Melt Wire Temperature Monitors (TM).

IL lg A j

. K~+-; e~p f,:@ "'j;~ ., --;g'~-:~ k~"-~+$>~'~<

,m,y,~g~c~'~ "v:y ', -~

~. ~<<,P'. 'q~jP,*,~'~%" ~ ~, 'fi P, MQ.4q,,P '@ '

g.A ~+0,

/ '

• FIGURE z quartz Encapsulated Melt Wires. Typical of both Capsules A and C.

Ne9 8602722-11 E-3

APPENDIX F CHEMICAL ANALYSIS DATA FOR NINE MILE POINT UNIT 1

0 F.1 PLATE DATA The plate chemistry data were taken from References [LU64],

[ST84], [MA85a], and [MA87] .

F-2

'I CHEMICAL AHALYSIS RESULTS (WMAL) FOR NINE MILE POINT UNIT 1. MODIFIED A382B MATERIAL [MA87]

BASE FROM BASE 38 DEGREE CAPSULE (1)

E1A(A) Elh(B) E1C EBK E2U E31 STANDARD RELATIVE MEAN DEVIATION DEVIATION,X Fe MATRIX MATRIX MATRIX MATRIX MATRIX MATRIX Cu 8.244 8.243 8.243 8.258 8.248 8.246 8.883 1.366 8.244 Ni 8.512 8.514 8.5ee 8.55e s.sle 8.519 8.817 3.386 8.518 Cu(2) 8.241 8.885 2.875 8.241 Hi (2) 8.468 ~ 8.825 5.288 8.468 P 8.841 8.839 8.842 8.841 8.839 8.839 8.881 3.389 8.848 Mn 1.377 1.361 1.369 1.362 1.344 1.368 8.814 1.855 1.369 Co e.sle e.sle 8.818 8.811 8.818 8.818 4.816 8.818 Mo 8.433 8.435 8.421 8.466 8.432 8.436 8.815 3.454 8.436 V <8.885 <e.ees < 8.885 <8.885 <8.885 < 8.885 <8.885 Cr 8.186 8.186 8.186 8.186 8.186 8.189 8.912 8.186 Ti 8.881 8.881 <8.881 <8.881 <e.eel <8.881 <8.881 C(3) 8.286 8.286 S(4) 8.822 8.822 Si (6) 8.226 8.226 8.226 8.226 NOTES: (1) ALL MEASUREMEHTS BY ICAP UNLESS OTHERWISE HOTED (2) BY ATOMIC ABSORPTION (3) BY LECO COMBUSTIOH (4) BY TITRIMETRIC COMBUSTIOH (5) BY GRAVIMETRY

CHEMICAL AHALYSIS RESI.TS (WMAL) FOR HIHE MILE POINT UNIT 1. MODIFIED A382B MATERIAL [14AB7]

BASE FROM BASE 388 DEGREE CAPSI.E (1)

E42(A) E42(B) ETE EILl E1U E3T E3T(R) STANDARD RELATIVE MEAN DEVIATIOH DEVIATION,X Fe MATRIX MATRIX MATRIX MATRIX MATRIX MATRIX MATRIX Cu 6.233 8.23S 8.258 8.243 8.241 8.24S 8.247 B.ess 2.871 8.243 Hi 8.479 8.617 8.639 8.432 6.4SS 8.543 8.522 . 8.827 5.339 8.518 Cu(2) 8.238 8.865 2.161 8.238 Ki (2) e.473 8.817 3.694 8.473 P 8.833 8.848 8.845 8.839 8.834 8.843 8.845 8.884 9.947 8.841 Mn 1.324 1.334 1.378 1.337 1.362 1.394 1.377 8.825 1.869 1.368 Co 8.818 8.818 8.818 8.818 8.818 8.811 8.818 s.ese 3.726 8.818 Mo 8.48S 8.428 8.462 e.4es 8.41S 8.452 8.445 8.822 6.17S 8.431 V <S.SSS <e.eeS <S.SSS <e.ess <e.sss <B.ees <B.ess <e.ees Cr 8.185 8.18S 8.189 8.166 8.183 8.113 8.116 3.495 8.169 Ti 8.681 8.681 <8.861 <8.881 8.881 <8.881 < 8.881 <8.681 C(3) 8.218 8.218 S(4) 8.823 8.823 Si (5) 8.22S 8.22S 8.226 8.22S HOTES: (1) ALL L!EASURHlENTS BY ICAP UNLESS OTHERlISE KOTED (2) BY ATOMIC ABSORPTIOH (3) BY LECO COMBUSTIOH (4) BY TITRIMETRIC COMBUSTIOH (5) BY GRAVIMETRY

k CHEMICAL ANALYSIS RESIA.TS (WMAL) FOR HIHE MILE POIHT UHIT 1. MODIFIED A3828 MATERIAL [MA87]

UHIRRADIATED ARCHIVE PLATE (1) 025(A) 025(B) D21 Dbl STANDARD RELATIVE MEAN DEV IATIOH DEVIATION,X Fe MATRIX MATRIX MATRIX MATRIX Cu 5.173 8.188 b.17d $ .175 8.883 1.673 8.17S Ni b.562 8.569 $ .52S $ .5$ 4 8.8LT 2.819 8.599 Cu(2) 8.178 8.177 8.883 1.698 8.178 Ni (2) 8.573 S.SSS 8.822 3.663 8.578 P b.b23 $ .823 8.523 $ .818 8.882 11.494 8.822 1.14d 1.168 1.143 1.151 5.517 1.471 1.155 Co 8.518 $ .$ 1$ $ .51$ $ .811 S.SSS 4.876 b.slb d.4dd $ .477 $ .491 $ .469 d.dll 2.231 8.47d V <S.SSS <$ .885 <S.bsS <S.SSS S.SSS S.SSS <S.SSS Cr d.563 $ .886 $ .864 S.d64 8.881 1.494 8.864 Ti <$ .881 <$ .881 <$ .881 <$ .$ 81 <d.ssl C(3) 8.249 8.249 S(4) S.bid S.eld Si (5) $ .153 $ .163 $ .153 5.163 XOTES: (1) ALL MEASUREMEIITS BY ICAP tRLESS OTHERSISE NOTED (2) BY ATOMIC ABSORPTIOX (3) BY LEO COMSSTION (4) BY TITRIMETRIC COSSTIOH (5) BY CRAVIMETRY

0 CHEMICAL ANALYSIS RESULTS (WMAL) FOR NINE MILE POINT UHIT 1. MODIFIED A382B MATERIAL fMA87]

BASE FROM HAZ 368 DECREE CAPSULE (1)

JIL(A) JIL(B) JAM JAE JIT JIP STANDARD RELATIVE MEAN DEVIATION DEVIATION,X Fe MATRIX MATRIX MATRIX MATRIX MATRIX MATRIX Cu 8.172 8.173 8.173 8.174 8.173 8.164 8.684 2.174 8.171 Hi 8.681 8.626 8.654 8.644 8.639 8.598 8.825 4.842 8.626 Cu(2). 8. 165 8.864 2.424 8.165 Ni(2) 8.564 8.819 3.369 8.564 P 6.823 8.626 8.822 8.823 8.821 8.822 8.862 7.543 8.823 Mn 1.142 1.163 1.135 1.152 1.149 1.854 S.B48 3.497 1.133 Co 8.818 8.811 S.811 8.811 8.811 8.818 8.881 4.641 8.811 Mo 8.456 8.463 6.561 8.495 8.466 8.461 8.816 3.213 8.464 V <8.665 <8.665 <8.865 < 6.665 <S.SBS <8.685 <8.665 Cr 8.884 8.665 8.892 8.864 8.863 6.677 8.685 5.694 8.864 Ti <8.881 <8.861 <8.861 <8.881 <8.861 < 8. 681 <8.881 C(3) 8.227 8.227 S(4) 8.818 8.818 Si(S) 8.168 8.168 8.168 8.168 NOTES: (1) ALL MEASUREMENTS BY ICAP UNLESS OTHERWISE NOTED (2) BY ATOMIC ABSORPTION (3) BY LEO COMBUSTION (4) BY TITRIMEfRIC COMBUSTION (5) BY GRAVIMETRY

ICAP CHEMICAL AHALYSIS REPORT (KL) FOR HIHE MILE POINT UHIT 1. IMA87)

UHIRRADIATED ARCHIVE PLATE (1)

D25(A) D25(B) D21 Del STANDARD RELATIVE MEAH DEVIATION DEVIATION,X Fe LNTRIX LNTRIX LNTRIX LNTRIX Cu 8.1ae 8.Ds 8.1ee 8.182 8.885 2.967 8.181 Hi d.ee3 e.sss d.s7s 8.685 8.828 3.384 8.588 Cu(2) 8.184 8.182 8.184 8.182 8.883 1.839 8.183 Ni (2) e.se4 d.s7s e.s78 $ .678 8.822 3.881 8.579 Mn 1.198 1.118 1.148 1.128 8.838 3.122 1.148 Co Mo $ .497 8.496 $ .493 $ .492 8.882 8.449 8.494 V

Cr 8.122 8.117 $ .118 $ .121 $ .882 1.992 e.128 Ti C(3)

S(4)

Si (6)

HOTES: (1) ALL MEASUREMEHTS BY ICAP MESS OTHERNISE NOTED (2) BY ATOMIC ABSORPTION (3) BY LECO COMSSTION (4) BY TITRIMETRIC COLRXSTION (6) BY CRAVIMETRY

EDAX CHEMICAL ANALYSIS RESU TS (BCL) FOR NINE MILE POIHT UNIT 1. MODIFIED 382B MATERIAL [MABSa)

BASE FROM BASE 38 DECREE CAPSll.E EIA(A) E71 EIC(182) EBK(lk2) E2U E31 STANDARD RELATIVE MEAN DEVIATION DEVIATION,8 Fe MATRIX MATRIX MATRIX MATRIX MATRIX MATRIX Cu 8.338 8.268 8.328 8.268 8.288 8.825 8.356 8.298 Hi 8.668 8.588 8.525 8.528 8.618 8.839 6.925 8.563 P .8.858 8.868 8.876 8.838 8.878 8.818 31.383 8.857 Mn Co Mo Y

Cr Ti C

S Si

EDAX CHEMICAL AHALYSIS RESU TS (BCL) FOR HIHE MILE POIHT UHIT 1. MODIFIED 3828 MATERIAL LST84 j BASE FROM BASE 388 DECREE CAPSUL=

E42(A) E42(B) ETE EIM ElU(112) E3T EOT(R) =TANOARO I RELET YE UCAH OEYIATIOH DEVIATIOH, X Fe MATRIX MATRIX MATRIX MATRIX MATRIX MATRIX MATRIX Cu 8.228 8.238 8.228 8.215 8.258 8.238 8.813 5.518 8.228 Hi 8.488 e.458 8.458 8.4vs e.458 8.458 8.813 2.T13 8.463 P 8.822 8.821 8.821 8.819 8.824 8.827 8.883 12.559 8.822 Mn Co Mo Y

Cr Ti C

S Si

1964 LUKENS DATA FROM TEST CERTIFICATE()

ELEMENT G-8-3 G-8-4 G-8-1 G-307-3 G-307-4 G-307-10 Fe MATRIX MATRIX MATRIX MATRIX MATRIX Cu 0.18 0.23 0.20 0.27 0.22 Ni 0.56 0.51 0.48 0.53 0.51 P 0.012 0.021 0. 018 0.019 0.018 Mn 1.16 1.34 1.45 1.25 1.43 Mo 0.47 0.45 0.45 0.52 0.51 C 0.20 0.19 0.18 0.20 0.20 S . 0.027 0.028 0.034 0.030 0.026 Si 0.17 0.21 0.26 0.21 0.26 (1) Based on discussions with Lukens, data from ladel analysis by atomic absorption.

F.2 WELD DATA The weld chemistry data were taken from references [LE64],

[CE90], and [ST84].

REACTOR VESSEL BELTLINE WELD INFORMATION Weld Seam Weld Wire Type Weld Flux Type Detailed Weld Number Location and Heat No. and Lot No. Procedure 2-564 A/C Lower-Intermediate RACO 3/86054 Arcos B-5/4E5F SAA-33-A (3 )

Shell Longitudinal RACO 3/1248 Arcos B-5/4K13F SAA-33-A (3)

E8018/HACD N/A MA-33-A(7)

E8018/JBGD N/A MA-33-A(7) 2-564 D/F Lower Shell RACO 3/86054 Arcos B-5/4E5F SAA-33-A ( 3 )

Longitudinal Seams E8018/HACD N/A MA-33-A(7)

E8018/JBGD N/A MA-33-A(7) 3-564 Lower Intermediate to RACO 3/1248 Arcos B-5/4M2F SAA-33-A (3)

Lower Shell Girth E8018/DBDE N/A MA-33-A(7)

E8018/IOGE N/A MA,-33-A(7)

Surveillance All Three Capsules RACO 3/W5214 Arcos B-5/5G13F SAA-33-A (3)

Capsule Weld Reference [CE90]

BELTLINE WELD CHEMISTRY DATA Weld Wire Weld Seam Heat Flux Lot Si Mo Ni Cu 2-564 A/C 86054/4E5F .12 1. 64 .015 .020 .34 .51 1248/4K13F .11 1.71 .005 .017 .38 .56 2-564 D/F 86054/4E5F .12 1.64 .015 .020 .34 .51 3-564 1248/4M2F .10 1.26 .015 .020 .22 .57 Surveillance Capsule (2) (2)

Weld 5214/5G13F .14 1.58 .018 .013 .25 .51 (. 18) (. 09)

( ~ 023)

(1) Data in parenthesis were measured using irradiated material and reported in [ST84].

(2) Reference [CE90] recommends the use of significantly higher values based on examination of generic .data Reference [CE90]

'T It F.3 SUPPLEMENTARY BASE METAL CHEMICAL ANALYSIS The base portion of the HAZ tensile specimen, JUD, from the

'300 degree capsule was analyzed. Also, the base portion of the HAZ Charpy specimen, JlM, from the 300 degree capsule was also analyzed. The objective was to show that the chemistries of the base from HAZ are similar and match the G-8-3 Lukens data. The attached data support this theory.

F-14

BASE METAL CHEMISTRY ANALYSIS FOR IRRADIATED SPECIMENS Concentration in Wei ht Percent JlM JUD (Base from (Base from Element Fe BALANCE BALANCE Co 0.011 0.013 Cr 0.075 0.092 CQ 0.183 0.168 Mn 1.044 1.129 Mo 0.472 0.474 Ni 0.642 0.579

<0.010 0.026

<0.010 <0.010

APPENDIX G TENSILE DATA

The unirradiated weld properties are given in references

[LE64] and [CE90]. The unirradiated plate properties are given in [ST64]. Irradiated properties are reported in Battelle reports [ST84], [MA85a], and [NA87]. The G-8-3 archive plate data and irradiated HAZ base metal miniature tensile data were measured as part of the current work.

G-2

G.l BASELINE TENSILE DATA G-3

TABLE G-1 UNIRRADIATED WELD TENSILE DATA Yield Strength Ultimate Tensile Elongation Reduction

~T e Heat No. Flux Lot No. Psi Stren th Psi In 2" % Of Area  %

RAC03 86054 4E5F 75, 500 90, 000 27.5 69.9 RAC03 1248 4K13F 66, 800 84, 100 26. 0 64. 9 RAC03 1248 4M2F 63, 000 80, 000 27.5 64.3 Surveillance Capsule Weld W5214 5G13F 65, 000 84, 000 27.5 67.0 (1) Data taken from [LE64] and [CE90]. The test records do not indicate the test temperature.

Therefore, RT is assumed.

TABLE G-2 UNIRRADIATED BASE METAL DATA Plate Yield Strength Ultimate Tensile Elongation Reduction Code Heat No. Psi Stren th Psi In 2" % Of Area  %

G-307-3 P2074-2 62,000 82, 000 28.0 69. 0 G-307-10 P2091-2 69,400 92, 900 25.0 67.0 A G-307-4 P2076-1 69,400 89,900 27.0 66.2 I

G-8-1 P2112-1 66, 600 87,500 27.0 66.0 G-8-3 P2130-1 65, 000 86, 200 26.0 65.4 G-8-4 P2130-2 59, 300 85,500 29.0 68.0 (1) Data taken from [ST64] The test records do not indicate the test temperature.

Therefore, RT is assumed.

G.2 IRRADIATED SURVEILLANCE SPECIMEN DATA

TABLE G-3 TENSILE PROPERTIES FOR THE IRRADIATED MATERIALS FROM THE NINE MILE POINT 300-DEGREE SURVEILLANCE CAPSULE~~

Test Fracture Reduction Specimen Material"'emp.'> Stren th si Stress in Area Elon ation ercent'~'otal No. Type (F) Yield Ultimate Fracture (psi) (percent) Uniform JJA Base RT 79, 170 99, 700 66, 060 192, 300 65.7 12.5 27.7 JDB Base 550 69, 410 92, 890 68, 090 161, 800 58.0 8.9 19.7 JLB Weld RT 73, 680 90, 240 59, 450 186, 300 68.1 13.0 23.2 JL7 Weld 550 67, 760 84, 690 59,180 157, 600 62 ' 10.5 20.9 JUD RT 63, 720 85, 060 54,880 181,200 69.7 7.5 19 ~ 8 JTU 550 59, 960 81,500 56, 910 145, 100 60.8 7.1 18.4 Data taken from [ST84]. The weld and HAZ specimens were most likely fabricated using plate G-8-3 material. The composition of the base tensile specimens is not known at present.

(2) Room temperature (RT) is approximately 75'.

(3) The elongation is for a 1-inch gauge length.

(4) The composition of the surveillance tensile specimens is not known.

Chemical analyses are recommended in the future.

TABLE G-4 TENSILE PROPERTIES FOR THE IRRADIATED BASE METAL FROM THE NINE MILE POINT 30-DEGREE SURVEILLANCE CAPSULE Test Fracture Reduction Specimen Material'" Temp. Stren th si Stress in Area Elon ation ercent Type (F)

+'o.

Yield Ultimate Fracture (psi) (percent) Uniform

'~'otal JDE Base RT 76, 078 96, 817 65, 708 193, 939 66.1 12.0 24.1 (1) Data taken from [MA87]. The composition of the base tensile specimens is not known at present.

(2) Room temperature (RT) is approximately 75'.

(3) The elongation is for a 1-inch gauge length.

(4) The composition of the surveillance tensile specimens is not known.

Chemical analyses are recommended in the future.

G.3 ARCHIVE PLATE G-8-3 DATA The axial extensometer slipped off specimen TN-2 during testing. As- a result, the stress/strain curve could not be accurately plotted and the uniform strain could not be determined.

G-9

TABLE G-5. TENSILE PROPERTIES FOR THE ARCHIVE PLATE G-8-3 Engineer>ng Strength (kss) True Red.

Hatl Test Fracture Yield Ultimate Fracture Fracture Bridgman In Elongatton Spec. Type Temp. Load .2X or Stress Correction Area Uni form Total (F) (lbs) upper loMer (ksi) (ksi) (X) (X) (X) tn-1 Base 78 13281. 67.4 65.4 87.2 66.8 178.3 152.5 62.5 13.8 45.2 tn-3 Base 250 11719. 64.7 61.9 80.4 58.5 156.5 133.8 62.6 10.1 33.3 tn-2(1) Base 550 12842. 57.9 -.-- 86.3 64.1 148.3 129.7 56.8 38 5 The elongation is for a(n) 1.0 inch gage length (1) The uniform elongation could not be determined because extensometer slipped during test

UNIRRADIATED ARCHIVE PLATE Q-8-3 SPECIMEN: TN- 't TESTED AT 78 F

't 00 IMQIRIIHIEKI%9MQ S7RKSS

+ IMANIIIMtLIM 50 87IRK88 FINA,CVUIRK 87MSS 0.00 0.1 0 0.20 0.30 0.40 0.50 STRAIN (IN/IN)

UNIRRAOIATED ARCHIVE PLATE G-8-3 SPECIMEN: TN-1 TESTED AT 78 F 200 r r<

r IKMQIR0HIKIKIR0HQ rr 87IRIK88 r rrr 7IRUIK r r 87IRIK88 100 SIRIISCIMAH COIRIRIKC7ION

/

65lAXOMUIM 87IRIK88 iFIRA,C7INIRIK 87IRIK88 0.00 0.20 0.40 0.60 0.80 'I.OO 1.20 STRAIN (IN/IN)

G-12

1'Y" C

k

~h h4 7 ."<~

ggcj) p4 28 iQ it~i A

~v]

FIGURE G-1 POST-TEST PHOTOGRAPHS OF THE UNIRRADIATED BASE METAL ARCHIVE PLATE G-8-3 TENSILE SPECIMEN TN-1 SHOWING BOTH THE REDUCED AREA AND FRACTURE SURFACE G-13

I}'

g4 4

1

I F GURE G-2 POST-TEST PHOTOGRAPHS OF THE UNZRRADZATED BASE METAL ARCHIVE PLATE G-8-3 TENSILE SPECIMEN TN-2 SHOWING BOTH THE REDUCED AREA AND FRACTURE SURFACE

UNIRRADIATED ARCHIVE PLATE Q-8-3 SPECIMEN: TN-3 TESTED AT 250 F IBQHIIIHIEIRIRIIM 87IRKSS

+ MAMIMUM 87IRKSS tF(M CVURIK 87(RIESS 0.00 0.1 0 0.20 0.30 0.40 0.50 STRAIN (IN/IN}

G-15

UNIRRADIATEO ARCHIVE PLATE G-8-3 SPECIMEN: TN-3 TESTED AT 250 F 200 r

rr r rr

'f 00 r Qg3IIDDQIgg(g CIRIMCYIIM I/ IMANOMIUIM SYIMSS ISA,C'rUIRIB SYM88 0.0 0 0.20 0.40 0.6 0 O.S 0 t.0 0 1.20 STRAIN (IN/IN)

k ~

,k

'd bl.

1~ v tl >i~

FIGURE G-3 POST-TEST PHOTOGRAPHS OF THE UNIRRADIATED BASE METAL ARCHIVE PLATE G-8-3 TENSILE SPECIMEN TN-3 SHOWING BOTH THE REDUCED AREA AND FRACTURE SURFACE

6.4 MINIATURE BASE METAL SPECIMEN DATA (BASE METAL TAKEN FROM 300 DEGREE CAPSULE WELD CHARPY SPECIMENS)

Based on analysis of base metal taken from likely that HAZ specimens, these specimens were prepared from plate G-8-3 it is G-18

TABLE G-6. TENSILE PROPERTIES FOR IRRADIATED BASE METAL TAKEN FROM WELD HAZ SPECIHENS (Based on HAZ Base Metal Analysis, the HateTial is Host Likely G-8-3)

Engineering Strength (ksi) True Red.

Hatl Test Fracture Yield Ultimate Fracture Fracture Bridgman in Elongation Spec. Type Temp. Load .2X or Stress Correction Area Uniform Total (F) ( lbs) upper lover (ksi) (ksi) (X) (X) (X) ejd Base 78 1079. 71 1 93.3 61.9 177.4 150.4 65.1 10.5 31.5 ejt Base 250 1001. 66.8 66.3 86.6 58.2 157.3 134.3 63.0 8.7 29.4 edt Base 550 1138. 62 8 90.0 66.1 150.8 132.2 56.2 26.6 31.5 The elongation is for a(n) .102 inch gage length

BASE METAL FROM WELD CVN SPECIMEN: E JD TESTED AT 7 8 F

'f 00 IRMQRIIIMIKIRRIMQ SYIRIKSS

+ MANIISlUM 50 SY(RlKM PIMC'rUM 8YlMSS 0.00 0.1 0 0.20 0.30 0.40 0.50 STRAIN (IN/IN)

G-20

BASE METAL FROM WELD CVN SPECIMEN: E JD TESTED AT 7 S F 200 rr rr rr r KIMCIROHKKIROHQ SVIRKS8 rr r VIRUK rrr SVIRKSS 100 p' SIRODOLIAH CIRIRKCVIOII OIANOMUM SVIRKSS IFIMCVUIRK SVIRKSS 0.00 0.20 0.40 0.80 0.80 'I.00 1.20 STRAIN (IN/IN)

0 J ':,24 28 p pvp 24 28 FIGURE G-4 POST-TEST PHOTOGRAPHS OF IRRADIATED BASE METAL TENSILE SPECIMEN EJD SHOWING BOTH THE REDUCED AREA AND FRACTURE SURFACE G-22

BASE METAL FROM WELD CVN SPECIMEN: E JT TESTED AT 260 F 100 IEIMIIIIHIMRIM 87MSS

+ MAMMUH 87MSS lF IRACVUIM 87IRKSS 0.00 0. 1 0 0.20 0.30 0.40 0.50 STRAIN (IN/IN)

G-23

BASE NIETAL FROM WELD CVN

- SPECIMEN: E JT TESTED AT 250 F 200 KHQIRIIHKKIR0HQ rr 87IRK88 7IRVK rrr 87IRK88 rrr QIRIIpDQIUQ,Ig 100 p CIRIRKCVIOIM MANIIIMUIM 87IRK88 IFIRACVUIRIF.

87IRK88 0.00 0.20 0.40 0.60 0.80 '!.00 'I.20 STRAIN (IN/IN)

G-24

p' lp l'

~rf rry!'!

y<~ 'r.

r

• ~

.;*!i+r r ',

4 8 1r. 24 28 Q $ D lA '0 J AW AC)

~rr ~i' - .* ~

g~;gpss;.~lpl

,jigjj ZB FIGURE G-6 POST-TEST PHOTOGRAPHS OF IRRADIATED BASE METAL TENSILE SPECIMEN EJT SHOWING BOTH THE REDUCED AREA AND FRACTURE SURFACE G-25

BASE METAL EROM WELD CVN SPECIMEN EDT TESTED AT 550 F sac IKMQIRIIMEIMIIIMQ 87MSS

+ lMAMSIUM 8YÃtK88 lFiMC'rUIRlK SYIMSS 0.0 0 0.20 0.40 0.80 0.80 1.00 STRAIN (IN/IN)

G-26

BASE METAL FROM WELD CVN SPECIMEN: EDT TESTED AT 550 F 200 KMRRIIHIKKIR!IMQ 87IRIK88 7IRUIK 87IM88 rr SIRISCHAH 100 / COIRMC7IIOM I

MANRilUM 87IRKSS IFIMCVUilK 87IR 888 0.00 0.20 0.40 0.80 0.80 'f.00 1.20 STRAIN (IN/IN)

G-27

qa~,

I

24 28 32 56

>4 3> ~0 g8

-yS F~4 FIGURE G-5 POST-TEST PHOTOGRAPHS OF IRRADIATED BASE METAL TENSILE SPECIMEN EDT SHOWING BOTH THE REDUCED AREA AND FRACTURE SURFACE

APPENDIX H CHARPY DATA

~-

The unirradiated weld properties are given in reference

[LE64]. The G-8-3 archive plate data for the T-L orientation were measured as part of the current work and the L-T orientation data was originally reported in Reference [MA87] . The unirradiated plate properties are given in [ST64]. Irradiated properties are reported in Battelle reports [ST84], [MA85a], and

[MA87] .

H-2

H.1 UNIRRADIATED DATA H-3

H.1.1 1964 UNIRRADIATED BASELINE CHARPY DATA

TABLE H-1. UNIRRADIATED WELD CHARPY DATA

~Te Heat No. Flux Lot No. +10'F Zm act Values ~Auexa e RAC03 86054b 4E5F 66.0, 64 ', 65.0 65. 2 RAC03 1248 4K13F 55.0, 51.0, 57 ' 54.4 RAC03 1248 4M2F 53 ', 57.0, 65.0 58 '

Surveillance Capsule W5214 5G13F 61, 52, 58 57.0 Data taken from [LE64] and [CE90]

TABLE H-2. UNIRRADIATED PLATE G-8-3 CHARPY DATA~~

Test Tem erature F Im act Ener ft-lb Avera e ft-1b

-80 9.0, 6.0 7.5

-40 32.0, 17.0 24.5

+10 50.0, 37.0, 47.5 44.8 60 77.0, 63 ' 70.0 110 90 ', 99.0 94.5 160 100.0, 96.0 98.0 212 87 ', 78.5 82.7 Data taken from reference [ST64]

~" 4l TABLE H-3. UNIRRADIATED PLATE G-8-4 CHARPY DATA Test Tem erature 'F Im act Ener ft-lb Aveia e ft-1b

-40 23.5, 26.0 24.7

+10 46.0, 61.0, 42 ' 49. 6 40 58.0, 64.0 61.0 60 72.5, 74 ' 73.5 110 98 ', 98.0 98.0 160 100.5, 100 ' 100.2 212 106.0, 106.0 106.0 Data taken from reference [ST64]

H-7

NINE MILE POINT UNIT I ~ EXPERIMENTAL UNIRRADIATED BASE METAL 083/Q84 (1 2)

DATA 120 WEIBVLL FIT TRANSITION 100 g+Jf j WEIBVLL FIT I 0 UPPER SHELF IL 80 HYPERBOLIC TANGENT FIT 60 CONFIDENCE J LIMIT (95%)

J 40 0 CONFIDENCE V 0 (~

/ LIMIT (95%)

4 X

20 CONFIDENCE LIMIT (95%)

CONFIDENCE

-100 -50 0 50 'I 00 150 200 250 LIMIT (95%)

TEST TEMPERATURE (F)

H-8

l C V'+Jul

TABLE H-4. UNIRRADIATED PLATE G-8-1 CHARPY DATA

• "*'Test Tem erature 'F Im act Ener ft-1b Avera e ft-1b

-40 13.0 13.0

+10 33.0, 33.5, 25.0 30.5 40 43.5, 45.0 44 '

60 47.0, 58.0, 55.0 53.3 110 80.0, 70.0, 70.0 73 3.

~

212 82.0, 95.0, 83.0 86.7 Data. taken from ref erence [ST64]

H-9

NINE MILE POINT UNIT UNIRRADIATED BASE METAL Q-8-1 (1,2)

DATA 120 WElBULL FIT TRANSITION 100 I

l HYPERBOLIC tL TANGENT FIT 80 CONFIDENCE 60 LIMIT (9SX)

CONFIDENCE 40 LIMIT (95%)

V 20 k CONFIDENCE k LIMIT (95%)

A CONFIDENCE

-100 -50 0 50 100 150 200 250 LIMIT (95m)

TEST TEMPERAl VRE {F)

H-10

TABLE H-5. UNIRRADIATED PLATE G-307-3 CHARPY'DATA~>>

Test Tem erature 'F Im act Ener ft-lb Avera e ft-lb

-40 8.0 8.0

+10 27.5, 37.5, 41.0 35.3 40 41.5, 52 ', 57.0 50.2 60 63. 5 63. 5

+75 58.0 58.0 90 71.5 71.5 110 82.0 82.0 140 90.0 '90.0 160 101.5 101.5 212 100.0, 106.5 103.3 Data taken from [ST64]

• ~ 4$

'I>

NINE MILE POINT UNIT I ~

UNIRRADIATED BASE METAL Q-307-3 (1 I) EXPERIMENTAL DATA 120 WEIBULL FiT Ql TRANSITION 100 t

I I- HYPERBOLIC LI 80 TANGENT FIT

/ 4 60 t a CONFIDENCE LIMIT (95%)

LLI i

/a

/~ CONFIDENCE 40 k LIMIT (95%)

V /

//e CONFIDENCE 20 LIMIT (95%)

0 CONFIDENCE

-100 -50 0 50 100 150 200 250 LIMIT (95%)

TEST TEMPERATURE (F)

H-12

TABLE H-6. UNIRRADIATED PLATE G-307-4 CHARPY DATA~ ~

Test Tem erature 'F Im act Ener ft-1b Avera e ft-lb

-40 13.0, 20.0, 12 ' 15.0

+10 37.0, 35.0, 38.0 36. 6 60 50.0, 41.0, 61.0 50. 6 110 82.0, 75.0, 84.5 80.5 212 80.0, 83.5, 80 ' 81.0 Data taken from [ST64)

NINE MILE POINT UNIT 1

~

VNIRRADIATEP BASE METAL G 307 4 (1 1) EXPERIMENTAL DATA 120 WEIBULL FIT TRANSlTION IXI 100 l WEIBULL FIT I- kkkk kkkk UPPER SHELF LL 80 P kkkk k H YPERBOLlC TANGENT FIT 60 CONFlD ENCE k/ LIMIT (95%)

LLI k/

k /

40 k CONFIDENCE V k k

k k

LlMIT (95%)

k CL 20 CONFlD ENCE uMIT (95%)

-'I 00 -50 CONFIDENCE 0 50 'j 00 150 2'00 250 uMIT (95%)

TEST TEMPERATURE (F)

H-14

'II%

TABLE H-7. UNIRRADIATED PLATE G-307-10 CHARPY DATA~i Test Tem erature 'F Im act Ener ft-lb Avera e ft-1b

-40 16. 5, 12. 5, 13. 0 13.3

+10 40.0, 33 ', 45.0 39.3 60 45.0, 56.5, 62.0 54.3 110 68.0, 80.0, 63.0, 70.3 212 97.0, 100 ', 100.0 99.0 Data taken from [ST64]

H-15

r V4 NINE MILE POINT UNIT 1 UNIRRADIATED BASE METAL Q 307 1 0 (1 2)e~ EXPERIMENTAL DATA 120 WEIBULL FiY TRANSITION 100 I

I- HYPERBOLIC LL TANGENT FIT 80 r 0 CONFIDENCE 60 / g 4 LIMIT (e5%)

J k

ILI o >r e CONFIDENCE 40 LIMIT (95%)

V 20 CONFIDENCE uMIT (e5v)

CONFIDENCE

-100 -50 0 50 'I 00 150 200 250 uMIT (e5v)

TEST TEMPERATURE (F j H-16

H.1.2. UNIRRADIATED ARCHIVE PLATE G-8-3 DATA H-17

C,r LE H-8. CHARPY V-NOTCH IMPACT RESULTS OR UNIRRADIATED BASE METAL SPECIMENS PREPARED FROM PLATE G-8-3 (L-T orientation)

Location Test Impact "'pecimen Lateral Fracture Within Temperature Energy. Expansion Appearance Identification Plate (F) (ft-lb) (mils) (% Shear)

C28 1/4T -80 8.4 12.4 5 C06 3/4T -80 9.2 10.2 5 C27 1/4T -40 14.5 18.2 10 C04 -40 24.0 29.6 10 C24 1/4T 10 85 ' 50.8 50 C03 3/4T 10 55.0 44.4 25 C22 1/4T 35 51 ' 43.8 40 C23 1/4T 35 107.0 77.4 100 C02 3/4T 35 67 ' 53.4 50 C21 1/4T 60 104.5 68 ' 65 C01 3/4T 60 97.5 82.0 70 C29 1/4T 110 108.0 73 ' 100 C07 3/4T 110 105.5 79.0 100 C31 1/4T 160 108.0 78.0 100 C09 3/4T 160 113.0 83.8 100 C30 1/4T 212 109.5 82.4 100 C08 3/4T 212 110 ' 81.4 100 Data taken from reference [MA87]

NINE MILE POINT DATA AND CURVES 8 Data Yatwe Wottwll fit TAlat t'tt OS Conttdonco t.tmtt ~

0 O

O

-l00 -50 0 50 l00 l50 200 250 TEST TEMPERATURE (F)

CVN IMPACT ENERGY VERSUS TEST TEMPERATURE FOR PLATE G-8-3 (L-T ORIENTATION)

H-19.

COMPAR l SON PLOT DATA AND CURVES O

0 plate ~-5, $ 67 teet 0 plod ~-b, $ 64 teet b pled ~-4> $ 54 teet 0

0 LXI I o EO 0

4 0

h cr 9 h LJJ 0

0 100 -50 0 50 100 150 200 250 TEST TEMPERATURE (F)

COMPARISON OF 1987 ARCHIVE PLATE G-8-3 DATA AND 1964 CVN IMPACT DATA ON PLATES G-8-3 and G-8-4 H-20

H.1.3. 1990 Current Stud UNIRRADIATED ARCHIVE PLATE G-8-3 DATA T-L orientation H-21

LE H-9. CHARPY V-NOTCH IMPACT RESULTS OR'UNIRRADIATED BASE METAL SPECIMENS PREPARED FROM PLATE G-8-3 (T-L orientation)

Location Test Impact Lateral Fracture Specimen Within Temperature Energy Expansion Appearance Identification Plate (F) (ft-lb) (mils) (% Shear) 3N2 3/4T -80 7.3 5 4 8.1 1N2 1/4T -80 9.0 6.4 6.5 3N1 3/4T -40 11.0 9.4 20.5 1N1 1/4T -40 12.2 11.2 15.3 3N4 3/4T 10 33.9 28.6 34.4 1N4 1/4T 10 41.0 34.4 37.6 3N5 3/4T 35 41.0 37.4 56. 0 1N5 1/4T 35 45.5 39.4 61. 1 3N9 3/4T 57 60.1 52.0 83.5 3/4T 53.1 48.2 81.0

'N8 57 1N8 1/4T 57 57.2 51.4 74.9 3N7 3/4T 79 53.0 48.6 95.6 1N7 1/4T 79 51.2 46.6 73.3 1N9 1/4T 79 68.3 62.0 81.0 3N6 3/4T 110 69.0 59 ' 100.0 1N6 1/4T 110 73 ' 61 ' 100 '

3N3 3/4T 212 68.0 61.2 100.0 1N3 1/4T 212 64.0 58.8 100.0

NINE MILE POINT UNIT I ~ EXPERIMENTAL UNIRRADIATED g4$ E METAL Q Q 3 (ff )( I 2)

DATA WEIBULL FIT CQ TRANSITION 100 I

f HYPERBOLIC U

80 TANGENT FIT CONFIDENCE 60 LIMIT (95%)

CONFIDENCE 40 LIMIT (95%)

0 20 CONFIDENCE LIMIT (95%)

0 CONFIDENCE

-100 -50 0 50 100 150 200 250 LIMIT (95%)

TEST TEMP ERAT V RE (F)

H-23

~g

<<4.'at

20 24 28 TESTED AT -800F l; ) 24 28 TESTED AT -80 F H-24

~ lg V ,a

TESTED AT -40"F I I l 1 io ~~ 24 28 TESTED AT -40 F H-25

tgg ll

24 28 TESTED AT 10 F g g I I I

~ 24 28 TESTED AT 10 F H-26

\

P

)

t

'2 'I ~'lK

I24 28 TESTED AT 35 F 8 u zO 24 2S TESTED AT 35 F H-27

Ci c

TESTED AT 57'F j f a TESTED AT 57 F

s

12 TESTED AT 57 F I a 8 12 z4 28 TESTED AT 79"F H-29

($ qf

ll ~

v>

I 8 24 28 TESTED AT 79 F ii

(>>

.g'))I

ggj~jpq4V TESTED AT 79 F H-30

p TESTED AT 110 F 4 8 > . 3 24 28 TESTED AT 110 F

I I

~ .

>'4 i

TESTED AT 2120'ESTED AT 212 F H-32

H.2. IRRADIATED SURVEILLANCE SPECIMEN DATA H-33

H.2.1. 1984 IRRADIATED 300 DEGREE CHARPY DATA 0

TABLE H-10. CHARPY V-NOTCH IMPACT RESULTS FOR IRRADIATED BASE METAL SPECIMENS FROM THE NINE MILE POINT 300-DEGREE SURVEILLANCE CAPSULE Impact Lateral Fracture Specimen Test Energy, Expansion Appearance Identification Temperature, F ft-lb mils  % Shear E1U-RC -40 6.3 2.8 1 E42 0 '10 9.2 9 ElJ 40 17.5 17.4 10 E1M 75 28 28.2 21 E3T 120 30 30 30 EA5 135 53 45.2 35 E1V 160 52.8 50 E1M-RC 200 52.5 47.8 40 E7E-RC 200 59 38 65 E7E 240 78 61.4 100 E3T-RC 280 99 72.2 100 ElK 320 104 78.4 100 RC = Reconstituted (1) Data taken from Reference [ST84]

(2) Based on chemistry data, this material is believed to be plate G-8-1 material.

NINE MILE POINT UNIT 1 IBB4pI4TEp B4QE METAL 3 pp pEG g4p Iq q t EXPERIMENT4L DATA 120 WEIBULL FIT TRANSITION 100 I WEIBULL FIT I- UPPER SHELF L

80 HYPERBOLIC Q TANGENT FIT 60 CONFIDENCE LIMIT (95%)

40 /4 CONFIDENCE O

LIMIT (95%)

20 X CONFIDENCE LIMIT (95%)

0 50 100 150 200 250 300 350 LIMIT (95%)

TEST TEMPERATURE (P)

H-36

TABLE H-11. CHARPY V-NOTCH IMPACT RESULTS FOR IRRADIATED WELD METAL SPECIMENS FROM THE NINE MILE POINT 300-DEGREE SURVEILLANCE CAPSULE Impact Lateral Fracture Specimen Test Energy, Expansion Appearance Identification Temperature, F ft-lb mils  % Shear E JD-RC -120 7.2 3 EDL -100 3.4 7 EDT-RC -80 33 30.4 17 EJD -40 28 25.2 22 EDJ -20 33 30.4 27 EJC 0 59 52 42 EDK 40 61 54.8 64 EDT 75 76.5 63.2 68 EDJ-RC 125 110.5 84.6 100 EDM 160 100 66 100 EJC-RC 240 116 78 100 EJT 280 112 85 100 RC = Reconstituted (1) Data taken from Reference [ST84]

>a l 1 Ph Il f>>,

NINE MILE POINT UNIT 1 RRADIATED ~ELp 3pp DEGREE CAPSULE(q 2) EXPERIMENTAL DATA 150 WEI BULL FIT kk TRANSITION 125 kk kk I kkkk Qkkkkk WEIBULL FIT I k

• UPPER SHELF LL. k 100 kk HYPERBOLIC k

j TANGENT FIT 75 k j/ ** CONFIDENCE k

k LIMIT (95%)

k 50 k CONFIDENCE V k

• 4 Ok k

k Ok LIMIT (95%)

25 k k k k k CONFIDENCE k

k k LIMIT (95%) ~

k

-150 'I 50 CONFIDENCE LIMIT (95%)

TEST TEMPERATURE (F)

H-38

I I

TABLE H-12. CHARPY V-NOTCH IMPACT RESULTS FOR IRRADIATED HAZ METAL SPECIMENS FROM THE NINE MILE POINT 300-DEGREE SURVEILLANCE CAPSULE Impact Lateral Fracture Specimen Test Energy, Expansion Appearance Identification Temperature, F ft-lb mils  %, Shear J2C -40 17.5 12.6 20 J1L 0 44 32.4 35 40 33 33.4 42 J1T 50 57.5 48 70 J1M 77 64 2

~ 9'0 160 82 59. 6 100 JlP 200 96 76 100 Jl J 280 96.5 79.2 100 (1) Data taken from Reference [ST84]

100 90 70 50 40 30 20 LO

-COO WO 0 50 LOO 150 200 250 300 350 Test Teaperstvre F CHARPY V-NOTCH IMPACT ENERGY VERSUS TEST TEMPERATURE FOR THE IRRADIATED HAZ METAL SPECIMENS FROM THE NINE MILE POINT 300-DEGREE SURVEILLANCE CAPSULE H-40

H.2.2. 1985 IRRADIATED 30 DEGREE CHARPY DATA

TABLE H-13.

SUMMARY

OF CHARPY IMPACT DATA FOR IRRADIATED BASE MATERIALS FROM THE 30-DEGREE CAPSULE~i'~i Lateral Fracture Specimen Type Test Energy, Expansion Appearance Identification Specimen Temp, F ft-lb Total Mil  % Shear E1A Base 85 47.0 43 ' 40 E2U Base 125 68.0 55 ' 50 I

c Elc Base 45 27.7 25.4 30 E31 Base 100 50.5 42.4 40 EBK Base 60 24.0 25.4 25 E71 Base 72 25.0 25 ' 30 (1) Date taken from reference [MA85a]

(2) Based on chemistry data, this material is believed to be plate G-8-1 material.

NINE MILE POINT UNIT 1 IRRADIATED BASE 30 DEG CAP ( I, 1 )

120 CG 100 I EXPERIMENTAL I- DATA L,

80 WEIBULL FIT TRANSITION Q HYPERBOLIC 60 TANGENT FIT CONFIDENCE 40 LIMIT (95%)

O J A CONFIDENCE 20 O LIMIT (9 5%) .

20 X

-100 -60 0 60 100 150 200 250 TEST TEMPERATURE (F)

H-43

H.2.3. 1990 current stud IRRADIATED BASE FROM HAZ FOR THE 300 DEGREE CAPSULE H-44

TABLE H-14. CHARPY V-NOTCH IMPACT RESULTS FOR IRRADIATED BASE METAL SPECIMENS PREPARED FROM HAZ SPECIMENS TAKEN FROM THE 300 DEGREE CAPSULE Impact Lateral Fracture Specimen Test Energy Expansion Appearance Identification Tem erature F ~ft-lb ~mile JlP -50 23.0 17.0 11.5

-50 8.2 3.0 11.0 J1J 42.0 32.6 29.1 J1T 25.7 20.6 27.0 35 57 ' 44.4 39.4 J1M 150 114.3 85.8 100.0 (1) Based on chemistry data, this material is believed to be Plate G-8-3 material.

0 NINE MILE POINT UNIT RRAPIATEP BASE(HAZ) 300 DEG CAP (1 2) EXPERIMENTAL DATA 120 WEIBULL FIT Cl TRANSITION 100 I

I- HYPERBOLIC 0

80 TANGENT FIT Q CONFIDENCE 60 LIMIT (9 5%)

CONFIDENCE ao J LIMIT (95%)

O /fA 20 CONFIDENCE X LIMIT (9 5%)

CONFIDENCE

-100 -50 0 50 100 150 200 250 LIMIT (95%)

TEST TEMPERATURE (P}

V'c 1 1

4 8 ti.vo ZO 24 28 32 TESTED AT -50"F

,>Qypgjjgf,lf t 4 8 iz 16 20 24 28 32 TESTED AT -50 F H-47

iL pj

l pi 32.

12 'lo-20 24 28 64 4 TESTED AT -5 F

'~f8; 4 8 12 . ~- -~v z4 28 32 TESTED AT -5 F H-48

A+

Aw" a'l'S

Q 12 16 2~

TESTED AT 35 F pro y

f34'~v iE 4 8 12 32 TESTED AT 150 F H-49

APPENDIX I HARDNESS DATA

0 TABLE I-1. ROCKWELL C HARDNESS DATA Specimen Indentation"'urface

"'verage Fluence Rockwell' Hardness Rockwell"'ardness Identification Material (n/cm~) B C 3N3 side G-8-3 0 3N3 notch G-8-3 0 89.8 1N8 side G-8-3 0 10.5 1N8 notch G-8-3 0 10.4 JAE side G-8-3 4.78x10'~ 92. 5 14.1 JAE notch G-8-3 4.78x10'~ 90 1 F 12.5 JIT side G-8-3 4.78x10" 92.5 14.0 JIT notch G-8-3 92.3 13.2 4.78x10~'.78x10'~

EIM notch G-8-1 4.78x10~'0.0 90.7 15.2 E42 (4) G-8-1 4.78x10'~ 93.8 14.3 RElU side G-8-1 4.78x10'~ 94.2 15.4 RElU notch G-8-1 91.2 15.2 "side" indicates the indentation was performed on the surfaces normal to the notch.

"notch" indicates the indentation was performed on the notched surface in the direction of crack propagation.

l (2) base specimens from the 300'apsules are believed to be from plate G-8-1 and base specimens from HAZ or Weld are believed to be G-8-3 material.

(3) the estimated uncertainty is + 1.0 for the calibration.

(4) since the fracture surface was removed prior to hardness testing, the notch orientation could not be determined.

APPENDIX J DROP WEIGHT DATA Drop weight tests were con'ducted on archive plate G-8-3 using ASTM type P-2 specimens. The NDT for plate G-8-3 is -25'F.

TABLE J-1. DROP WEIGH ST DATA FOR PLATE G-8-3 Specimen Test Identification Location Within Plate Temperature (F) Break D21 1/4T -70 Yes D22 1/4T -50 Yes D02 3/4T -40 Yes D04 3/4T -35 Yes D24 1/4T -35 Yes D01 3/4T -30 No D25 1/4T -30 No Test D26 1/4T -30 No D03 3/4T -30 Yes D23 1/4T -30 Yes D05 3/4T -25 Yes D06 3/4T -25 No D27 1/4T -20 No D28 1/4T -20 No D07 3/4T -20 No D08 3/4T -20 No

• Specimen run at 320 ft-lb instead of 300 ft-lb.

APPENDIX K FLUX AND FLUENCE DATA

K.1 30 DEGREE CAPSULE DATA K-2

<ap K. 1. 1 Dosimetry Table K-1 contains the results of the analysis of the dosimetry wires from the 30-degree position surveillance capsule.

The wire identifications, measured radionuclide activities (in dps-mg '), and wire compositions are tabulated. The measured activities have been decay corrected to November 30, 1984. The composition of the Cu and Ni wires was taken from Reference

[HI69]. The percent Fe for the iron wire was taken from Reference [LO84].

As indicated in Table K-l, the two Ni wires, Pl-Ni.l and P2-Ni.2, are listed for packet P1. This is because packet Pl contained two Ni wires but no Cu wire. Both Ni wires were anlayzed witht he results listed. For the Co-58 measurement on he Ni wires, no Co-58 activity was measured due to the long decay time following removal of the capsule from the reactor and the relatively short half-life of the isotope (70.8 days). The values which are listed in the table for Co-58 are upper limits for the-specific activity based on detection limits.

K.1.2 Fluence Calculation To accurately determine the neutron flux at the capsule and in the pressure vessel wall, the neutron spectrum must be calculated. The spectrum was calculated for the 300-degree capsule by analyzing the octant adjacent to the capsule, and the results were reported in Reference [ST84]. An analysis was performed to determine if the assumption of octal symmetry for he 30-degree capsule is reasonable. This assumption precludes the need to calculate the spectrum for the 30-degree capsule.

K-3

Wt

~

~

Niagara Mohawk supplied bundle-cycle-exposure data for the core octant adjacent to the 30-degree capsule position. We compared the ratio of the bundle-cycle exposure to the core-average-cycle exposure for the appropriate geometrically symmetric bundles in the outer core region. Most of these ratios were within a few percent. Therefore, the assumption of octal symmetry is reasonable and results in a small error in the effective cross sections.

Calculations of the flux and fluence were made using the DECAY code. The reactor power history was supplied in a private communication'C084]. The effective cross sections and nuclear constants were reported in Reference [ST84].

The E>0.1 MeV and E>1 ' MeV full power flux and fluence calculated from initial startup to March, 1979, are given in Tables K-2 and K-3 respectively. For each of the dosimeter wires, the average of the flux and fluence was determined by averaging over all the Fe and Cu wires. It was not possible to estimate the flux for the Ni dosimeters, since the Co-58 activity was not detectable because of the long decay time following removal of the capsule.

The accuracy of the fluence values calculated is approximately +5 percent accuracy, uncertainties in neutron spectrum and spectrum-averaged cross sections result in the larger variances in the computed flux and fluence values.

K-4

TABLE K-1. MEASURED ACTIVITY OF THE 30-D GREE SURVEILLANCE CAPSULE DOSIMETRY WIRES Sample Measured Activity* Wire Composition ID Nuclide (dps mg ') Percent Pl-Ni.l Co-58 <0.043 99.9271 Ni, 0 '729 Co Co-60 192 + 8 Pl-Ni.2 Co-58 <0 '48 Co-60 198 + 8 P2-Ni Co-58 <0.040 Co-60 176 + 7 P3-Ni Co-58 <0.041 Co-60 162 Pl-Fe Mn-54 1.78 + 0.07 Commercially Pure Fe (99. 865)

P2-Fe Mn-54 1.68 + 0.07 P3-Fe Mn-54 1 64

~ + 0 07 P2-CU Co-60 8.48 + 0.32 99.999 CU P3-CU Co-60 7.87 + 0.31

• Decay corrected to November 30, 1984.

TABLE K-2. FLUX AND FLUENCE VALUES WITH ENERGY GREATER THAN 0.1 MeV AT THE NINE MILE POINT-UNIT 1 SURVEILLANCE CAPSULE (30-DEGREE AZIMUTHAL POSITION)

Dosimeter Full Power Flux Energy Material (n/cm /sec) x 10 Fluence'n/cm

) x 10'.1 MeV Fe (P1-Fe) 3.8 6.9 (P2-Fe) 3.6 6.6 (P3-Fe) 3.5 6 4

~

Average of Fe 3.6 6.6 Cu (P2-Cu) 3.3 6.1 (P3-CU) 3.1 5.7 Average of Cu 3.2 5.9 Average of Cu and Fe 3.4 6.3

• Fluence based on 2117.8 equivalent full power days of operation.

Reference [Ma85a]

TABLE K-3. FLUX AND FLUENCE VALUES WITH ENERGY GREATER THAN 1.0 MeV AT THE NINE MILE -POINT-UNIT 1 SURVEILLANCE CAPSULE (30-DEGREE AZIMUTHAL POSITION)

Energy Dosimeter Material (n/cm'/sec) x 10'luence Full Power Flux (n/cm~) x 10'~

1.0 MeV Fe (P1-Fe) 2.2 4.0 (P2-Fe) 2' 3.8 (P3-Fe) 2.0 3.7 Average of Fe- 2.1 3.8 Cu (P2-Cu) 1.9 3.5 (P3-Cu) 1.8 3.3 Average of Cu 1.9 3.4 Average of Cu and Fe 2.0 3.6

• Fluence based on 2117.8. equivalent full power days of operation.

Reference [Ma85a]

K.2 300 DEGREE CAPSULE DATA

'%ll

.2.1 Anal tical Method The determination of the neutron flux at the capsule, and subsequently in the pressure vessel wall, requires the completion of thre'e procedures. First, the disintegration rate of the product isotope per unit mass of the flux monitor must be determined. Second, in order to find a spectrum-averaged reaction cross section at the capsule location, the neutron energy spectrum must be calculated. Third, the neutron flux at the capsule must be found by calculations involving the counting rate data, the spectrum-averaged cross sections, and the operating history of the reactor.

The energy and spatial distribution of neutron flux in the reactor were calculated using the DOT 4.3 computer program '~~.

OT solves the Boltzmann transport equation in two-dimensional geometry using the method of discrete ordinates. Balance equations are solved for the density of particles moving along discrete directions in each cell of a two-dimensional spatial mesh.'nisotropic scattering is treated using a Legendre expansion of arbitrary order.

The two-dimensional geometry that was used to model the Nine Mile Point-Unit 1 Reactor is shown in Figure K-1. As seen, there are 16 circumferential meshes and 66 radial meshes. The capsule includes circumferential meshes 6 and 7 and radial meshes 54 and 55. Third order scattering was used (P~) and 48 angular directions of neutron travel (24 positive and 24 negative) were used (S8 quadrature). Neutron energies were divided into 47 roups with energies from 17.3 MeV to 10 eV. The 47 group

~

K-9

k structure is that of the RSIC Data Library DLC75/BUGLE 80 and neutron absorption, scattering, and fission cross sections used are those supplied by this librar'y. The core shroud is Type 304 stainless steel. The capsule is also modeled as a solid piece of Type 304 stainless steel. The reactor pressure vessel wall is SA302B steel. The reactor core was mocked up as homogenized fuel and water having the densities found in the operating reactor. The water in the core region has a density consistent with saturation conditions at the operating pressure of 1050 psia and a core-averaged steam volume fraction of 0.30 ~

The water in 'the downcomer has a density consistent with an inlet subcooling of 23 Btu/pound. Finally, the fuel was a source of neutrons having a U-235 fission energy spectrum. The relative ower in the assemblies nearest the capsule, during he interval the capsule was in the reactor, is shown in Figure K-17'~~"'. A plane view of the Nine Mile Point Reactor physical geometry at the core midplane is shown in Figure K-1 and because of symmetry includes only a 1/8th segment.

The neutron spectrum at the capsule center, as calculated by DOT, is shown in Figure K-2. Also shown for comparison is the fission spectrum. The fission spectrum was normalized to contain one neutron/cm~/sec above 1.0 MeV. The neutron spectrum at the capsule center was normalized to contain the same flux as the fission spectrum at 1.0 MeV energy. As can be seen, the capsule spectrum is considerably harder than the fission spectrum. This is caused by neutron travel through water.

K-10

di 0

15 13 280 300 Capsule 240 200 5

.56 $6 $3 .49 .39 O~

) C V .95 .99 "'.93 .93 .79 $0 ~a 4y 160 1.02 1,08 1.06 1.05 .93 .85 ~ 50 Ng E V 0 4p 4y u 1,09 1,19 1.12 1.15 1.04 1.04 ~ 53 Cg

~ 52

.99 .88 120 1.0 1,07 4~

4~

80 40 0

0 80 120 240 280 Distance From Core. Center (cm)

FIGURE K-l. NINE MILE POINT CORE, INTERNAL VESSEL STRUCTURES, AND VESSEL WALL GEOMETRY USED IN THE DOT CALCULATION

f'I 1.0 Fission Spectrum Neutron Spectrum at Capsule Le I

10.1 L

I L

I I

I I

I I

L~)

> 102 I g I I

ul I C

Leans E I I

I k I I

Le w~~

I 10'3 I I

I I

I I

I lcm mme 10 I 10' 12 16 24 Neutron Energy (MeV)

FZGURE K-2 COMPARZSON OF DOT SPECTRUM AT 300 DEGREE NZNE MILE POZNT SURVEILLANCE CAPSULE WZTH FZSSZON SPECTRUM K-12

Based upon the fluxes calculated by DOT at radial meshes 54 and 55 and circumfenential meshes 6 and 7 (the meshes used to represent the capsule and the region in which the flux monitors were placed), effective cross sections a~ (E > 0.1 MeV) and G~ (E

> 1.0 MeV) defined as:

cr(E > E,)

0 (E) dE E,

were calculated for iron, nickel, and copper. The results are shown in Table K-4.

Other nuclear constants needed in the third step of the flux-finding procedure are given in Table K-5.

In the third step, the full power flux at the capsule location is determined from the radioactivity induced in the monitor foils, the effective cross sections calculated for the monitor elements, and the power history of the reactor during capsule exposure. The fluence at the capsule is then calculated from the integrated power output of the reactor during the exposure interval using the following equation:

) (E > E ) = A/ [Na(E > E ) C]

This equation was used to find fluxes based on the measured surveillance capsule activations. The time intervals were taken as one month each and a time integrated relative power value for each month was used for the fractional power level values.

Calculations of the flux and fluence were made with the DECAY code. The reactor power history was supplied in a private communication'

~'

TABLE K-4. CROSS-SECTIONS FOR THE IRRADIATED FLUX MONITORS CALCULATED FROM FLUXES AT CAPSULE CENTER OF THE NINE MILE POINT 300-DEGREE SURVEILLANCE CAPSULE Dosimeter Cross-Sections Barns Material E > 0.1 MeV E > 1.0 MeV Fe 1.01 x 10 1.76 x 10 i Cu 1.77 x 10 ~

3.09 x 10 ~

Ni 1.28 x 10 i 2.23 x 10 i TABLE K-5. CONSTANTS USED'IIN DOSIMETRY CALCULATIONS FOR THE NINE MILE POINT 300-DEGREE SURVEILLANCE CAPSULE Isotopic Threshold Target Abundance Energy, Product Reaction Percent Percent MeV Half-Life

~'Fe (n, p) ~'Mn 99.865 Fe 5.82 2.5 312.6 days "Cu (n, a) "Co 99.999 Cu 69.17 6.1 5.27 years seNi (n, p) ssC 99.927 Ni 67.77 2.1 71.2 days K-14

0 K.2.2. Dosimetr Results

~ ~ ~

The surveillance capsule was located at the 300-degree azimuthal position at approximately the core midplane position and at 7/16- inch from the inner pressure vessel wall. This capsule was in the reactor for" 2913 equivalent full power days or about 7.98 equivalent full power years. The Nine Mile Point Nuclear Generating Plant design thermal output is 1850 MW.

The neutron monitor wires from Charpy packets P7 and P8 wee counted to determine their specific activity. The recommended ASTM procedures ~~~" ~~"'~'" As~"'~' were followed in determining the specific activity of the wires. Each dosimeter monitor consisted of an approximately 4-inch length of wire which was rolled into a small coil for counting. The count rate was determined for each wire ~ The fast flux and fluence calculated using the count rate therefore represented an average over the 4-inch length of that wire. The E > O.l MeV and E > 1.0 MeV full power flux and fluence calculated from initial startup to March 1982 are given in Table K-6 and Table K-7., respectively, for each f the dosimeter wires along with the average of the flux and fluence derived from each wire and the average values for Fe, Cu, and Ni. In addition, the average values of the results for Fe and Cu are given. The Ni results were not used because the very short half life makes its results dependent on only the latest operating history.

Using the average fluxes of 3.32 x 10 n/cm /sec for E > 0.1 MeV and 1.90 x 10'/cm'/sec for E > 1.0 MeV, the fluxes at full power at the inside of the pressure vessel wall, at 1/4 T and at

k tf A4

3/4 T directly behind the capsule (300-degree orientation) and at the maximum position (285.66-degree orientation) were calculated.

The flux results are tabulated in Table K-8. The end of life (EOL) fluences were also calculated and tabulated in Table K-8 assuming a reactor pressure vessel lifetime of 40 years and the reactor operated at 80 percent of full power. The fine mesh and time integrated relative power values shown in Figure K-8 for each fuel assembly was used in the DOT 4.3 code to generate the values in Table K-8. A plot of neutron flux (E > 1.0 MeV) as a function of azimuthal angle (in degrees) is shown in Figure K-3.

The fluence values at the maximum position for inner vessel wall, 1/4 T and 3/4 T are plotted as a function of time in equivalent full power years (EFPY) for the Nine Mile Point pressure vessel in Figure K-4. The lead factor, ice., the ratio of the flux (E >

~

1.0 MeV) at the surveillance capsule to the largest flux (E > 1.0

~

MeV) received by the vessel wall at any azimuthal location, is approximately 0.68 (1.90 x 10'/2.80 x 10') at the vessel surface.

This result indicates that the flux at the capsule actually lags the flux at certain vessel wall positions. The lead factors at the pressure vessel 1/4 T and 3/4 T positions were calculated to be 0.99 (1.90 x 10 /1.91 x 10 ) and 3.70 (1.90 x 10 /5.14 x 10 ),

respectively.

The accuracy of the fluence values generated is estimated to be +20 percent. Although specific activities of fluence monitor wires can be determined to +5 percent accuracies, uncertainties in neutron spectrum and spectrum averaged cross K-16

0 1010

~ Position of 300 Degree Capsule A

N h 10> inner gq~~

C X

u C

0 '<<r z

~h u

3/g y 108 270 290 300 310 320 330 Azimuthal Angle (degree)

FIGURE K-3. CALCULATED FLUX AT PRESSURE VESSEL INNER WALL, 1/4 T THICKNESS AND 3/4 T THICKNESS AS A FUNCTION OF AZIMUTHAL ANGLE K-17

1018 g(4'1 1017 1016 0 10 15 20 25 30 Time (full power yean)

FIGURE K-4. FLUENCE AT INNER WALL 1/4 T AND 3/4 T POSITION AS A FUNCTION OF TIME FOR THE NINE MILE POINT UNIT 1 REACTOR VESSEL

p's sections result in the larger variances in the computed flux and fluence values.

The rate of displacements per atom was also calculated using the cross sections for displacement available with the DETAN code. Table K-9 shows calculated values of displacements per atom per second at full power in the pressure vessel wall behind the capsule and in the pressure vessel wall at the angle of peak fluence in the wall. Table K-9 also shows calculated values of displacements per atom at these same positions in the wall for 7.98 effective full power years of operation (to time of

'capsule removal) and for 32 effective full power years of operation.

K-l9

0 TABLE K-6. FLUX AND FLUENCE VALUES WITH ENERGY GREATER THAN 0.1 MeV AT THE NINE MILE POINT-UNIT 1 SURVEILLANCE CAPSULE (300-DEGREE AZIMUTHAL POSITION)

Energy Dosimeter Material (n/cm /sec) x 10'luence*

Full Power Flux (n/cm') x 10" O.l MeV Fe (P7) 3.549 8.933 (P8) 3 '10 8.078 Average of Fe 3.380 8.506 Cu (P7) 3.253 8.188 (P8) 3.253 8.188 Average of Cu 3.253 8.188 Ni (P7) 3.230 8 '30 (P8) 3.117 7.844 Average of Ni 3.174 7.987 Average of Fe and Cu 3.32 8.35

• Fluence based on 2913.1 equivalent full power days of operation.

a) P7 refers to bottom packet b) P9 refers to top packet Reference -[ST84]

0 TABLE K-7. FLUX AND FLUENCE VALUES WITH ENERGY GREATER THAN 1.0 MeV AT THE NINE MILE POINT-UNIT 1 SURVEILLANCE CAPSULE (300-DEGREE AZIMUTHAL POSITION)

Energy Dosimeter Material Full (n/cm /sec) x 10'luence Power Flux (n/cm ) x 10" 1.0 MeV Fe (P7) 2.033 5.118

~

'P8) 1.839 4.629 Average of Fe 1.936 4.874 CU (P7) 1. 864 4.691 (P8) 1. 864 4.691 Average of Cu 1. 864 4.691 Ni (P7) 1.851 4. 659 (P8) 1.786 4. 494 Average of Ni 1.819 4.577 Average of Fe and Cu 1.90 4.78

• Fluence based on 2913.1 equivalent full power days of operation.

a) P7 refers to bottom packet b) P9 refers to top packet Reference [ST84]

TABLE K-8. FLUX AND FLUENCE IN THE PRESSURE VESSEL WALL OF THE NINE MILE POINT-UNIT 1 REACTOR BEHIND THE SURVEILLANCE CAPSULE (300-DEGREE) AND AT THE AZIMUTHAL ANGLE OF MAXIMUM FLUX IN THE VESSEL WALL (285.66-DEGREE)

Fluence in Vessel Wall Full Power Flux in Vessel Wall Behind Capsule (300') Maximum (285.66')

Energy Location Behind Capsule (300') Maximum (285.66') March 82 (1) EOL (2) March 82 (1) EOL (2)

(MeV) (n/cm'/sec x 10') (n/cm'/sec x 10') (n/cm x 10' (n/cm x 10") (n/cm~ x 10' (n/cm x 10")

>0.1 Surface 3.19 5.56 8.03 3.22 1.40 5.61

>0. 1 1/4T 2.85 5.01 7.17 2.88 1 ~ 26 5.08 I

>O.l 3/4T 1.31 2.26 3.30 1.32 0.569 2.28

>1.0 Surface 1 ~ 51 2.80 3.80 1.52 0.705 2.83

>1.0 1/4T 1.02 1.91 2.57 1.03 0 '81 1.93

>1.0 3/4T 0.281 0.514 0.707 0.284 0.129 0.51 (1) Fluence based on 7.98 effective full power years of operation.

• (2) Fluence based on 32 effective full power years of operation.

Reference [ST84]

g k

TABLE K-9. ATOM DISPLACEMENTS IN THE PRESSURE VESSEL WALL OF THE NINE MILE POINT-UNIT 1 REACTOR Dis lacements er Atom Dis lacements er Atom er Second Behind Capsule Maximum (285. 66')

Location Behind Capsule Maximum (285 66')

~ March 82 (1) EOL (2) March 82 (1) EOL (2)

Surface 2.56x10 ~

4.82xlO ~

0. 644 2.582 1.214 4.868 1/4T 1.79x10 9 3.28xlO 9 0.452 1.813 0.825 3 '08 3/4T 6.19x10 " 1.10xlO 9 0.156 0.626 0.277 1.111 (1) Fluence based on 7.98 effective full power years of operation.

(2) Fluence based on 32 effective full power years of operation.

Reference [ST84]

APPENDIX L NINE MILE POINT VNIT 1 RT~ DETERMINATION

i, FINAL REPORT entitled NINE MILE POINT UNIT 1 RT r DETERMINATION to

, Niagara Mohawk Power Corporation 301 Plainfield Road Syracuse, NY 13212 by Dr. Michael P. Manahan, Sr.

September 28, 1990 NPN Consultin~

213 Teaherry Circle State CO11egeg PA 16803 (814) 234-8860

0 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT This document was prepared by Dr. Michael P. Manahan. The information contained in this report is believed by Dr. Manahan to be an accurate and true representation of the facts known, obtained or provided to Dr. Manahan at the time this report was prepared.

Dr. Manahan does not make any representation or warranty (express or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor does he assume any responsibility for liability or damage of any kind which may result from such use of such information.

L-3

t ia >i COPYRIGHT NOTICE l

The Ropy methodo ogy is the property of Dr . Manahan . Permission is hereby granted to Niagara Mohawk to use the results obtained using

,.this methodology, for the Nine Mile Point Unit 1 P-T curve analysis.

Use of the methodology, other than for Nine Mile Point Unit 1, is forbidden without prior written approval.

M.~ P.~ Manahan, 1990 L-4

W4 TABLE OF CONTENTS

1.0 INTRODUCTION

0 ~ ~ 7 1.1 Purpose 2.0 WELD MATERIAL ANALYSIS ~ ~ ~ ~ ~ 8 2.1 Weld W5214/5G13F Analysis 8 2.1.1 cr, Analysis 9 2 1 2

~ ~

Ropy Ana lysis 11 2.1.3 Summary of Findings 16 2.2 Beltline Weld Analysis 16

-.'.3.0 PLATE MATERIAL ANALYSIS ~ ~ ~ 19 3.1 Base Metal cr, ~ ~ ~ 19 3.2 RTp~ Analysis. 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 19

4.0 CONCLUSION

S 28

.0 REFERENCES. . . . . . . . . . . . . . . . . . . . '. . . 30

LIST OF FIGURES 2-1 Charpy Impact Energy Versus Test Temperature for Irradiated Weld Specimens from the Nine Mile Point Unit 1 300 Degree Capsule . . . . . . . . . . . . . 10 2-2 RT~~ Analysis for Surveillance Capsule Weld Material W5214/5G13F . . . . . . . . . . . .

. . . . . . 15 2-3 RT~~ Analysis for Beltline Weld 86054B/4E5F 18 3-1 Charpy Data Fit for Base Metal Plate G-8-3/G-8-4 22 3-2 Charpy Data Fit for Base Metal

.Plate G-8-3/G-8-4 (T-L) 23 3-3 Charpy Data Fit for Base Metal Plate G-8-1-5. .24 Charpy Data Fit for Base Metal Plate G-307-3 .25

- Charpy Data Fit Plate G-307-4 for Base Metal

.26 3-6 Charpy Data Fit for Base Metal Plate G-307-10 ~ ~ ~ 27 L-6

LIST OF TABLES 2-1 Summary of RTQQ7 0,, and Charpy Indices for Weld W5214/5G13F . . . . . . . . . . . . . . . . . . 16 2-2 RT~~ Data for Beltline Welds . ... . . . . . . . . . . . 17 3-1 RT+~, 0' and Charpy Indices for Unzrradiated Beltline Plates . . . . . . . . . . . . . . 21 L-7

1.0 INTRODUCTION

1.1 Pur ose An analysis was perf1ormed to establish best estimate values for the RT~~ and a, term used in Regulatory Guide 1.99 (Rev. 2) calculations for the Nine Mile Point Unit 1 (NMP1) beltline plate and weld materials. Sufficient data are not available to determine the RTgp~ in strict conformance with the current ASME code rules. In the case of the weld metals, only three unirradiated Charpy tests were conducted at 10'F.

Charpy transition behavior data are available for the plate materials, however, there are no drop weight data for all of the beltline materials except for plate G-8-3/G-8-4. The methodology used was first described in {MA85] and was specifically designed to address ,the problem of RTz~

determination in cases where there are not sufficient data available to satisfy the ASME code requirements. The results of these analyses are presented in Sections 2.0 and 3.0 of this report. Conclusions and recommendations concerning the use of these data are provided in Section 4.0.

L-8

2. 0 WELD MATERIAL ANALYSIS The objective of the weld material analysis is to calculate a statistically meaningful o, for the beltline welds which can be used in the Margin term of the Regulatory Guide 1.99-Rev. 2 (RG1.99(2)) model. Also, the initial RTpT of each of the beltline welds and the surveillance capsule material was calculated using Manahan's statistically based calculative procedure [MA85b].

The only unirradiated fracture behavior data available for the weld materials is Charpy data at 10'F [MA90, BY90]. Irradiated data for the surveillance capsule material, designated weld W5214/5G13F (wire heat no./flux lot no.), is also presented in Reference [MA90]. Both the irradiated surveillance data and the unirradiated data for each weld were used to calculate o, and the RTpT of each beltline weld. As described in Reference [MA90], the base metal Charpy specimens were machined with their long axes rallel to the plate rolling direction. Therefore, the crack plane is oriented transverse to the material grain (L-T orientation). The weld specimen long axis was machined transverse to the weld direction. Since there are no microstructural data available for the welds, it 'has been assumed that the material is isotropic and no correction for grain orientation was made in the RTgpT analysis . Unl ike base metal pl ates, the weld material generally does not exhibit grain orientation effects and, therefore, the Charpy specimen orientation is not critical for weld.

materials.

2.1 Weld W5214 5G13F Anal sis Weld W5214/5G13F is the surveillance capsule weld.

Unfortunately, this weld was not made using the same wire or flux as the beltline welds. However, the weld materials were manufactured by the same suppliers for the beltline materials (RACO g3 wire, Arcos B5 flux) and the Cu and Ni content is representative of the beltline weld 1248/4M2F. [BY90, MA90]

(Data on the Cu and Ni content of the other two beltline welds L-9

is not available.) Therefore, it has been assumed that the capsule weld material is similar to the beltline welds in terms of its mechanical behavior trend.

The irradiated Charpy data for the capsule weld material was analyzed using the SAM McFRAC code [McFRAC).

The SAM McFRAC code has been QA verified in accordance with The Pennsylvania State University, Nuclear Quality Assurance Program. This code is based on a non-linear, least squares, regression analysis using the Weibull statistic. The Weibull statistic has been shown to be the correct statistic for analysis of fracture data by considering the microstructural mechanisms involved in the fracture of ferritic, pressure vessel steels [MA85a].

The confidence bands are measures of "the goodness of fit" and do not indicate the classical 954 statistical error spread. This uncertainty must be analyzed using conventional statistical methods. However, the McFRAC confidence intervals are used to measure confidence in the fit of a particular data set as well as the inherent scatter due to the fracture process. These error bands must be calculated, particularly, for sparse data sets, because in many cases the ability to fit sparse data drives the uncertainty. The McFRAC analysis for the irradiated capsule weld is shown in Figure 2-1.

The o, term is the uncertainty in the initial RTgpy determination.Section III of the ASME code requires that, both drop weight tests, to determine the NDTT, and Charpy tests be performed. The NDTT temperature deter-mined by following ASTM E-208 is the RT>> provided that, at 60'F above the NDTT, at least 50 ft-lbs. of energy and 35 mils lateral expansion are obtained in Charpy speci-mens with crack planes oriented in the weak direction.

The weak direction is transverse to the direction of maximum working (T-L orientation) for base metal.

e 10 NINE INILE POINT UNIT 5

~

IRRADIATED WELD 300 DEGREE CAPSULE{'I,2) EXPERIMENTAL DATA 150 WEIBULL FIT Cl TRANSITION 125 gg Jk I-I WEIBULL FIT UPPER SHELF IL k 100 HYPERBOLIC C9 TANGENT FIT 75 4

CONFIDENCE 4

/ja LIMIT (95%)

50 V CONFIDENCE LIMIT (95%)

26 4 j 4 A

CONFIDENCE LIMIT (95%)

JJ

'I 50 CONFIDENCE 150 LIMIT (95%)

TEST TEMPERATURE (F)

FIGURE 2-1: CHARPY IMPACT ENERGY VERSUS TEST TEMPERATURE FOR IRRADIATED WELD SPECIMENS FROM THE NINE MILE POINT UNIT 1 300 DEGREE CAPSULE

11 Therefore, cr, should reflect the uncertainty in either the drop weight test or the Charpy test depending on which test ultimately determined the RTDz. As described below, Charpy data are used to determine the Ropy and there fore, a, is determined by estimating the uncertainty in temperature at the 50 ft-lb. energy level.

Based on examination of the confidence in the McFRAC fit at 50 ft-lb., cr,(fit) = 14.0'F. Analysis of the statistical 954 confidence interval for the data yields cr,(experimental) = 12.5'F. Based on examination of the LWR data base, Odette [OD86] has shown for welds that, cr,(data base) = 18.0'F.

The uncertainty for weld W5214/5G13F is higher than that of the LWR data base because of the sparsity of data in the transition region. Additional testing would reduce o,(fit) and e,(experimental). Therefore, the final o, was determined using a weighted average of plant specific and generic data conservatively assuming there were only ten data points in the data base analyzed by Odette, to yield cr, = 17'F. This value for cr, is recommended for use in [RG1.99] Margin term calculations for all of the beltline weld materials.

Many operating power plants were built when current regulations concerning establishment of the initial RT~~

were not in force. Zn many cases, insufficient drop weight and Charpy data are available to determine the RT~~ in accordance with the current Section XII of the ASME code. Prior to 1972, the ASME code recpxired that the average of three Charpy specimens be at least 30 ft-lb. at a designated temperature, with no single impact energy less than 25 ft-lb. These are the rules which were used for the vessel. The NRC Branch Technical Position, MTEB 5-2, contains a procedure for estimating the RT~~ based on generic data. However, this approach

12 is overly conservative for some materials and does not take advantage of the use of limited unirradiated nor irradiated data in the analysis. GE also has a generic model which is based on a transition region slope of .5 ft-lb./'F. However, this model, like the NRC model, is generic and lacks statistical rigor.

Neutron irradiation of pressure vessel materials causes:

an increase in the 30 ft-lb. transition temperature

2) a lowering of the upper shelf

3) . a decrease in the slope of the transition region [OD86]

In addition, Odette has shown that the transition region occurs over an approximately constant interval of temperature. Odette pointed out that this fact, in concert with a continuous decrease of upper shelf energy (USE), requires that the 'transition region slope must decrease with irradiation. The average value of the transition temperature range, 6T~, is about 200 F +45 F for welds. These data, along with Odette's yield strength model, can be used to provide an accurate estimate of the RTQ7 of the beltline weld materials assuming that the NDTT occurs at a lower temperature than the temperature at which three Charpy specimens would exhibit 50 ft-lbs. of energy minus 60. This assumption is accurate for most pressure vessel materials because the NDTT is expected to occur at or near the brittleness transition region near the onset of the lower shelf. Analysis of the LWR ASTM A533B data base substantiates this assumption. In the 23 cases analyzed, the NDTT was less than or equal to the RT~~. In 15 out of 23 cases, the RT~~ was determined from the Charpy data. The average difference between the actual RTgp~ and the RTgp7 estimated using the Ref erence [ MA85 ] approach

I I

13 was 8'F for welds and 10'F for base metal. The average net deviation is about 4.3'F for plate and weld. A similar analysis was conducted for ASTM A302B data. In all of the 17 cases analyzed the NDTT was less than or equal to the RT~T. The RTgpT was determined from Charpy data in 13 of the 17 cases. The average between the ASME code determined RT~T and that determined using Reference

[MA85] was 8', and the average net deviation was 4.4'F for the base metal.

Given these facts and observations, the approach to RTgp7 determination is to determine a temperature, T50T (for T-L orientation), at which three Charpy specimens would be expected to yield greater than 50 ft-lbs of absorbed energy. The T50T is determined by analyzing the uncertainty in the Charpy data in the transition region.

The Tzo~ temperature is defined as the, temperature at which the mean Charpy curve minus 50 ft-lbs equals 20E.

Once the T50~ temperature is determined, the RTgp7 is taken to be T50~'minus 60'F. This basic approach was used for both plates and welds. In the weld analysis, additional steps were followed since full Charpy curves were not available for the unirradiated welds. The procedure used to calculate the RTN07 for the Oyster Creek welds is as follows:

~Ste 1 Using Odette's yield strength model [OD86],

calculate the unirradiated upper shelf energy (USE ).

~Ste 2 Draw a horizontal line at the USE'evel and pass a line through the unirradiated data which intersects the USE'ine such that the transition region spans 2004F. Verify the reasonableness of the slope by comparison with the irradiated transition region slope.

~Ste 3 Estimate the 954 confidence band for energy at the 50 ft-lb. level.

L-14

v 14

~Ste 4 Using the results of Steps 2) and 3),

determine the temperature at which 50 ft-lb.

is achieved and subtract 60'F to obtain the Rior

~Ste S (Optional) Zf sufficient data exists, use the Odette yield strength model to calculate the transition temperature shift at the 30 ft-lb.

level (6T>Q) and compare with the ST3Q obtained above and with the RG1.99(2) model prediction.

For weld 85214/5G13F, the unirradiated USE was calculated to be: USE' 128 ft-lb. The results of the Step 2) graphical analyses are presented in Figure 2-2.

As discussed in Reference [OD86], the entire irradiated Charpy curve can be predicted knowing only the change in yield strength due to irradiation and the unirradiated Charpy curve. This procedure was woxked in reverse using the irradiated Charpy curve and the change in yield strength. The measured slope of the irradiated Charpy curve is 0.539 ft-lb/'F which is in good agreement with the irradiated data as shown'n Figure 2-2. The calculated slope of the unirradiated transition region is 0.645 ft-lb/'F, which is slightly larger than the irradiated curve, as expected.

The third step of the RTQT analysis calls for an assessment of the statistical 954 confidence band (2u E )

for energy measurement at the 50 ft-lb. level. Two separate calculations were performed. The first approach uses the 2'ariation in energy at the 50 ft-lb. level.

This results in: 2aE(fit) = 13.5 ft-lb. The second approach consists of an assessment of the 954 confidence in the experimental data. This approach yields:

2v~(experimental) = 12.0 ft-lb. The procedure defines 2@~ as the larger of 20E (fit) and 2'~ (experimental),

with the lower bound being 10 ft-lb. Therefore, for weld W5214/5G13F: 2aE = 10 ft-lb.

~ e 'I VI A.

15 NINE INILE POINT UNIT 1 VfELD 52 1 4/661 3F (SURVEILLANCE %ELD) ~ IRRADIATED DATA 150 NEIBULL FIT TRANSITION 126 NEIBULL FIT I gal@

A giikgi UPPER SHELF I- HYPERBOLIC 100 r TANGENT FIT C9 76

/r CONFIDENCE LIMIT (85%)"

CONFIDENCE K re a LIMIT (95%)

4 CONFIDENCE 4

4 L LIMIT (85%)

Oa A CONFIDENCE LIMIT (95%)

UNIRRAOIATED DATA 300 UNIRRADIATED CHABPY CURVE TEST TEMPERATURE (F)

I F GURE 2 2 RTg) 7 ANALYSIS FOR SURVE ILLANCE CAPSULE WELD MATERIAL W5214/5G13F

16 Since the three data points for weld W5214/5G13F were over 50 ft-lb., the RT~z is taken to be:

RT~~ (weld W5214/5G13F) = 10'F - 60 F = -50'F. In order to demonstrate the validity of the RT~~ approach, the Reference [MA85) procedure was followed. In accordance with Step 4, three Charpy specimens with energies in excess of 50 ft-lb. would be expected at a test temperature of 20'F. Therefore, the RTz~ using this approach would be -40'F, which is in close agreement and slightly conservative when compared with the measured Charpy data.

2.1.3 Summar of Findin s Based on the statistical analysis presented above, a value of o> of 17'F is justified for the beltline welds. An initial RTgQ7 of -50 has been established for weld W5214/5G13F. Also, unirradiated Charpy indices have been established for the capsule weld material. These data are summarized in Table 2-1..

TABLE 2 1

SUMMARY

OF RT ()7~0) AND CHARPY INDICES FOR WELD W5214 5G13F Descri tion Best, Estimate Value cr, 17 F RTvov -50 F 2G~ 13.5 ft-lb.

T3Q unirradiated -32 F Upper Shelf Energy (unirradiated) 128 ft-lb.

2.2 Beltline Weld Anal sis The RT~~ determination for the beltline weld 86054B/4E5F was made as described in Section 2.1 for weld W5214/5G13F L-17

0 17 because the data at +10 are well above 50 ft-lb. However, the calculated RTgQ7 using the [MA85] procedure is -53'F. Since there is a small difference (3'F) between the calculated and experimental value, the RT~~ was taken to be -50'F. The data for weldS 1248/4M2F AND 1248/4K13F were above 50 ft-lb. at 10 F and theref ol e the Ropy was determined experimenta l1 y for these materials as well. The results are presented in Table 2-2 and Figure 2-3.

TABLE 2 2 RT ()~

DATA FOR BELTLINE WELDS Weld Identification RT or~F 86054B/4E5F -50 1248/4M2F -50

'248/4K13F

-50

E' I

18 NINE MILE POINT UNIT I NELD 860548/4ESF 150 125 I

I 100

+ UNIRRRDIATED C DATA V5 K UNIRRADIATED CHARPY CURVE 50 300 TEST TEMPERATURE (F)

F FIGURE 2 3 I RT~)~ ANALYS S FOR BELTLINE WELD 8 6 054 B/4E4F L-19

t' 19

3. 0 PLATE MATERIAL ANALYSIS 1

In the case of the base metal plates, full unirradiated Charpy curves are available for each beltline plate. The surveillance capsule base metal'pecimens were fabricated from plate G-8-1

[MA90).

It is possible to calculate a o> for each plate as was done for the irradiated weld. However, a value of a, = 0 is recommended for all the plates except G-8-3/G-8-4. The o< for G-8-3/G-8-4 is taken as 2'F since experimental data was used and conventional ASME test practices followed. Justification of u, = 0 for the remaining plates comes from the fact that Manahan's RTz~ determination is generally conservative, the T)pg is taken to be a point where the ASME Charpy criterion is expected to be satisfied with high confidence, and the L-T to T-L orientation correction was taken to be 24 F which is 4 F higher than that recommended in Branch Technical Position MTEB 5-2. The value of cr, = 2 was obtained by taking the Ropy found using Manahan's method, subtracting the ASME determined l

Ropy and dividing by 2 ( rounding up ) . The resu ts of the o',

analysis are given in Table 3-1.

The RTgpz was determined in a similar manner as described in Section 2.1.2, except that the temperature at which 50 ft-lb. of energy would be obtained in three Charpy specimens could be determined directly from the Charpy curves. First, the unirradiated data was- fit using the SAM McFRAC code.

Then, the 954 confidence interval in energy (2cr<) was determined using the more conservative value of either the fit uncertainty at the 50 ft-lb level (2cr~(fit)) or the deviation of the data from the mean in the transition region (2v~

(experimental)). If both of these measures are less than 10 ft-lbs, then 10 ft-lb. was taken as the 2cr< value. Otherwise, the larger of 2az(fit) or 2vz(experimental) was used. The 10

20 ft-lb. uncertainty was judged to be a reasonable and conservative value based on observation of many data sets in the LWR data base.

After 2a'~ was determined, the temperature, T><< (L-T orientation), at which three Charpy specimens would exhibit 50 ft-lbs of absorbed energy was determined. This temperature was taken to be the temperature at which the mean Charpy curve minus 50 ft-lbs. equals 2cr~. Once, the T><< temperature is determined, the RTzz (L-T) is taken to be T><< minus 60'F.

The Charpy specimens tested for the plate materials had an L-T orientation. The ASME code requires testing using the T-L orientation because it is the limiting orientation.

General Electric recommends adding 30'F to the RT~~ obtained C

using L-T specimens. Similarly, Branch Technical Position MTEB 5-2 recommends adding 20 F. Analysis of the EPRI data base presented in'MA89] indicates that 30'F is appropriate.

Therefore, in the absence of material specific data, MPM Consulting recommends using 30'F. However, reference [MA90]

reported an L-T to T-L Charpy curve transition of 24'F at the 30 ft-lb. level. Therefore, based on the material specific data, the L-T to T-L correction used herein was 24'F. The results of the analysis are summarized in Table 3-1 and the McFRAC fits are given in Figures 3-1 through 3-6.

I,kr TABLE 3-1: RT oT~<Tr AND CHARPY INDICES FOR UNIRRADIATED BELTLINE PLATES Best Estimate Data Plate Plate Plate Plate Plate Descri tion G-8-3 G-8-4 G-8-1 G-307-3 G-307-4 G-307-10

<7) 2 F 0 F OF 0 F 0 F RT07 (1) -24 F(L-7) 12 F(L-7) 40F <<-7) 16 F (L-7) -4 F (L-T)

(2) -3'F<7-L) 36 F(7-L) F (7 L) 40 F (T-L) 20 F (7-L)

RTNoT 2 <TE 11 ft-lb. 10 ft-lb. 11 ft-lb. 10 ft-lb. 10 ft-lb.

T30 -26OF<L-7) 8 F -14 F 0 F -3 F O'(7-L)

Upper Shelf Energy (

'unirradiated) 96 ft-lb.<L-T) 88 ft-lb. 112 ft-lb. 81ft-lb. 115 ft-lb.

68 ft-lb. (7-L)

L-T orientation. P-T calculations require RT+7 to be determined for the T-L orientation.

' T-L orientation. Obtained by adding 24 'F to the L-T orientation data for all plates except G-8-3/G-8-4. The value was determined experimentally.

(3) Fracture appearance data are not available for any beltline plates. Upper shelf data was identified by the shape of the curve fit.

he ' r 22 NINE MILE POINT UNIT I ~ EXPERIMENTAL UNIRRADIATED BASE METAL G&3/G&4 (0 2)

DATA WEIBULL FIT A

A TRANSITION CQ A L

100 I A+Jl S WEIBULL FIT I- UPPER SHELF U

80 HYPER8OLIC A

TANGENT FIT e

A A

CONFIDENCE LIMIT (95%)

CONFIDENCE A LIMIT (95%)

CONFIDENCE A

LIMIT (95%)

A

-'I 00 -50 0 50 'I 00 'I 60 200 260 CONFIDENCE LIMIT (95%)

TEST TEMPERATURE (F j FIGURE 3-1: CHARPY DATA FIT FOR BASE METAL PLATE G-8-3/G-8-4 L-23

23 NINE MILE POINT UNIT I ~ EXPERIMENTAL UNIRRADIArPD BASP METAL Q g 3 (QL)( I 2)

DATA 120 WEIBULL FIY IXt TRANSITION 100 I

I- HYPERBOLIC 80 TANGENT FIT Q CONFIDENCE eo LIMIT (95%)

CONFIDENCE 40 LIMIT (95%)

V 20 CONFIDENCE LIMIT (95%)

A CONFIDENCE

-'IOO -60,0 50 100 160 200 260 L'IMIT {95%)

TEST TEMPERATURE (F)

FIGURE 3-2: CHARPY DATA FIT FOR BASE METAL PLATE G-8-3/G-8-4 (T-L)

L-24

24 NINE MILE POINT UNIT I ~

UNIRRADIATED BASE'ETAL G 'I (l,2) EXPERIMENTAL DATA CZ0 WEIBULL FIY Ql TRANSITION 400 I

HYPERBOLIC IL TANGENT FIT 80 C CONFIDENCE ILI eO uMn (95%)

CONFIDENCE 40 uMn (95%)

V zo CONFIDENCE LIMIT (95%)

CONFIDENCE

.'IOQ-50 0 50 'I 00 160 200 260 uMn'95%)

TEST TEMPERATURE (F)

Fj:GURE 3-3: CHARPY DATA FIT FOR BASE METAL PLATE G-8-1 L-25

'C NINE MILE POINT UNIT I ~ EXPERIMENTAL UNIRRADIATED BASE lNETAL G-30V-3 (1 a)

DATA WEIBULL Fl'Y TRANSITION

'I 00 I

I HYPERBOLIC 80 TANGENT FIT I

PA C 0 CONFIDENCE 60 LIMIT (95%)

L L

SL l~ CONFIDENCE 40 LIMIT (95%)

V 20 CONFIDENCE LIMIT (95%)

CONFlD ENCE

-100 -50 0 50 'I 00 'I 50 200 260 LIMIT (95%)

TEST TEMPERATVRE (F j FIGURE 3-4: CHARPY DATA FIT FOR BASE METAL PLATE G-307-3

26 NINE INILE POINT UNIT 1

~

UNIRRADIATEP BASE METAL g 307 4 (1 1) EXPERIMENTAL DATA 120 WEIBULL FIT TRANSITION CQ 100 I WEIBULL FIT I AIL< UPPER SHELF LL 80 HYPERBOLIC TANGENT FIT CONFIDENCE uMn. (95%)

40 J L CONFIDENCE V A L

LIMn (e5%)

L 20 C CONFIDENCE uMn (95%)

0 CONFIDENCE

-100 -50 0 50 100 15Q 200 250 LIMIT (95%)

TEST TEMPERATURE {F)

FXGURE 3-5: CHARPY DATA FIT FOR BASE METAL PLATE G-307-4 L-27

ll 0

27 NINE MILE POINT UNIT 1 UNIRRADIATED BASE METAL Q-307-10 (1,2) ~ EXPERIMENTAL DATA 120 WEIBULL FIY TRANSITION 100 I-I HYPERBOLIC 80 TANGENT FIT J'

C CONFIDENCE 60 gk uMn (95%)

0 ~ 0 CONFIDENCE LIMIT (95%)

CONFIDENCE LIMIT (95%)

CONFIDENCE

-100 -50 0 50 'I 00 150 200 250 uMn (95%)

TEST TEMPERATURE (F)

FIGURE 3-6: CHARPY DATA FIT FOR BASE METAL PLATE G-307-10 L-28

28

4.0 CONCLUSION

S The following RT~z results were obtained:

RT~pz (T L) C)

Plate oF oF G-8-3/G-8-4 G-8-1 36

'-307-3 28 G-307-4 40

. G-3 07-.10 20 RTunz 0)

Weld oF oF W5214/5G13F -50 17 86054B/4E5F -50 17 1248/4K13F -50 17 1248/4M2F -50 17 The data reported here was compared with the calculations re-ported in, reference [MA90a]. The surveillance welds for Oyster Creek and NMPl are identical. Also, the beltline welds 1248/4M2F and 86054B/4E5F are the same fox the two plants. The RT+z results for the surveillance weld and weld 1248/4M2F were exactly the same for the two plants, and the RT+zs for weld 86054B/4E5F differed by only 8'F using the [MA85] procedure. Thus, the agreement for the two independent calculations was quite good. The weld RT~zs are also consistent with the generic RT~z value of -56'F reported in

[BY90].

In addition to firmly establishing the RT~z for all of the beltline materials, a further benefit of this analysis has been the determination of the surveillance weld T>Q and USE. Up to the resent, the weld surveillance data has been of little value since

0 29 the 4T>0 and 4USE cannot be calculated. However, the calculations reported herein have determined the unirradiated Tzo to be -32'F and the USE to be 128 ft-lb.

In the future, consideration should be given to obtaining a cooperative agreement with GPU Nuclear to share surveillances data for NMP1 and Oyster Creek. These two units are of the same design and vintage and have several identical materials. The two data bases can be combined to yield accurate plant-specific trend curves. Also, it would be desirable in the future to anneal the surveillance weld material and directly measure the Charpy properties of the material. These data could be used to verify the calculations and possibly reduce the uncertainty in the models.

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5.0 REFERENCES

[BY90] Byrne, S.T., "Niagara Mohawk Power Corporation Nine Mile Point Unit 1 Reactor Vessel Weld Materials", Report No.

86390 MCC 001 6/13/90 g ~

[MA87] Manahan, M.P., BCD-763-87-1, "Surveillance Capsules C'or Nine Mile Point Unit 1", Final Report from A'nd Battelle to Niagara Mohawk, September 30, 1987.

[MA85] Manahan, M.P., "Procedure for the Determination of Initial RT~~ in Cases where Limited Baseline Data are Available", 11/85

[MA85a] Manahan, M.P., Quayle, S.F., Rosenfield, A.R., and Shetty, D.K., <<Statistical Analysis of Cleavage-Fracture Data", Invited paper, Conference Proceedin s of the International Conference and Exhibition on Fati ue, Corrosion Cracking, Fracture Mechanics, and Failure Analysis, Salt Lake City (December 2-6, 1985).

[MA89] Manahan, M.P., "A Statistically Based Procedure for Determination of RT~~ When Limited Materials Data are Available", in preparation for The Journal of Nuclear Technology.

[MA90] Manahan, M.P., NMEL-90001, "Nine Mile Point. Unit 1 Surveillance Capsule Program", Final Report from Penn State/Battelle to Niagara Mohawk, December, 1990.

[MA90a] Manahan, M.P., "RT~~ and cr, Analysis for the Oyster Creek Nuclear Generating Station Pressure Vessel Beltline Materials", Final Report to GPU Nuclear Corporation, April 11, 1990.

[McFRAC) Manahan, M.P., et.al., "Statistical Analysis Methodology for Mechanics of Fracture", Final Report to Battelle's Corporate Technology Development Office, 1984

[OD86] Odette, G.R., Lombrazo, P.M., "The Relation Between Irradiation -Hardening and Embrittlement of Pressure Vessel Steels", Proceedings of the 12th ASTM Symposium on the Effects of Irradiation on Materials, 1986

[RG1.99] Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials", May, 1988 L-32