ML18038A518

Response to GL 92-01 for Nine Mile Point Unit 1

ML18038A518
Person / Time
Site:	Nine Mile Point
Issue date:	06/12/1992
From:	Michelle Manahan, Soong Y MPM RESEARCH & CONSULTING, NIAGARA MOHAWK POWER CORP.
To:
Shared Package
ML17058A976	List: ML17058A975 ML18038A518
References
GL-92-01, GL-92-1, NUDOCS 9207100009
Download: ML18038A518 (146)

Text

NMPC Project 03-9425 MPM-GL-692713 Response to NRC Generic Letter 92-01 for Nine Mile Point Unit 1

~ 4 Dr. M. P. Manahan, Sr.

MPM Research

& Consulting 213 Teaberry Circle State College, PA 16803-1269 Mr. Y. Soong Niagara Mohawk Power Corp.

301 Plainfield Road

Syracuse, NY 13212 June 12, 1992 9207100009 920702 PDR ADOCN 05000220 P

PDR

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Table of Contents 1.0 Introduction 2.0 Information Regarding Appendix H to 10 CFR Part 50 3.0 Information Regarding Appendix G to 10 CFR Part 50 4.0 Information Regarding Generic Letter 88-11 5.0 References Appendices Appendix A Surveillance Capsule Specimen Types and Materials Appendix B NMP-1 Beltline Material Chemistry Data Appendix B.1 1964 Lukens Ladle Analysis Appendix B.2 CE Determined Beltline Weld Chemistry Appendix B.3 Battelle and Westinghouse Measured Chemistry Data 12 33 49 51 52 61 62 64 66

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1.0 Introduction The U.S. Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 92-01 on March 6, 1992. This generic letter requests information needed to assess compliance with requirements and commitments made regarding reactor vessel integrity.

In particular, the generic letter requests information pertaining to the following reactor vessel structural integrity assurance requirements:

10 CFR 50.60 10 CFR 50.61 10 CFR 50 Appendix G 10 CFR 50 Appendix H Regulatory Guide 1.99 (Rev. 2), GL 88-11 This report contains data and information which document NMPC's past and on-going analyses and evaluations of NMP-1 in light of these regulations and requirements.

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2.0 Information Re ardin A

endix H to 10 CFR Part 50 Certain addressees are requested to provide the following information regarding Appendix H to 10 CFR Part 50:

Addressees who do not have a surveillance program meeting ASTM E 185-73, -79, or -82 and who do not have an integrated surveillance program approved by the NRC (see Enclosure 2), are requested to describe actions taken or to be taken to ensure compliance with Appendix H to 10 CFR Part 50.

Addressees who plan to revise the surveillance program to meet Appendix H to 10 CFR Part 50 are requested to indicate when the revised program willbe submitted to the NRC staff for review. lfthe surveillance program is not to be revised to meet Appendix H to 10 CFR Part 50, addressees are requested to indicate when they plan to request an exemption from Appendix H to 10 CFR Part 50 under 10 CFR 50.60(b).

The reactor vessel surveillance program at NMP-1 is in compliance with ASTM E1 85-66.

(1S66).

The description of the response to the requirements of ASTM E185-66 is provided below.

The NMP-1 surveillance program satisfies all of the current 10CFR50 Appendix H requirements with two exceptions:

the initial RTNpT was not determined in accordance with the current ASME Section III rules for all beltline materials; and the surveillance specimen orientation is longitudinal-transverse (I=T).

As described

later, NMPC has taken steps to address properly these requirements.

Res onse to ASTM E185-66 Re uirements:

(1)

ASTM E185-66 Section 3.1 Test Material As shown in Table 1-1, the NMP-1 surveillance program consists of three capsules, and each capsule contains a minimum of eight (8) specimens for each type (base, weld, and HAZ).

All specimens were prepared from the actual materials used to fabricate the beltline region of the reactor vessel.

Plates G-8-1 and G-8-3/G-8-4 (see Table 2-8),

which represent more than one heat of base metal, are both being irradiated in the NMP-1 surveillance program.

The surveillance welds were made with the same type of weld wire and with the same welding procedures used in fabrication of the beltline region.

All surveillance test materials received heat treatment and stress relief so that they represent the metallurgical conditions of the beltline region plates of the reactor vessel after fabrication.

Also, for each beltline plate and weld, chemical composition measurements were made for Cu, Ni, P, Mn, Mo, C, S, and Si. Thus, all of the requirements of ASTM E185-66 Paragraph 3.1 were met.

(2)

ASTM E185-66 Section 3.2 T e of S ecimen Both tension and impact specimens have been included in the capsules.

ASTM E23 Type A conventional Charpy specimens have been used to measure beltline properties.

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Tension specimens were prepared in compliance with ASTM E8.

(3)

ASTM E185-66 Section 3.3 Number of S ecimens As shown in Table 1-1, the number of capsule specimens complies with the minimum number required to establish the Charpy transition temperature curve.

(4)

ASTM E185-66 Section 3.4 Correlation Monitors Correlation monitors are not required and, based on current understanding, were not included in the NMP-1 surveillance program. As shown in Table 1-1, Capsule B contains specimens marked "APED" for which documentation has not yet been found.

(5)

ASTM E185-66 Section 4.1 Location of S ecimens The azimuthal location of the NMP-1 surveillance capsules are shown in Table 1-1. The lead factor at the 1/4T was calculated to be 0.99 for the originally installed capsules.

The maximum 1/4T fast neutron fluence after 25 EFPY of operation was calculated to be 1.44 x 10" n/cm'.

As part of the reinsertion program, the capsule A & C (currently called A' C') locations have been modified to be closer to the core than the original locations to increase the neutron flux and thus to increase the lead factor to 1.97 (see Table 1-2).

Each surveillance capsule contains two Charpy specimen packets.

Inside of each packet there are flux monitor wires; one each of iron (Fe), copper (Cu), and nickel (Ni).

(6)

ASTM E185-66 Section 4.2 Accelerated or Reduced Irradiation The use of accelerated capsules is not a requirement and the NMP-1 surveillance program does not include accelerated capsules.

(7)

ASTM E185-66 Section 4.3 Thermal Control S ecimens The initial capsules did not contain thermal monitors.

However, the reinsertion capsules contain two different types of temperature monitors (TMs):

conventional melt wires (MWs), which are the primary source of temperature monitoring; and solid state track recorder-temperature monitor (SSTR-TMs), which are used as the backup to the melt wires.

The advantage of the SSTR-TM monitors is that they are potentially capable of providing an estimation of the average temperature in the capsule.

(8)

ASTM E185-66 Section 4.4 Test Ca sules The test specimens are placed in a leak tight stainless steel container.

Prior to final weld closure, the specimen packets are evacuated and backfilled with an inert gas.

The specimen packet and basket has been designed to allow good heat transfer with the downcomer water while maintaining a corrosion resistant environment.

(9)

ASTM E185-66 Section 4.5 Corrosion-Resistant Material Com onents See reply in (8) above.

(10)

ASTM E185-66 Section 4.6 S ecimen Withdrawal As shown in Table 1-1, two capsules have been removed to date, Capsule A and

Capsule C. Both of these capsules were tested and analyzed in accordance with ASTM E185-82.

The NMP-1 capsule withdrawal schedules are shown below:

Original Program -

First Capsule (A) - 5.8 EFPY Second Capsule (C) - 7.98 EFPY Third Capsule (B) - 16 EFPY (expected in 1996)

Current Program -

Fourth Capsule (A') - 24 EFPY Fifth Capsule (C') - 32 EFPY Sixth Capsule (B') - 40 EFPY (planned)

A substantial amount of additional work has been undertaken to update the NMP-1 surveillance program and to correct deficiencies in the original program.

This work is described in detail below.

Historical Back round:

Three surveillance capsules were installed in the NMP-1 reactor vessel in 1969 prior to initial operation. The number and the types of specimens installed in the original program, as well as the capsule identification and location within the reactor

vessel, are summarized in Table 1-1. Capsule C was removed after 7.98 EFPY and capsule A was removed after 5.8 EFPY. The Charpy data for the beltline plate specimens (originally thought to be G-8-3) had demonstrated an unexpected large shift in the 30 ft-Ib index temperature of 114 F (over four standard deviations higher than expected using the R.G.

1.99, Rev. 2 model).

After extensive study, archive material testing, and careful examination of all data available, it has been conclusively demonstrated that a material mix-up has occured in the NMP-1 surveillance program during fabrication ofthe surveillance capsule specimens.

The base Charpy specimens were fabricated from plate G-8-1, and not plate G-8-3 as originally specified. However, the base metal portion of Charpy weld and HAZspecimens is composed of plate G-8-3 material, As a result, the plate G-8-3 measured shift (AT3Q),

originally thought to be 114'F, was correctly established to be 11'F at a fast fluence of 4.78x10" n/cm'. Since the surveillance program is irradiating two plate materials (G-8-3 and G-8-1), the Charpy dT>> can be determined for plate G-8-1 as well. For plate G-8-1, BT~Q 79'F at a fluence of 4.78x10'" n/cm',

and bT,Q 55'F at fluence of 3.6x10" n/cm'.

Therefore, the resolution of the material mix-up has resulted in three measured Charpy bT,Q instead of two, and the measured shifts are much lower than believed earlier. At the request ofthe NRC, NMPC submitted on August 30,1991 a comprehensive report [MA91]

describing the material mix-up issue and its resolution.

Status of Current Surveillance Pro ram:

Since plant life extension is being considered, NMPC decided to re-insert two capsules (A'nd C'). The prime is used to designate the new capsules in the same azimuthal location as the original capsules. The radial location of the reinsertion capsules relative

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to the reactor vessel wall was modified to increase the fast neutron flux at the capsule location by a factor of 1.97 over the maximum flux at the reactor vessel peak flux 1/4 T location. The mechanical behavior specimens which were included in the new capsules are summarized in Table 1-2. The re-insertion, capsules are a second generation BWR design and include spectral radiometric monitors (RMs), helium accumulation fluence monitors (HAFMs), solid state track recorders (SSTRs), melt wires, and solid state track recorder-temperature monitors (SSTR-TMs),

reconstituted and conventional Charpy specimens, miniature and conventional tensile specimens, and archive plate material

~

The initial RTN>> for Plate G-8-3 was measured in accordance with the current ASME Section III requirements after the surveillance archive plate was discovered.

The initial RT>> for the remaining beltline materials was calculated in accordance with the intent of Section III of the ASME code and these calculations were reported in Reference [MA91).

Allsurveillance Charpy specimens, including the reinsertion capsule specimens, have an L-Torientation. An L-Tto T-Ladjustment factor has been applied to the surveillance data to meet the current Appendix H requirements.

Summa of Im ortant Data:

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Reactor Vessel Fabrication Code:

ASME Section 1, 1962 Edition and Addenda through 1963 plus the nuclear code cases applicable on December 1963.

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Surveillance Program: ASTM E-185:

1966

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Original Surveillance Program:

See Table 1-1

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Current (after re-insertion) Surveillance Program: See Table 1-2

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Configuration & Identification of Beltline Region: See Figure 1-1

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All capsule specimens are L-T orientation

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Measured Transition Temperature Shift (30 ft-Ib) in L-T orientation:

(1) 79.3'F after 7.98 EFPY (G-8-1 base in capsule C)

Equivalent shift of 85'F for G-307-4 plate (2) 55.1'F after 5.8 EFPY (G-8-1 base in Capsule A)

Equivalent shift of 60'F for G-307-4 plate (3) 11.2'F after 7.98 EFPY (G-8-3 base in Capsule C)

Equivalent shift of 23.TF for G-307-4 plate

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Limiting Material & RTN>>'late G-307-4, 40'F

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Minimum Upper Shelf Energy (USE) in L-T orientation:

78 ft-Ib for G-8-1 (after 7.98 EFPY) 114 ft-Ib for G-8-3 (after 7.98 EFPY, only one data point) 80 ft-Ib for G-307-4 (initial value) 78.5 ft-Ib for G-8-3 (initial value) 82 ft-Ib for G-8-1 (initial value)

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Minimum Upper Shelf Energy (USE) in T-L orientation:

64 ft-Ib for G-8-3 (initial value)

Accumulated Capsule Fluence (E>1.0 Mev) after 7.98 EFPY: 4.78x10"n/cm' Current P-T Curves: Comply with RG 1.99 (Rev. 2)

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Discovery and resolution of the material mixup: resulted in 3 valid surveillance

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data.

See Figure 3-4 Capsule Withdrawal Schedule:

Original program:

First capsule (A)- 5.8 EFPY Second capsule (C)- 7.98 EFPY Third capsule (B) -16 EFPY (Expected in 1996)

Current program:

Fourth capsule (A')-24 EFPY Fifth capsule (C')-32 EFPY Sixth capsule (B')-40 EFPY (planned)

Based on the NMP-1 past plant capacity factor, and projecting to the end-of-license(EOL) date of 2009 using a 70% capacity factor, the total exposure at EOL is not expected to exceed 25 EFPY. This exposure level is used in all analyses to predict the EOL vessel damage.

A endix H Com liance:

In summary, the NMP-1 surveillance program was originally designed to and meets the requirements of ASTM E185-66.

The surveillance specimens removed to date were tested in accordance with ASTM E1 85-82. The initial RT>> of the beltline materials has been calculated in accordance with ASME Section III~ An L-T to T-L adjustment factor has been applied to the surveillance data to meet'the current 10 CFR 50, Appendix H requirements.

Therefore, the current surveillance program for NMP-1 meets the requirements of Appendix H to 10 CFR 50.

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TABLE 1-1 SURVEILLANCECAPSULE MECHANICALBEHAVIOR 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 A(18) 30 1979 5.8 12 Base 12 Weld 12 HAZ 3 Base 2 Weld 3 HAZ 120 Not Removed (In Vessel) 16(expected 10 Base in 1996) 8 Weld 9 HAZ 9 APED 3 Base 3 Weld 2 HAZ 2 APED(')

C(1,4) 300 1982 7.98 8 Base 8 Weld 8 HAZ 2 Base 2 Weld 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.

10 TABLE 1-2 CURRENT SURVEILLANCECAPSULES AFTER REINSERTION FOR NINE MILE POINT - UNIT 1

~Ca sate Ca sule Contents"'Char Tensile Lead Factor Capsule Exposure to 1/4 T at Withdrawal Position EFPY A'

Base-0 6 Base-R 12 Weld - 0 12HAZ -0 2 Base - 0 1.97 2 Base-M 2 Weld-0 3HAZ -0 24 10 Base 8 Weld 9 HAZ 9 APED 3 Base 2 Weld 3 HAZ 2 APED 0.99 16 (40 atter withdrawal and re-insertion)

C'2 Base-R 12 Weld-R 12 Base-U 4 Base - M 1.97 4 Weld-M 3 Base-U 32 0 = Original irradiated (untested) specimen M = Miniature tensile specimens from broken ends of tested specimen R = Reconstituted Charpy specimen U = Unirradiated specimen

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12 3.0 Information Re ardin A

endix 6 to 10 CFR Part 50 2.

Certain addressees are requested to provide the following information regarding Appendix G to 10 CFR Part 50:

Addressees of plants for which the Charpy upper shelf energy is predicted to be less than 50 foot-pounds at the end of their licenses using the guidance in Paragraphs C.1.2 or C.2.2 in Regulatory Guide 1.99, Revision 2, are requested to provide to the NRC the Charpy upper shelf energy predicted for December 16, 1991, and for the end of their current license for the limiting beltline weld and the plate or forging and are requested to describe the actions taken pursuant to Paragraphs IKA.1 or V.C of Appendix G to 10 CFR Part 50.

The estimated upper shelf energy (USE) for the NMP-1 beltline plates and welds, as of December 16, 1991, and at EOL, are given in Table 2-1. These estimates were calculated using the guidance in paragraph C.1.2 of Regulatory Guide 1.99 (Rev. 2).

Since only L-T orientation data are available for most of the beltline materials, the reference [BTP81] guidance was used to convert from the L-Tto T-L orientation.

In particular, the L-T values were multiplied by 0.65 to obtain the T-L orientation estimates.

Based on these conservative models, plates 6-307-4 and 6-8-1 are predicted to be below the 50 ft-Ib screening criterion at the present time.

Plate 6-307-4 is also the critical plate material from an ARTQ7 perspective.

The unirradiated USE data listed in Table 2-1 were compared with the data reported in Reference

[GE92].

Since the Reference

[GE92] table which summarizes the NMP-1 data does not contain up-to-date information, the comparisons were made with the Reference [GE92] Figures 3-6 and 3-7.

Figure 3-6 of [GE92] presents the modified A302B lower bound USE fit vs. the Charpy energy at 10'F. Similarly, Figure 3-7 of [GE92] presents the lower bound USE vs.

the Charpy energy at 10'F.

The [GE92] model predicts the A302B plate USE to be 90 ft-lb.

The measured plate G-307-4 USE is significantly lower than the

[GE92] model prediction.

However, it is important to note that the NMP-1 USE data was not used in the [GE92] regression analysis.

For welds, the [GE92] model predicts the USE to be approximately 80 ft-lb. This is roughly 10 ft-Ibs lower than that calculated in Table 2-1.

Nevertheless, even if the conservative [GE92] weld estimates were used, the USE of the beltline welds would not be expected to fall below 50 ft-Ib at EOL.

Examination of the irradiated upper shelf data presented in Reference

[MA91]

suggests that the shelf drop is negligible. However, this conclusion is tentative for plate 6-8-3 since there are not sufficient USE data available for statistical analysis.

Capsule B is scheduled for withdrawal during the 1996 outage.

This capsule can

13 provide the data needed for verification of a small upper shelf energy decrease for both the 6-8-1 and G-8-3 materials.

In the case of plate 6-8-1, there are three irradiated and three unirradiated USE points available for analysis.

These data are summarized in Table 2-2.

Comparison of the linear averages suggests that the AUSE is so small that it is within the measurement uncertainty.

If the BUSE is conservatively calculated using the mean of the unirradiated data and the lowest irradiated data point, the d USE is 10%. Similarly, ifthe AUSE is calculated using the lowest irradiated and unirradiated points, the dUSE is 5%.

These data are compared with the Regulatory Guide 1.99 (Rev. 2) model in Figure 2-1. The G-8-1 Cu content (0.23 Wt. %) is close to the 6-307-4 Cu content (0.27 Wt. %). Therefore, a chemistry correction was not applied.

The Reference [BTP81] L-T to T-L conversion factor of 0.65 appears to be overly conservative for the NMP-1 beltline plates.

In particular, the measured L-T to T-L conversion is 0.82 [MA91]. Applying these material-specific factors, the best estimate USE data for plates G-8-1 and 6-307-4 are given in Table 2-3. The b,USE estimates in Table 2-3 were obtained using the guidance of paragraph 2.2 of Regulatory Guide 1.99 (Rev. 2) with an L-T to T-L conversion factor of.8 and an assumed B,USE of 10% at 7.98 EFPY. The L-T to T-L conversion factor of 0.8 was obtained using the plate G-8-3 lowest measured USE data measured in the L-T and T-L orientations.

Based on this analysis, it is predicted that the critical plate USE will not fall below 50 ft-Ib prior to EOL.

It is recognized that additional data and analyses will be needed to confirm this estimation.

The on-going NMPC work to develop material-specific models is described below.

During review of the NMP-1 reactor vessel pressure-temperature limits technical specification amendment, the NRC requested additional information regarding NMPC's plans for addressing the low USE issue.

NMPC submitted the response to this request on September 27, 1991, [SY91].

This letter summarized the technical program, which is currently underway, to resolve the USE issue.

This technical approach is briefly summarized below:

An elastic-plastic fracture mechanics analysis based on the J-integral tearing modulus methodology is being performed.

The results of this analysis will be a calculation of the lowest USE allowed as a function of cooling rate for both normal operation and accident conditions for the limitingbeltline materials. The calculative procedures which willbe followed in the elastic-plastic fracture mechanics analysis are described in the draft Appendix X to the ASME Code.

The 65% L-T to T-L conversion rule proposed by the NRC in Reference

[BTP81] will be examined on a materials-specific basis to develop a physically based model for accurate conversion of L-T to T-L data.

The

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recently released NRC light water reactor surveillance data base will be used in the study.

Niagara Mohawk is partially supporting, through an Empire State Electric Energy Research Corporation (ESEERCO) research

contract, work at Columbia University to develop a physically based USE decrement model for prediction of the shelf drop as a function of fluence. A physically based ductile fracture model will be used, in conjunction with the NRC light water reactor surveillance data base, to develop a new correlation for predicting USE drop.

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Niagara Mohawk is partially funding Project FERMI at Penn State University.

One of the FERMI projects is focused on the USE issue.

In particular, microstructural data on NMP-1 plate 6-8-3 (surveillance plate) are being measured in both the irradiated and unirradiated conditions to provide evidence for the physically based ductile fracture model theory for predicting USE drop.

A related ESEERCO project which is currently underway, willprovide support for microstructural examination of two plates and two welds to provide further evidence for the physically based ductile fracture model theory for predicting USE drop.

The above described technical approaches will result in:

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A physically based, material-specific, USE decrement model.

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An accurate, material-specific, L-T to T-L USE data conversion model.

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Calculation of the lowest USE allowable, based on elastic-plastic fracture mechanics, such that the margins of safety for fracture are equivalent to those specified in 10 CFR 50, Appendix G.

These results will be synthesized to define the exposure level, in effective full power years, which can be achieved with adequate protection against ductile fracture.

In summary, NMPC believes that the models used to calculate the Table 2-1 data are overly conservative for the NMP-1 beltline materials.

Microstructural data obtained to date indicates a large population of MnS inclusions, MO,C precipitates, and Fe,C precipitates in the unirradiated plate [FR92].

These precipitates and inclusions have also been shown to be stable under irradiation.

It has been proposed [MA91b]that the lowering of the upper shelf due to neutron damage in steels with initially high concentrations of particles is expected to be negligible since the irradiation induced defects (Cu rich precipitates, microvoids) will not significantly influence the fracture process on the upper shelf.

As discussed earlier, the Reference

[MA91] data support this proposition.

Accordingly, it is inappropriate to apply generic correlations, developed using data for relatively clean steels (A5338), to predict the b,USE for the NMP-1 plate materials.

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15 Therefore, NMPC is developing material-specific models, which accurately model the physics of ductile fracture, which will yield accurate and conservative predictions of the effects of neutron damage on ductile fracture properties.

Based on the current work schedule, NMPC is planning to submit a report which documents resolution of the low USE issue by June, 1994.

In the unlikely event that the material-specific models predict USE levels below 50 ft-Ibs prior to EOL, NMPC will perform measurements, inspections, and analyses as described in paragraph V.C. of Appendix G to 10 CFR 50.

Certain addressees are requested to provide. the following information regarding Appendix G to 10 CFR Part 50:

b.

Addressees whose reactor vessels were constructed to an ASMECode earlier than the Summer 1972 Addenda of the 1971 Edition are requested to describe the consideration given to the followingmaterial properties in their evaluations performed pursuant to 10 CFR 50.61 and Paragraph III.Aof 10 CFR Part 50, Appendix G:

(1) the results from all Charpy and drop weight tests for all unirradiated beltline materials, the unirradiated reference temperature for each beltline material, and the method of determining the unirradiated reference temperature from the Charpy and drop weight test; The original baseline vessel material properties consisted of L-Torientation Charpy data for each beltline plate and only three Charpy measurements at +10'F for the beltline welds.

These data are summarized in Reference

[MA91]. After discovery of the G-8-3 archive plate material, the RTDT was determined in strict conformance with the current ASME Section III requirements (T-L Charpy data and drop weight tests).

During this work, it was observed that the L-Tto T-L orientation change results in an upward shift of the Charpy curve of 24'F. The archive plate G-8-3 drop weight and Charpy data are summarized in Table 2-4.

Direct measurement of the initial RT>>, for the remaining beltline materials is not possible because archive plate is not available for these materials.

However, in September, 1990, the Reference

[MA90] statistical method was applied to all of the NMP-1 beltline materials to determine the RT>> in conformance with the current version of Section IIIto the ASME Code. These data are presented in Reference [MA91]and summarized in Table 2-5. A detailed description of the technical approach is provided in [MA91],and the key features of the approach are described below.

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16 Weld Material RT>>~Anal sis Many operating power plants were builtwhen current regulations concerning establishment of the initial RTpT were not in force.

In many cases, insufficient drop weight and Charpy data are available to determine the RTNpT in accordance with the current Section III of the ASME code.

Prior to 1972, the ASME code required 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.

The NRC Branch Technical Position, MTEB 5-2, contains a procedure for estimating the RT>>, based on generic data. However, this approach 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 [GE85] which is based on a temperature corresponding to the 50 ft-Ib longitudinal energy level (T,OL) which is estimated by adding 2'F per ft-Ib to the testing temperature, using the lowest Charpy energy value as the starting value. The GE model results in a transition region slope of 0.5 ft-Ib/'F. However, this model is also generic.

Neutron irradiation of pressure vessel materials causes:

1) 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, is about 200'F 245'F for welds. These data, along with Odette's yield strength model, can be used to provide an accurate estimate of the RTNpT 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'F. 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 RTND7 In 15 out of 23

cases, the RTNpT was determined from the Charpy data.

The average difference between the actual RT>>T and the RT>> estimated using the Reference

[MA85] approach 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>>,. The RTNpT was determined from Charpy data in 13 of the 17 cases.

The average between

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the ASME code determined RTN>> and that determined using Reference

[MA85]was 8', and the average net deviation was 4.4'F forthe base metal

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Given these facts and observations, the approach to RT>> determination is to determine a temperature, T,>> (for T-L orientation), at which three Charpy specimens would be expected to yield greater'than 50 ft-Ibs of absorbed energy.

The T<<T is determined by analyzing the uncertainty in the Charpy data in the transition region.

The T,>> temperature is defined as the temperature at which the mean Charpy curve minus 50 ft-Ibs equals 2a,.

The 2c, term is defined as the statistical 95% confidence band for energy assessment at the 50 ft-Ib level.

Once the T,>> temperature is determined, the RTN>> is taken to be T<<T 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 RTN>> for the NMP-1 welds is as follows:

Ste 1

~Ste 2

~Ste 3

Ste 4

Ste 5

Using Odette's yield strength model [OD86], calculate the unirradiated upper shelf energy (USE') ~

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 200'F.

Verify the reasonableness of the slope by comparison with the irradiated transition region slope.

Estimate the 95% confidence band for energy at the 50 ft-lb.

level.

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 RT>>.

(Optional)

If sufficient data

exists, use the Odette yield strength model to calculate the transition temperature shift at the 30 ft-lb. level (bT3Q) and compare with the hT>> obtained above and with the RG1.99(2) model prediction.

For weld W5214/5G13F, the unirradiated USE was calculated to be: USE' 128 ft-lb. 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 worked 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-Ib/'F which is in good agreement with the irradiated data.

The calculated slope of the unirradiated transition region is 0.645 ft-Ib/'F, which is slightly larger than the irradiated curve, as expected.

l

The third step of the RT>>, analysis calls for an assessment of the statistical 95% confidence band (2a,) for energy measurement at the 50 ft-lb. level.

Two separate calculations were performed. The first approach uses the 2a variation in energy at the 50 ft-lb. level. This results in: 2a,(fit)'

13.5 ft-lb. The second approach consists of an assessment of the 95% confidence in the experimental data.

This approach yields: 2o,(experimental)'

12.0 ft-lb.

The procedure defines 2o, as the larger of 2a, (fit) and

2cr, (experimental), with the lower bound being 10 ft-lb. Therefore, for weld W5214/5613F:

2o'E = 10 ft-lb.

Since the three data points for weld W5214/5613F were over 50 ft-lb., the RTN>> is taken to be:

RT(weld W521 4/5613F) = 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 RT>> using this approach would be -40'F, which is in close agreement and slightly conservative when compared with the measured Charpy data.

The RT>>, determination for the beltline weld 86054B/4E5F was made as described forweld W5214/5613F because the data at+10'F are well above 50 ft It),

However the calculated RTNpT 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 1248/4K1 3F were above 50 ft-lb. at 10'F, and therefore, the RTN>> was determined experimentally for these materials as well.

Plate Material RT>>~Anal eie In the case of the base metal plates, full unirradiated Charpy curves are available for each beltline plate.

The surveillance capsule base metal specimens were fabricated from plate 6-8-1.

The RT>>, was determined in a similar manner as described above for the weld materials, 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 [McFRAC]. Then, the 95% confidence interval in energy

'2a,(fit) is an estimate, based on the least squares regression

analysis, of the goodness of the fit to the data.

26~

(experimental) is an estimate of the scatter in the experimental data

19 (2a,) was determined using the more conservative value of either the fit uncertainty at the 50 ft-Ib level (2a,(fit)) or the deviation of the data from the mean in the transition region (2a, (experimental)).

If both of these measures are less than 10 ft-lbs, then 10 ft-lb. was taken as the 2a, value.

Otherwise, the larger of 2a,(fit) or 2a,(experimental) was used.

The 10 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-Torientation), at which three Charpy specimens would exhibit 50 ft-Ibs of absorbed energy was determined.

This temperature was taken to be the temperature at which the mean Charpy curve minus 50 ft-lbs. equals 2a,.

Once, the T<<

temperature is determined, the RT>> (L-T) is taken to be T<<minus 60'F.

The Charpy specimens tested forthe plate materials had an L-Torientation.

The ASME code requires testing using the T-L orientation because it is the limiting orientation. General Electric recommends adding 30'F to the RTN>>

obtained 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-Tto 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 for NMP-1 was 24'F.

Certain addressees are requested to provide the following information regarding Appendix 8 to 10 CFR Part 50:

Addressees whose reactor vessels were constructed to an ASMECode earlier than the Summer 1972 Addenda of the 1971 Edition are requested to describe the consideration given to the followingmaterial properties in their evaluations performed pursuant to 10 CFR 50.61 and Paragraph III.Aof 10 CFR Part 50, Appendix 6:

(2) the heat treatment received by all beltline and surveillance materials; 1.

Heat treatment of beltline materials (before welding)

The shell plate was heat treated at 1550-1600'F, held at temperature four (4) hours and then dip quenched in agitated water at 1225 2 25'F for four (4) hours.

A test sample was then removed and given an additional heat treatment of 1150 2 25'F, held at temperature forthirty (30) hours, and then furnace cooled to 600'F.

2.

Heat treatment after welding 20 For the beltline region, two welding procedures (SAA-33-A(3), MA-33-A(7))

were used.

Post-weld heat treatment 1150'F + 25'F. Typically, the indicated heat treatment during fabrication was augmented in most cases by subassembly and assembly heat treatments that could have resulted in cumulative stress relief anneals of up to 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> at 1150'F.

3.

Heat treatment of reactor vessel The reactor vessel was heat treated for 10 1/2 hours at 1150 2 25'F, with heating and cooling at a stow rate.

4.

Heat treatment of surveillance test materials (base, weld, and heat affected zone)

A simulated stress relief for thirty (30) hours at 1150'F 2 25'F was conducted to ensure that the test specimens represent the metallurgical conditions of the lower shell plates of the reactor vessel after fabrication.

Certain addressees are requested to provide the following information regarding Appendix G to 10 CFR Part 50:

b.

Addressees whose reactor vessels were constructed to an ASMECode earlier than the Summer 1972 Addenda of the 1971 Edition are requested to describe the consideration given to the followingmaterial properties in their evaluations performed pursuant to 10 CFR 50.61 and Paragraph III.Aof 10 CFR Part 50, Appendix G:

(3) the heat number for each beltline plate or forging and the heat number of wire and flux lot number used to fabricate each beltline weld; The beltline consists of the lower shell and lower-intermediate shell.

Each shell was formed using three plates, so the beltline includes a total of six plates, six vertical welds, and one circumferential weld.

(See Figure 1-1).

The heat and flux lot numbers are given in Tables 2-6 and 2-7.

Certain addressees are requested to provide the following information regarding Appendix G to 10 CFR Part 50:

b.

Addressees whose reactor vessels were constructed to an ASMECode earlier than the Summer 1972 Addenda of the 1971 Edition are requested to describe the consideration given to the followingmaterial

5

21 properties in their evaluations performed pursuant to 10 CFR 50.61 and Paragraph III.Aof 10 CFR Part 50, Appendix G:

(4) the heat number for each surveillance plate or forging and the heat number of wire and flux lot number used to fabricate the surveillance weld; Based on a careful examination of all of the NMP-1 surveillance data, it has been concluded that a material mix-up did occur in the NMP-1 surveillance program. This conclusion is based on definitive chemical analysis data and has been further substantiated by examining the mechanical behavior trends. The base metal Charpy specimens were prepared from plate 6-8-1 material and the HAZ and weld specimens were prepared using 6-8-3 material. The HAZand weld tensile specimens were produced using 6-8-3 material.

However, the base metal tensile specimens were prepared from G-8-1 material.

The resolution of the material mixup is summarized in Table 2-8.

The capsule specimen material compositions are listed in Appendix A.

As with many plants in operation today, NMP-1 surveillance material is not the limiting beltline material from an embrittlement perspective.

At NMP-1, the peak flux occurs above the core mid-plane in the region of the 6-307 plates (see Figure 1-1) ~

Based on the chemistry data and analysis of the initial RTN>> forthe base materials, plate 6-307-4 willexperience the largest Charpy shift (bT>>), since its copper and nickel content is higher than any of the other beltline plates.

Therefore, a correction which adjusts for the chemistry difference between 6-307-4 and the surveillance materials (G 3, G-8-1) has been developed.

This approach allows NMP-1 to use the surveillance materials to directly monitor the embrittlement of the limiting plate G-307-4 materials.

Further details are provided in [MA91] and in information response 3.b.

Certain addressees are requested to provide the following information regarding Appendix G to 10 CFR Part 50:

Addressees whose reactor vessels were constructed to an ASMECode earlier than the Summer 1972 Addenda of the 1971 Edition are requested to describe the consideration given to the followingmaterial properties in their evaluations performed pursuant to 10 CFR 50.61 and Paragraph III.Aof 10 CFR Part 50, Appendix G:

(5) the chemical composition, in particular the weight in percent of copper, nickel, phosphorous, and sulfur for each beltline and surveillance material;

l l

22 The chemical composition data for the beltline materials can be organized into three classifications:

~

The 1964 Lukens measured (atomic absorption) chemistry for all six plates in the beltline region. These data are shown in Appendix B.1.

~

The CE determined RPV beltline weld chemistry including the surveillance capsule weld. These data are given in Appendix B.2.

The Battelle and Westinghouse measured (ICAP, AA, Leco Combusion, Titrimetric Combustion, and gravimetric) data.

These data are shown in Appendix B.3.

Earlier EDAX data on plate material was discarded because of large scatter due to the material heterogeneity.

The chemistry data was linearly averaged to provide the best estimate plate and weld composition.

These data are summarized in Table 2-9.

Certain addressees are requested to provide the following information regarding Appendix G to 10 CFR Part 50:

b.

Addressees whose reactor vessels were constructed to an ASMECode earlier than the Summer 1972 Addenda of the 1971 Edition are requested to describe the consideration given to the followingmaterial properties in their evaluations performed pursuant to 10 CFR 50.61 and Paragraph III.Aof 10 CFR Part 50, Appendix G:

(6) the heat number ofthe wire used fordetermining the weld metal chemical composition ifdifferent than Item (3) above.

Not applicable to NMP-1.

.1 l

23 Table 2-1 Estimated Upper Shelf Energy for NMP-1 Beltline Materials Material Plates Wt. %

cu Minimum Unirrad.

USE (ft-Ib)

L-T

'inimum Unirrad.

USE (ft-Ib)

T-L'rradiation Decrement AUSE (%)

12/16/91 Irradiation Decrement AUSE (%)

EOL(25 efpy)'redicted USE (T-L)'2/16/91 (ft-Ib)

Predicted USE(T-L)'t EOL(25 efpy)'ft-Ib)

G-8-3/6-8-4 0.1 8 6-8-1 0.23 G-307-3 0.20 G-307-4 0.27 G-307-10 0.22 78 82 100 80 97 64 /50.7 53.3'5.0'2.063.1'5 17 16 20 17 17 20 19 23 20 54.4 44.2 54.6 41.6 52.4 53.1 42.6 52.7 40.0 50.5 Welds W5214/5G13F 0.1 8 86054B/4E5F 0.22 1248/4K1 3F 0.22 1248/4M2F 0.22 100 904 904 904 17 20 20 20 20 23 23 23 83.0 72.0 72.0 72.0 80.0 69.3 69.3 69.3

'he L-T and T-L designations apply to plate material only

'easured using archive plate in the T-L orientation

'rradiated value measured at a fluence of 4.78 x 10" n/cm' Conservatively estimated using data in [MA90] and [MA91]

'ast fluence of 7.26 x 10" n/cm't the peak 1/4T position

'ast fluence of 1.44 x 101e n/cm't the peak 1/4T position

'ata from Reference [CE90]

'urveillance Weld

'alculated by multiplying L-T data by 0.65

l

24 Table 2-2 USE Data for Plate G-8-1 Irradiated to a Fast Fluence of 4.78 x10" n/cm'nirradiated USE (ft-Ib)

Irradiated USE (ft-Ib) b,USE Measured data Measured data Measured data 82 83 95 78 99 104 Average of Measured Data 86.7 93.6 01 Shift based on Lowest Measured Data 82 78 Shift Conservatively Based on Mean Unirradiated and Lowest Irradiated Data 86.7 78 10

'hift is negligible and within experimental scatter

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25 Table 2-3 Best Estimate Upper Shelf Energy for Plates G-8-1 and G-3074 Material G-8-1 6-307-4 Minimum Unirrad.

USE (ft-Ib)

L-T 82 80 Minimum Unirrad.

USE (ft-Ib)

T-L'5.6 64.0 Irrad.

Decre-ment'USE(%)

12/1 6/91 Irrad.

Decre-ment'USE(%)

EOL (25efpy)'3 13 Predicted USE (T-L) 12/1 6/91 (ft-Ib) 58.4 56.9 Predicted USE (T-L) at EOL (25 efpy)'ft-Ib) 57.1 55.7

'late 6-8-3 measured L-T to T-L conversion of 0.8 applied

'ast fluence of 7.26 x 10" n/cm't the peak 1/4T position

'aragraph 2.2 of RG1.99 (Rev. 2) used.

hUSE conservatively calculated using average unirradiated data and lowest irradiated datum

'ast fluence of 1.44 x 10" n/cm't the peak 1/4T position

I l

l

26 Table 2-4 Summary of Drop Weight Charpy impact Properties for Unirradiated Base Metal Plate G-8-3/G-8-4 30 ft-Ib Transition Temperature Orientation Remarks

('F) 50 ft-Ib Transition Temperature

('F)

Upper Shelf Energy (ft-Ib)

L-T (1987)

L-T (1964)

T-L (1990)

Archive (1987)

C.E. Data (1964)

Archive (1990)

-21

-26.5

- 0.2 14.4 46.5 108 99.5 68.3 Orientation Plate Drop Wt.

ft-Ib NDT,'F Transverse Cha Data MLE Temp oF RTNov T-L(1990)

G-8-3

-25 60, 53, 57 52, 58, 51 57

l l

27 Table 2-5 Beltline Material RTNpT Data for Nine Mile Point Unit 1 Plate G-8-3/6-8-4 6-8-1 6-307-3 6-307-4 6-307-10 RTNoT (T-L) oF

-3 36 28 40 20 G(

oF Plate W5214/5613F2 86054 B/4E5F 1248/4K1 3F 1248/4M2F RTNDT (T L) oF

-50

-50

-50

-50 6(

oF 17 17 17 17

'easured in accordance with ASME Code

'urveillance Weld

.l f

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28 Table 2-6 NMP-1 Reactor Vessel Beltline Plates Heat No.

P2074 P2076 P2091 P2112 P2130 Plate ID G-307-3

'-307-4'-307-5 6-307-10 6-8-1 6-8-3 6-8-4

'imiting plate from a radiation damage perspective (highest ART+$7)

~

~

~

1 I,

(g l

29 Weld Seam TABLE 2-7 REACTOR VESSEL BELTLINEWELD INFORMATION Number Location Weld Wire Type and Heat No.

Weld Flux Type and Lot No.

Detailed Weld Procedure 2-564 A/C Lower-Intermediate Shell Longitudinal 2-564 D/F 3-564 Lower Shell Longitudinal Seams Lower Intermediate to Lower Shell Girth Surveillance All Three Capsules Capsule Weld RACO 3/86054 RACO 3/1 248 E8018/HAC0 E8018/JBGD RACO 3/86054 E8018/HAG D E8018/JBGD RACO 3/1248 E8018/DB DE E8018/IOG E RACO 3/W521 4 Arcos B-5/4E5F Arcos B-5/4K1 3F N/A N/A Arcos B-5/4E5F N/A N/A Arcos B-5/4M2F N/A N/A Arcos B-5/5G13F SAA-33-A(3)

SAA-33-A(3)

MA-33-A(7)

MA-33-A(7)

SAA-33-A(3)

MA-33-A(7)

MA-33-A(7)

SAA-33-A(3)

MA-33-A(7)

MA-33-A(7)

SAA-33-A(3)

Reference [CE90]

l

TABLE 2-8 HEAT NUMBER IDENTIFICATIONOF SURVEILLANCEPLATES 1.

Original Surveillance Program Prior to Identification of Material Mixup SURVEILLANCE PLATE 30 CAPSULE BASE 6-8-3 6-8-3 G-8-3 WELD 6-8-3 G-8-3 6-8-3 HAZ G-8-3 6-8-3 6-8-3 HEAT NO.

P2130 P2130 P2130 P2130 P2130 P2130 P2130 P2130 P2130 2.

Current Surveillance Program After Resolution of Material Mixup Issue and Recapsulation SURVEILLANCE PLATE CAPSULE CASE G-8-1 6-8-3'-8-1 6-8-3" 6-8-1 6-8-3'ELD N/A 6-8-3 6-8-3 HAZ no HAZ reinserted 6-8-3 G-8-3 HEAT NO.

P2112 P2130 P2112 P2130 P2130 P2112 P2130 P2130 NOTES: '.

Includes archive 6-8-3 Charpy specimens fabricated during re-insertion Material in capsule B should be confirmed by chemical analysis when the capsule is pulled.

Reconstituted from original weld material Contains G-8-3 material which could be reconstituted into Charpy or tensile specimens

l

~l

~

ff lf 4

l

31 TABLE 2-9 NMP-1 Beltline and Surveillance Materials Best Estimate Chemistry IDENTIFICATION CHEMICALCOMPOSITION (WT%).

SURVEILLANCE PLATES':

6-8-1 6-8-3 Cu 0.242 0.176 Ni 0.495 0.586 0.041 0.023 0.023 0.017 Plate G-307-3'late 6-307-4/5'late 6-307-1 0'late 6-8-1'late 6-8-3/4 Weld Seam 2-564 A/C'eld Seam 2-564 D/F'eld Seam 3-564'URVEILLANCE WELD':

0.20 0.27 0.22 0.23 0.18 0.22 0.22 0.22 0.17 0.48 0.53 0.51 0.51 0.56 0.2 0.2 0.2 0.07 0.018 0.034 0.019 0.030 0.018 0.026 0.021 0.028 0.012 0.027 0.015 0.020 0.015 0.020 0.015 0.020 0.022 0.013

'verage of Battelle/Westinghouse data [MA91]

'ukens ladel analysis [LU64]

'E recommendation (CE90]

I ct

.I

~

I

~

~

I

UPPER SHELF ENERGY vs.

FLUENCE NINE MILE POINT UNIT 1

32 100 90 80 EQUIVALENT 6-307< DATA BASED ON 6-8-1/6-8-3 DATA N

II-U.

70 60 50 40 RG1.99(2)

PARA 1.2 RG1.99(2)

PARA 2.2 EQUIVALENT G-3074 EQUIVALENT 6-307-4 30 20 10 BASED ON CHEMISTRY DATA 0

1 e+17 1e+18 1e+1S FAST NEUTRON FLUENCE (n/em~+2)

Model Descri tion

-Plate G-8-3 measured L-T to T-L conversion of 0.8 applied

-Equivalent 6-307-4 data used

-b,USE conservatively calculated using average unirradiated data and lowest irradiated datum Figure 2-1 Comparison of USE Prediction Using RG1.99 (2)

Paragraph C.1.2 and C.2.2 Methods for Limiting Plate G-307%

l

33 4.0 Information Re ardin Generic Letter 88-11 3.

Addressees are requested to provide the followinginformation regarding commitments made to respond to GL 88-1 f:

ao How the embrittlement effects of operating at an irradiation temperature (cold leg or recirculation suction temperature) below 525'F were considered.

In particular, licensees are requested to describe consideration given to determining the effect of lower irradiation temperature on the reference temperature and on the Charpy upper shelf energy.

A best estimate of the vessel beltline metal temperature during neutron irradiation was calculated.

The mixing enthalpy was calculated to estimate the water temperature in the region of mixing above the core.

These data were then averaged with the recirculation water thermocouple measurements and compared with beltline thermocouple data.

In addition, the effects of temperature changes during startup, shutdown, and low power operation were accounted for in the calculations to yield the best estimate effective irradiation temperature.

The temperature of the water in the downcomer above the active fuel where the feedwater mixes with saturated water from the steam separators can be calculated since the pressure, temperature, and mass flow rates of the feedwater and steam separator water are known. The temperature of the mixture approximately 4 feet above the top of the active fuel was calculated to be 528.TF at full power.

The temperature in the recirculation line is monitored during plant operation.

During full power operation, the recirculation thermocouple average reading is 523.8'F, and the vessel OD thermocouple average reading one foot above the core midplane is 530.6'F.

The vessel temperature in the beltline region during full power operation at rated conditions is taken to be the average of the mixing region water temperature (528.TF) and the recirculation line water temperature (523.8'F). This temperature is 526.2'F.

As water flows from the mixing region, it will be heated due to radiation, heat flux from the core, and kinetic effects.

There are also radiative losses from the vessel OD surface.

As the water flows into the recirculation system, heating and radiative losses also occur. Since a detailed analysis of these effects is difficult,the best estimate temperature of the vessel beltline at full power and rated conditions is taken as the average of the mixing region water temperature and the recirculation line water temperature.

This approach is believed to be accurate and conservative since heat losses between the mixing region and the bottom of the active fuel are expected to be small in comparison with the losses which occur in the recirculation line.

.l l

t l

I

34 In order to account for the effects of low temperature neutron irradiation of the vessel, the past operating modes of the plant were reviewed.

It was verified that the only operating modes, other than full power operation at rated conditions, are startup, normal shutdown, SCRAM, and low power operation near rated conditions.

A model was constructed to account for the neutron irradiation of the vessel beltline at times when the downcomer water temperature is below the normal operating temperature.

Based on the projected number of startups and normal shutdowns through end-of-license (EOL), it was estimated that 0.54 EFPY of neutron exposure to the vessel would occur at a temperature significantly below the normal operating temperature at rated conditions.

In addition, 0.48 EFPY of exposure at low power and near rated conditions was also modelled.

Using exposure weighting, the best estimate of the effective vessel irradiation temperature was calculated to be 525'F.

The lower bound OD vessel thermocouple reading was 523'F.

Since the total exposure below 525'F at 25 EFPY is estimated to be 0.54 EFPY, the total fast neutron load on the vessel ID surface at the peak flux position at low temperature is calculated to be below 5 x 10"n/cm'.

Regulatory Guide 1.99 (Rev. 2) states that the procedures for evaluation of the ARTNpT are valid for a nominal irradiation temperature of 550'F, and irradiation below 525'F should be considered to produce greater embrittlement. The NMP-1 surveillance program is within all of the limitations required for use of the Regulatory Guide 1.99 (Rev. 2) calculative procedures.

It is, however, recognized that the vessel irradiation temperature is at the 525'F temperature limitation.

NMPC Research 8 Development is sponsoring the future development and use of plant-specific trend curves.

The plant-specific approach inherently models the actual plant operating conditions when embrittlement data are analyzed.

Work is currently underway to develop the plant-specific models for NMP-1.

The new models, if effective, will be used in future P-T curve calculations.

The current P-T curves, which are valid for up to 18 EFPY, were analyzed under regulatory position 2 of RG1.99(2),

since there are at present three valid surveillance data points.

Therefore, the chemistry factor was determined using actual plant data. Accordingly, the effect of irradiation near or slightly below 525'F has been incorporated into the current P-T curve limits since a plant-specific chemistry factor was calculated using the surveillance data.

Addressees are requested to provide the following information regarding commitments made to respond to GL 88-11:

b.

How their surveillance results on the predicted amount of embrittlement were considered.

lf

~

I' r

I

35 Surveillance Data Credibilit Use of data from surveillance capsules in P-T curve calculations depends on the data credibility as judged by the following criteria defined in Regulatory Guide 1.99 (Rev. 2):

"1.

Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement according to the recommendations'of this guide.

2.

Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-Ib temperature and the upper shelf energy unambiguously.

3.

When there are two or more sets of surveillance data from one reactor, the scatter of ERTgpy values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28'F for welds and 1TF for base metal.

Even ifthe fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values.

Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper shelf energy if the upper shelf can be clearly determined, followingthe definition given in ASTM E185-82 (Ref. 1) ~

4, The irradiation temperature of the Charpy specimens in the capsule should match vessel wall temperature at the cladding/base metal interface within

+25'F.

5.

The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the data base for that material."

Criterion 1

As shown in Table 3-1, plate G-307-4 is the limiting material in terms of radiation embrittlement.

Since the Cu and Ni content of the surveillance plate materials fairly closely matches the 6-307-4 chemistry, a chemistry adjustment factor has been used so that the surveillance materials can be used to represent the behavior of plate 6-307-4. Further details are provided below and in Reference

[MA91] and are summarized below.

Criterion 2 Plots of the Charpy energy versus temperature for the surveillance plates are shown in Figures 3-1 through 3-3. As discussed in detail in [MA91],35'F - 40'F is a typical 2o level for LWR materials in the transition region.

Similarly, 10-15 ft-Ib is a typical 2a level for upper shelf data [MA91]. As shown in Figures 3-1 through 3-3, the NMP-1 data scatter is typical of that observed in U.S. LWR vessel steels.

l I

l

,II I

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I

36 Criterion 3 The equivalent plate G-307-4 measured shift data are compared with the Regulatory Guide 1.99 (Rev. 2) BRTg>> calculation based on the paragraph 2.1 methodology in Table 3-2. Allof the measured shift data are essentially within the 2a (34'F) criterion and one datum falls within one a (17'F). Some deviation from the shape of the Regulatory Guide 1.99 (Rev. 2) model in the low fluence region is expected since the model was developed largely from higher fluence PWR data.

NMPC is currently developing a plant-specific B,RT>> model for NMP-1.

It is anticipated that the new model will more accurately represent the low fluence BWR vessel embrittlement behavior.

Criterion 4 The irradiation temperature of the surveillance specimens and the cladding/base metal interface temperature are very nearly in thermal equilibrium with the downcomer water during full power steady-state operation.

The specimens are packed tight in a thin-walled stainless steel container which was evaluated and backfilled with one atmosphere of helium.

Therefore, good heat transfer between the specimens and the downcomer water has been achieved.

The specimens and the cladding/base metal interface are expected to be within+5'F during power operation.

Capsules A'nd C'ontain meltwires and the capsule temperature will be measured when these capsules are pulled in the future.

Criterion 5 The NMP-1 surveillance capsules pulled to date did not include correlation monitor materials.

~ART>> Determination As a result of the material mixup during fabrication of the original surveillance program specimens, the current surveillance capsules contain both G-8-1 and G 3 plate material.

A detailed discussion of the material mixup issue and its resolution is provided in Reference

[MA91]. As shown in Table 3-1, the limiting beltline material at present and through EOL is plate G-307-4. The results shown in Table 3-1 were obtained using the Regulatory Guide 1.99 (Rev. 2) analytical model.

As discussed earlier, the surveillance material is not the limiting material from a radiation damage perspective.

Therefore, in order to be able to use the surveillance data directly in P-T curve calculations, correction factors, which account forthe small chemistry differences between the surveillance plates (G-8-1, G-8-3) and the limiting plate (G-307-4), have been developed.

The Regulatory Guide 1.99 (Rev. 2) chemistry factor data base was used to provide the correction terms.

The chemistry adjustment factor developed is given by:

l l

I l

I l

I

37

~=JF't0 28-0 a [g~- o - -/pe]

30 30

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

CF = Regulatory Guide 1.99 (Rev. 2) chemistry factor dT>>'

measured 30 ft-Ib shift j = G-8-1 or G-8-3 material The results of application of the chemistry adjustment factor to the NMP-1 surveillance data is shown in Table 3-3.

The equivalent G-307-4 data were used, in conjunction with the guidance provided in Regulatory Guide 1.99 (Rev. 2) Paragraph C.1.2, to determine the amount of embrittlement expected at 18 EFPY. A plot of the data and d RT>> fit to the data are provided in Figure 3-4.

The ART>> data at 18 EFPY is shown in Table 3-4 and was used to update the NMP-1 P-T curves and revise the Plant Technical Specifications.

These proposed changes to the Plant Technical Specifications, which are fully responsive to the guidance provided in NRC GL 88-11, were submitted to the NRC on May 13, 1991 [SY91a], and approved through issuance of Amendment 127 on February 10, 1992.

Addressees are requested to provide the followinginformation regarding commitments made to respond to GL 88-11:

C.

Ifa measured increase in reference temperature exceeds the mean-plus-two standard deviations predicted by Regulatory Guide 1.99, Revision 2, or ifa measured decrease in Charpy upper shelf energy exceeds the value predicted using the guidance in Paragraph C.1.2in Regulatory Guide 1.99, Revision 2, the licensee is requested to report the information and describe the effect of the surveillance results on the adjusted reference temperature and Charpy upper shelf energy for each beltline material as predicted for December 16, 1991, and forthe end of its current license.

As shown in Table 3-5, all but one of the d RT>> and all of the b,USE data points for the NMP-1 surveillance materials are within the range anticipated by the Regulatory Guide 1.99 (Rev. 2) models.

The G-8-1 plate hT>> for the specimens

I I

I I

I

38 irradiated to 4.78 x 10" n/cm's 1.3'F higher than the Regulatory Guide 1.99 (Rev. 1) model prediction.

Since this difference is small, and the other two dT>>

measurements are much lower than the Regulatory Guide 1.99 (Rev.

2) predictions, the measured bT3Q data are judged to be within the RG1.99 (2) prediction.

Therefore, the P-T curves have been revised in accordance with the guidance provided in Regulatory Guide 1.99 (Rev. 2). Additional actions to resolve the low USE issue are described in the response to information request 2.a.

l I

TABLE3-1a ADJUSTED RT~~ FOR NINE MILE POINT UNIT 1 BELTLINEMATERIALS 39 Heat No.

Wt.%

Wt.%

Cu Ni Plates - Beltline 18 EFPY'nitial

'4'T('F) d RTMargin ART

~T-L

~F

~F

~F 25 EFPY EOL d RTD, Margin ART,~

~F

~F

~F G-8-3/G-8-4"'-8-1~'~

G-307-3 G-307-4"~

G-307-10 0.18 0.23 0.20 0.27 0.22 0.56 0.51 0.48 0.53 0.51

-3 36 28 40 20 55 65 57 73 63 34 86 34 135 34 119 34 147 34 117 65 76 67 86 74 34 96 34 146 34 129 34 160 34 128 NOTE:

(Peak fast (E>1MeV) fluence:

1.027 x 10" n/cm't vessel 1/4T position at 18 EFPY; 1.426 x 10" n/cm't vessel 1/4T position at 25 EFPY.

"'hese materials are also in the reactor vessel surveillance program

"'imiting plate

'" Calculations performed at vessel 1/4T position using peak beltline flux

[MA91]

I I

40 TABLE3-1b ADJUSTED RTp~ FOR NINE MILE POINT UNIT 1 BELTLINE MATERIALS Welds -

Beltline'"'eat No.

18 EFPY'nitial Ropy ( F)

ARTgpy Margin ART>>

~T-L

~F

~F

~F 25 EFPY@

EOL

ART, Margin ART

~F

~F

~F W5214/5813F

'6054/4E5F 1248/4K1 3F 1248/4M2F

-50

-50

-50

-50 115 66 115 66 115 66 115 66 131 131 131 131 134 66 150 134 66 150 134 66 150 134 66 150 (Peak fast (E>1MeV) fluence:

1.027 x 10" n/cm't vessel 1/4T position at 18 EFPY; 1.426 x 10" n/cm't vessel 1/4T position at 25 EFPY.

"'his material is in the reactor vessel surveillance program.

'" Calculations performed at vessel 1/4T position using peak beltline flux.

[MA91]

"'he most conservative approach described in RG1.99(2) was used to demonstrate that the welds are not limiting in comparison with the plate material.

I I'

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41 Fast Fluence (n/cm')

Table 3-2 Comparison Between Measured Shift Data and Regulatory Guide 1.99 (Rev. 2) Paragraph 2.1 Calculated d RT Equivalent Plate B,RTNQ7 RG1.99 (2) Difference Between G-307-4 G-307-4 Shift Paragraph 2.1 Shift and RG 1.99 (2)

('F)

Paragraph 2.1 BRTNpy 4.78 x 10" 3.6Q x 1Q 4.78 x 10" 85.0 60.0 23.7 58.6 50.1 58.6 26.4 9.9 34.9

I.

I

42 Adjustment Factor"'o Obtain Plate G-307-4 Material E uivalent Data Fluence

~ncm~

Measured Shift F

Equivalent Plate G-307-4 Shift F

G-8-1 G-8-1 G-8-3 FF (19. 9)

+ bT FF (19. 9)

+ bTq FF (43. 7)

+ hT~

4.78 z 10'~

3.60 x 10"

4. 78 x 10~~

79.3 55 '

11.2 85.0 60.0 23 '

(1)

FF = f' '~ " f = fast fluence

1Q" (n/cm')

Table 3-3 Surveillance Data Chemistry Adjustment Factors to Obtain Plate G-307-4 Equivalent Data for NMP-1

I I

I

43

SUMMARY

OF THE ARTNDT FOR PLATE G 307 4 AT 1 8 EFPY USING REVISION (2) OF REGULATORY GUIDE 1.99 Exposure Fluence"'G1.99(2)

Location

~EFPY

~n/cm'~RT>>~F 1/4 T 3/4 T 18 1.027 x 10'8 4.286 x 10'44 113 (1)

The RG1.99(2) DPA attenuation model was used to calculate the 1/4 T and 3/4 T fluences.

Table 3-4 Summary of the ART>> Data for Limiting Plate G-307< at 18 EFPY

I I

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Table 3-5 44 Comparison of Charpy Impact Properties for Irradiated Materials from the NMP-1 Surveillance Program with the Regulatory Guide 1.99 (Rev. 2) Model Prediction 30 ft-lb Reg.Guide Fluence Transition 1.99 (2) Shift (E>1.0MeV) Temp.

Plus 2a Material (n/cm')

Shift ('F)

('F)

Upr. Shelf Reg.Guide Energy 1.99 (2)

Drop USE Drop

(%)

(%)

G-8-1 4.78x10'9.3 G-8-1 3.60x10'5.1 G-8-3 4.78x10'"

11.2 78.0 71.6 71.2 00) pP) 16 15 13

"'he irradiated USE is slightly higher than the unirradiated USE

"'nsufficient data to determine USE

"'nlyone USE point available. The irradiated USE is higher than the unirradiated USE

45 NINE MILE POINT UNIT 1

PLATE G-8-1 SHIFT AT 8,6X10oo17(N/CMaa2) 120 100 IXI II-IL 80 60 40 20 I

UNIRRADIATED EXP. DATA IRRADIATED EXP. DATA WEIBULL FIT TRANSITION HYPERBOLIC TANGENT FIT WEIBULL FIT TRANSITION HYPERBOLIC TANGENT FIT

-100 -50 0

50 100

'I 50 200 250 TEST TEMPERATURE (F)

Figure 3-1 Effect of Irradiation on Plate G-8-1 After a Fast Fluence of 3.6 x 10" n/cm'

46 NINE MILE POINT UNIT PLATE G-8-1 SHIFT AT 4.8X10+a'$7(N/CMaa2)

EXP. DATA 120 IRRADIATED EXP. DATA 100 80 WEIBULL FIT TRANSITION 60 WEIBULL FIT UPPER SHELF i

I-i I

40 20 HYPERBOLIC TANGENT FIT WEIBULL FIT TRANSITION HYPERBOLIC

-50 0

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

I Figure 3-2 Effect of Irradiation on Plate G-8-1 After a Fast Fluence of 4.78 x10'/cm

I I

I

NINE MILE POINT UNIT PLATE Q-S-S SHIFT AT 4.8X10++17(N/CM++2) 120 o

UNIRRADIATED EXP. DATA IRRADIATED EXP. DATA 100 I

t t

I 40 i

~~

2o I

o

/I I

WEIBULL FIT TRANSITION WEIBULL FIT UPPER SHELF HYPERBOLIC TANGENT FIT

/

lg p a~wc WEIBUL FIT TRANSITION HYPERBOLIC TANGENT FIT

-'IOO -60 0

60 100 150 200 250 TEST TEMPERATURE (F) 1 Figure 3-3 Effect of irradiation on Plate G-8-3 After a Fast Fluence of 4.78 x 10" n/cm'

48 NINE MILE POINT UNIT PRESSURE VESSEL EMBRITTLEMENT TREND 1000 O

o r r

rr SHIFT IN RTndt ADJUSTED RTndt CAPSULE DATA

'IO 0 e+17 1e+18 1e+19 1e+20 FAST NEUTRON FLUENCE (E~1.0 Mev)

Figure 3-4 ARTand d,RT, for NMP-1 Limiting Belttine Plate G-307-4

I I

I

5.0 References 49

[BTP81] NRC Branch Technical Position MTEB5-2, "Fracture Toughness Requirements",

Revision 1, July 1981.

[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, Inc., Windsor, Connecticut, June, 1990.

[FR92]

Freyer, P., Manahan, M.P., Presentation to Project FERMI, "Plant Life Extension Technology: Non-Destructive Reactor Materials Embrittlement Monitoring Using Positron Annihilation", May, 1992.

[6E85]

Caine, T., Ranganath, S., "BWR Owners Group Evaluation of Regulatory Guide 1.99 Proposed Revision 2 Impact on the BWR", MDE-190-0985, October, 1985.

[6E92]

Caine, T., Ranganath, S., "BWR Beltline Material Upper Shelf Energy Estimation Methods", GE-NE-523-1 8-1191, June, 1992.

[LU64]

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

[MA85] Manahan, M.P., "Procedure forthe Determination of Initial RT>>in Cases where Limited Baseline Data are Available", November, 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.

[MA87]. Manahan, M. P., "Surveillance Capsules A'nd C'or Nine Mile Point - Unit 1",

September 30, 1987, Draft Final Report to Niagara Mohawk.

[MA90] Manahan, M.P., "Nine Mile Point Unit 1 Ropy Determination", Final Report from MPM Research R Consulting to NMPC, Sept. 28, 1990.

[MA91] Manahan, M. P., "Nine Mile Point Unit 1 Surveillance Capsule Program", NMEL-90001, January 4, 1991

[MA91b] Private communication, M.P. Manahan (MPM Research

&Consulting) to J. Helm (Columbia University), "Physically Based Upper Shelf Fracture Model for Ferritic Pressure Vessel Steels", January, 1991.

[Mc FRAC]

Manahan, M.P., et.al., "Statistical Analysis Methodology for Mechanics of Fracture", User's Manual, Version 2.2, September, 1989.

I

50

[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.

[ST64]

Stone, W.A., Mechanical Test Report (contract 164 Niagara Mohawk), from W.

A. Stone Metallurgical R 8 D Department, Combustion Engineering, July-October, 1964.

[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.

[SY91]

Letter from B. Ralph Sylvia to USNRC,

Subject:

"Response to Request for Additional Information Regarding Reactor Vessel Pressure-Temperature Limits Technical Specification Amendment", September 27, 1991.

[SY91a] Letter from B. Ralph Sylvia to USNRC,

Subject:

Nine Mile Point-Unit 1 Docket No. 50-220, DPR-63", May 13, 1991.

[USAR] Nine Mile Point Nuclear Station Unit 1, Final Safety Analysis Report (Updated)

Revision 10, June, 1992.

Appendices

I I

52 Appendix A Surveillance Capsule Specimen Types and Materials

53 S ecimen Identification S ecimen T e

TABLE A-1 SURVEILLANCECAPSULE A'PECIMEN TYPES AND MATERIALS'ase Metal Plate Material E71A E12 E31A E2E E2T E2UA E17 E1AA E2Y E1CA E1D EBKA ED1 ED2 Reconstituted Charpy Base Charpy Base Reconstituted Charpy Base Charpy Base Charpy Base Reconstituted Charpy Base Charpy Base Reconstituted Charpy Base Charpy Base Reconstituted Charpy Base Charpy Base Reconstituted Charpy Base Charpy Weld Charpy Weld G-8-1 6-8-1 6-8-1 6-8-1 6-8-1 6-8-1 6-8-1 6-8-1 6-8-1 6-8-1 G-8-1 6-8-1 6-8-3 6-8-3

54 TABLE A-1 SURVEILLANCECAPSULE A'PECIMEN TYPES AND MATERIALS'ase Metal S ecimen Identification S ecimen T e

Plate Material ED3 ED4 ED5 ED6 ED7 EDA EDB EDC EDD EDE J12 J13 J14 J15 J16 J17 Charpy Weld Charpy Weld Charpy Weld Charpy Weld Charpy Weld Charpy Weld Charpy Weld Charpy Weld Charpy Weld Charpy Weld Charpy HAZ Charpy HAZ Charpy HAZ Charpy HAZ Charpy HAZ Charpy HAZ Charpy HAZ 6-8-3 6-8-3 6-8-3 6-8-3 G-8-3 6-8-3 6-8-3 G-8-3 6-8-3 6-8-3 6-8-3 6-8-3 6-8-3 6-8-3 6-8-3 6-8-3 G-8-3

I

55 TABLE A-1 SURVEILLANCECAPSULE A'PECIMEN TYPES AND MATERIALS" Base Metal S ecimen Identification S ecimen T e

Plate Material J1A J1B J1C J1D JD1 JD2 JLK JL2 JTA JUL JUJ Charpy HAZ Charpy HAZ Charpy HAZ Charpy HAZ Charpy HAZ Tensile Base Tensile Base Tensile Base Tensile Base Tensile Weld Tensile Weld Tensile HAZ Tensile HAZ Tensile HAZ 6-8-3 6-8-3 G-8-3 6-8-3 6-8-3 6-8-1 G-8-1 6-8-1 G-8-1 n/a n/a 6-8-3 6-8-3 G-8-3 (1)

The base metal tensile and base from Charpy weld should be confirmed by chemical analysis when the capsule is pulled.

I

56 TABLE A-2 SURVEILLANCECAPSULE B SPECIMEN TYPES AND MATERIALS'pecimen Identification All Base Metal Specimens AllWeld Specimens All HAZ Specimens Specimen

~TB Charpy Tensile Charpy Tensile Charpy Tensile Base Metal Plate Material 6-8-1 6-8-1 G-8-3 n/a 6-8-3 6-8-3

'hese materials should be confirmed by chemical analysis when the capsule is pulled,

l f

57 TABLE A-3 SURVEILLANCECAPSULE C'PECIMEN TYPES AND MATERIALS'ase Metal identification NC01 NC21 NC02 NC22 NC03 NC23 NC04 NC24 NC05 NC25 NC06 NC26 E1JA E1JB E1KA S ecimen 7 e

Charpy Base Charpy Base Charpy Base Charpy Base Charpy Base Charpy Base Charpy Base Charpy Base Charpy Base Charpy Base Charpy Base Charpy Base Reconstituted Charpy Base Reconstituted Charpy Base Reconstituted Charpy Base S ecimen Plate Material 6-8-3 6-8-3 6-8-3 6-8-3 6-8-3 6-8-3 6-8-3 G-8-3 6-8-3 6-8-3 6-8-3 6-8-3 6-8-1 G-8-1 6-8-1

58 TABLE A-3 SURVEILLANCECAPSULE C'PECIMEN TYPES AND MATERIALS'ase Metal Identification S ecirnen T e

~s Plate Material E1KB EA5A EA5B E42A E1MA E1UA

/

E3TA E7EA J2CB EDKA EDLA Reconstituted Charpy Base Reconstituted Charpy Base Reconstituted Charpy Base Reconstituted Charpy Base Reconstituted Charpy Base Reconstituted Charpy Base Reconstituted Charpy Base Reconstituted Charpy Base Reconstituted Charpy Base Reconstituted Weld Reconstituted Weld 6-8-1 6-8-1 6-8-1 6-8-1 6-8-1 6-8-1 6-8-1 6-8-1 6-8-3 n/a n/a

59 TABLE A-3 SURVEILLANCECAPSULE C'PECIMEN TYPES AND MATERIALS

'ase Metal Identification S ecimen T e

~S Plate Material EDMA EJTA JAEA Reconstituted Weld Reconstituted Weld Reconstituted Weld.

n/a n/a n/a JAMA J2CA J1LA J1MA Reconstituted Weld Reconstituted Weld Reconstituted Weld Reconstituted Weld n/a n/a n/a n/a J1PA J1TA J1JA T01 Reconstituted Weld Reconstituted Weld Reconstituted Weld Tensile Base n/a n/a n/a 6-8-3

I

60 TABLE A-3 SURVEILLANCECAPSULE C'PECIMEN TYPES AND MATERIALS' ecimen Identification S ecimen T e

Base Metal Plate Material T02 T21 10 Tensile Base Tensile Base Tensile Base Tensile Base Tensile Base Tensile Base Tensile Weld Tensile Weld Tensile Weld Tensile Weld 6-8-3 G-8-3

.6-8-1 6-8-1 G-8-1 6-8-1 n/a n/a n/a n/a (1)

The base metal tensile and base from Charpy weld should be confirmed by chemical analyses when the capsule is pulled.

Appendix B NMP-1 Beltline Material Chemistry Data

I

62 Appendix B.1 1964 Lukens Ladle Analysis

.I

63 1964 Lukens'easured'hemistry of Beltline Plates for NMP-1 [LU64]

CHEMICALCOMPOSITION (WT %)

Heat No.

Plate ID Fe Cu Ni Mn Mo C

S Si P2074 G-307-3 MATRIX 0.20 0.48 0.018 1.45 0.45 0.18 0.034 0.26 P2076 6-307-4 6-307-5 MATRIX 0.27 0.53 0.019 1.25 0.52 0.2 0.03 0.21 P2091 P2112 P2130 6-8-1 G-8-3 6-8-4 MATRIX MATRIX 6-307-10 MATRIX 0.22 0.23 0.18 0.51 0.018 0.51 0.021 0.56 0.012 1.43 0.51 0.2 1.34 0.45 0.19 1.16 0.47 0.2 0.026 0.26 0.028 0.21 0.027 0.17

'ased on discussions with Lukens, data from ladle analysis by atomic absorption

'imiting plate from a radiation damage perspective

l

.I

64 Appendix B.2 CE Determined Beltline Weld Chemistry

l f

65 BELTLINEWELD CHEMISTRY DATA[CE90]

2-564 A/C 86054/4E5F 1248/4K1 3F

.12

.11 1.64 1.71

.015

.005

.020

.017

.34

.38

.51

.56 0.2 0.2 0.22 0.22 2-564 D/F 3-564 86054/4E5F 1248/4M2F

.12 1.64,

.10 1.26

.015

.015

.020

.34

.51 0.2

.020

.22

.57 0.2 0.22 0.22 Surveillance Capsule Weld 5214/5G13F

.14 1.58

.018 (0.022)

.013

.25

.51 0.2(0.07)'.18 (0.17)

'ata recommended by Reference [CE90] based on CE database

'ata in parenthesis were measured using irradiated material as reported in [ST84]

I l'

.I I

66 Appendix B.3 Battelle and Westinghouse Measured Chemistry Data.

l l'I

egg/ ~'

~ ~

gg

~ ~

g

~p

SUMMARY

OF CHEMICALANALYSIS'EASUREMENTS MADE BY BATTELLEAND WESTINGHOUSE 67 IDENTIFICATION Cu Ni CHEMICALCOMPOSITION (WT. %), MEAN c'o S6 Note Si PLATE: BASE FROM BASE SPECIMENS 6-8-1 (Capsule C) 6-8-1 (Capsule A) 0.243 0.238 0.244 0.241 0.510 0.473 0.518 0.480 0.041 1.360 0.04 1.369 0.430 0.210 0.438 0.206 0.023 0.022 0.226 0.228 (4)

(4)

WELDS:

6-8-3 (Capsule C) 0.17 0.070 0.022 (3)

PLATE: BASE FROM HAZ SPECIMENS 6-8-3 (Capsule C) 0.171 0.165 0.626 0.564 0.023 1.133 0.484 0.227 0.018 0.160 (4)

UNIRRADIATEDARCHIVE:

6-8-3 0.176 0.181 0.178 0.183 0.599 0.580 0.570 0.579 0.022 1.155 1.140 0.476 0.494 0.249 0.016 0.163 (4)

(4)

NOTES: 1.

2.

3.

4.

5.

6.

7.

All measurements by ICAP unless otherwise noted For the details of analytical data for other elements, see attached sheets By x-ray fluorescense (XRF)

By atomic absorption By LECO combustion By titrimetric combustion By gravimetry

I*

f I'

~

Ol 1

I I

I.

CHEMICAL ANALYSIS RESll.TS (WMAL) FOR NINE MILE POINT UNIT 1.

MODIFIED A382B MATERIAL

[MAST]

BASE FROLl BASE 366 DEGREE CAPSULE (1)

E42(A)

E42(B)

ETE E1M E1U E3T E3T(R)

STANDARD RELATIVE MEAN DEVIATION DEVIATION,X Fe Cu Hi Cu(2)

Hi (2)

P Mn Co Mo V

Cr Ti C(3)

S(4)

Si (5)

LlATRIX MATRIX MATRIX 6.236 8.236 8.258 8.479 8.517 8.539 8.236 8.473 8.836 8.848 8.845 1.324 1.334 1.378 8.818 8.818 8.818 8.486 8.426 8.462

<8.665

<8.885

<8.865 8.185 8.166 8.189 8.681 8.881

<8.861 8.218 8.823 8.22S 8.226 8.22S MATRIX MATRIX 8.243 8.241 8.462 8.466 8.839 8.834 1.36T 1.352 8.818 8.818 8.466 8.416

< 8.885

<8.685 8.186 8.186

<8.681 8.681 8.843 8.845 1.394 1.377 8.811 8.818 8.452 8.445

<6.665

<8.665 8.113 8.115

<8.881

<8.661 MATRIX MATRIX 8.24S 8.247 6.665 8.543 8.522 8.827 8.665 8.817 8.664 8.825 8.886 8.822 2.871 8.243 5.339 8.518 2.181 8.236 3.594 8.473 9.947 8.841 1.669 1.368 3.72S 8.618 5.17S 8.431

<6.885 3.495 8.189

<8.881 6.218 8.823 8.226 NOTES:

(1) ALL MEASUREMENTS BY ICAP UNLESS OTHERNISE NOTED (2)

BY ATOLIIC ABSORPTION (3)

BY LEO COLlBUSTION (4)

BY TITRIllETRIC COMBUSTIOH (5)

BY GRAVIMETRY

l

.fi I'

I

cHEMIcAL ANALYsls REslLTs (wMAL) F0R NINE MILE PDIHT UNIT 1. MQDIFIED A362B MATERIAL (l4AB7)

BASE FROM BASE 38 DEGREE CAPSULE (1)

EIA(A)

ElA(B)

E1C EBK E2U E31 STANDARD RELATIVE MEAN DEVIATIOH DEVIATION,X Fe Cu Hi Cu(2)

Hi (2)

P Mn Co Mo V

Cr Ti C(3)

S(4)

Si (5)

MATRIX 8.244 8.512 8.241 8.488 6.841 1.377 8.818 8.433

<8.665 8.188 8.861 8.286 8.622 8.228 8.839 1.361 8.818 8.435

<e.ses 8.186 8.881 8.842 1.369 S.S18 8.421

<8.685 8.166

<8.861 6.641 1.362 8.811 8.466

<8.665 8.186

<8.881 6.639 1.344 8.818 8.432

<8.885 8.186

<8.861 8.839 1.388 8.818 8.436

< 8.885 8.189

<8.881 8.226 8.228 MATRIX MATRIX MATRIX

. MATRIX MATRIX 8.243 8.243 8.256 8.246 8.246 8.514 8.568 8.558 8.518 8.519 8.863 8.817 B.ess

~ 8.825 8.661 8.814 B.ees 8.815 1.386 8.244 3.366 8.516 2.875 8.241 5.286 8.468 3.369 8.848 1.855 1.369 4.816 8.818 3.454 8.436

<B.ses 8.912 8.166

<8.861 8.286 8.822 8.226 NOTES:

(1) AU. MEASUREMENTS BY ICAP UNLESS OTHERWISE NOTED (2)

BY ATOMIC ABSORPTION (3)

BY LECO COMBUSTIOH (4)

BY TITRIMETRIC COMBUSTIOH (5)

BY GRAVIMETRY

~

f

.l

.I

CHEMICAL ANALYSIS RESR.TS (WMAL) FOR NINE MILE POINT UHIT 1. MODIFIED A382B MATERIAL [MA87]

BASE FROM HAZ 388 DEGREE CAPSR.E (1)

JIL(A)

JIL(B)

JAM JAE JIT JIP STANDARD RELATIVE MEAN DEVIATION DEVIATION,X Fe Cu Ni Cu(2)

Hi (2)

P Mn Co Mo V

Cr Ti C(3)

S(4)

Si (5)

MATRIX 8.172 8.681 8.165 8.564 8.823 1.142 8.818 8.456

<8.885 8.864

<8.881 8.227 8.818 8.168 8.826 1.163 8.811 8.463

<8.885 8.865

<8.881 8.822 1.135 8.811 8.581

<8.885 8.892

<8.881 8.823 1.152 8.811 8.495

<8.885 8.864

<8.881 8.821 1.149 8.811 8.466

<8.885 8.863

<8.881 8.822 1.854 8.818 8.461

<8.885 8.877

<8.881 8.168 8.168 MATRIX MATRIX MATRIX MATRIX MATRIX 8.173 8.173 8.174 8.173 8.164 8.626 8.654 8.644 8.639 8.598 8.884 8.825 8.884 8.819 8.882 8.848 8.881 8.816 2.174 8.171 4.842 8.626 2.424 8.165 3.369 8.564 7.543 8.823 3.497 1.133 4.641 8.811 3.213 8.464

<8.885 5.694 8.864

<8.881 8.227 8.816 8.168 NOTES:

(1) ALL MEASUREMEHTS BY ICAP UNLESS OTHERWISE NOTED (2)

BY ATOMIC ABSORPTIOH (3)

BY LECO COMBUSTIOH (4)

BY TITRIMETRIC COMBUSTIOH (5)

BY GRAVIMETRY

.I I

I I

CHEMICAL ANALYSIS RESULTS (eMAL) FOR NIHE MILE POIHT UNIT 1.

MODIFIED A382B MATERIAL [HA87]

UNIRRADIATED ARCHIYE PLATE (1)

D25(A)

D25(B)

D21 Del STANDARD RELATIVE MEAH DEVIATIOH DEVIATION,X Fe Cu Hi CQ(2)

Hi (2)

P Mo Co Mo Y

Cr Ti C(3)

S(4)

Si (5)

MATRIX 8.173 8.562 8.176 8.573 8.823 1.146 8.81$

b.468

<b.885

$.883

<b.eel b.249 S.bid

$.163 MATRIX MATRIX 8.188 b.176 8.589 b.628 8.177 d.566 8.823

$.823 1.188 1.143

$.81$

b.bib 8.477 b.491

<$.885

<8.886 S.bed

$.864

<8.$$ 1

<S.ddl S.ld3

$.163 MATRIX 8.175 8.684 8.816 1.151

$.811

$.469

<S.ees 8.884

<8.881 8.883 8.817 8.883 8.822 8.882 8.817 8.888 b.dll S.eee 8.881 1.673 8.176 2.619 8.599 1.698 8.178 3.663 8.578 11.494 8.822 1.471 1.155 4.878 8.818 2.231 8.476 S.eee

<8.885 1.494 8.884

<8.881 8.249 8.816

8. 163 NOTES:

(1) All MEASURB!ELITS BY ICAP QLESS OTHERIISE NOTED (2)

BY ATOLIIC ABSORPTION (3)

BY LEO COLSITIOH (4)

BY TITRIMETRIC COMRSTIOH (5) BY GRAVIMETRY

I' l

I

ICAP CHEMICAL ANALYSIS REPORT (KL) FOR HINE LIILE POINT UNIT l.

[MA87]

UNIRRADIATED ARCHIVE PLATE (1)

D25(A)

D25(B)

D21 Dbl STANDARD RELATIVE MEAN DEVIATION DEVIATIOH,X MATRIX MATRIX LlATRIX 8.188 8.175 8.168 8.683 8.656 8.67S 8.184 8.162 8.164 b.684 8.575

- 8.678 1.198 l.lib 1.148 Fe Cu Ni Cu(2)

Ni (2) ih Co Mo 8.497

$.495 8.493 V

Cr 8.122 8.117 8.118 Ti C(3)

S(4)

Si (S)

MATRIX 8.162 8.885 8.585 8.828 8.182

$.883 8.578 8.822 1.128 8.836 8.492 8.$ 82 8.121 8.882 2.967 8.181 3.364 8.588 1.639 8.183 3.681 B.S79 3.122 1.148 8 449 8 494 1.992 8.128 NOTES:

(1) ALL MEASUREMENTS BY ICAP UNLESS OTHERIISE HOTED (2) BY ATOLIIC ABSORPTION (3)

BY LECO COLIUSTIOH (4)

BY TITRIMETRIC COLSSTIQH (S)

BY GRAVIMETRY

~

l l'

CHEMICALANALYSIS'ESULTS FOR NINE MILE POINT WELD METALSPECIMENS 73 Specimen No.

EDK EDK EJD EJC EDT EDL EDL Material Type'eld (I)

Weld (I)

Weld (I)

Weld (I)

Weld (I)

Weld (I)

Weld (I) 0.17 0.18 0.16 0.17 0.16 0.16 0.17 Elements Wei ht Percent Cu Ni P

0.05 0.05 0.06 0.08 0.09 0.08 0.08 0.020 0.022 0.020 0.022 0.022 0.023 0.022

'DAX

'I) - Irradiated

I' k

r r

I