ML19323B706
| ML19323B706 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities, Zion  |
| Issue date: | 04/01/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19323B700 | List: |
| References | |
| NUDOCS 8005140106 | |
| Download: ML19323B706 (5) | |
Text
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NUCLEAR REGULATORY CCimvilSSION
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ATYPICAL WELD fiETAL Introduction During 1978, B&W initiated work contracted with the B&W Owners Group on a program for evaluating the material properties of "early vintage" 177-fuel assembly reactor vessel welds. One of the work phases in this program had the objective of characterizing the chemistry of reactor vessel (RVi beltline welds. Extensive chemical analyses of the archive sources of RV welds have been performed as part of this work. Two samples of test weldments made for the Crystal River 3 reactor vessel surveillance program were part of the weld metal archives subjected to chemical analysis. The results of these analyses, perfo'ued by the Mt. Vernon Works Quality Assurance Laboratory, indicated that one of these samples had atypical concentrations of nickel and silicone, while the concentrations of the other elements were in the normal range for MnMcNi:Linde 80 submerged-arc RV weldments. The other sample had the nominal chemistry.
The atypical weld was made with weld wire designated by heat number 72105. This heat of weld wire was used in the fabrication of 12 reactor vessels. These vessels and the location of possible atypical welds are listed in Table 1.
Charpy V-notch tests on the atypical weld metal resulted in a higher than normal value of RTNOT, partially because of unusually high scatter. Therefore, we re-quested that the licensees of the above plants administratively apply revised pressure-temperature operating limits that reflected the possible presence of atypical weld metal.
In calculating these limits the atypical weld was assumed to have an unirradiated RTNDT of 1200F and radiation damage is predicted by the upper limit line in Regulatory Guide 1.99.
Currently all the affected plants are operating under such revised limits.
Discussion 10 CFR Part 50, Appendix G " Fracture Toughness Requirements", requires that pressure-temperature limits be established for reactor coolant system heatup and cooldown operations, inservice leak and hydrostatic tests, and reactor core operation. These limits are required to ensure that the stresses in the reactor vessel remain within acceptable limits. They are intended to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences.
The pressure-temperature limits depend upon the metallurgical properties of the reactor vessel materials. The properties of materials in the vessel beltline region vary over the lifetime of the vessel because of the effects of neutron irradiation. One principle effect of the neutron irradiation is that it causes the vessel material nil-ductility temperature (RTNDT) to increase with time.
The pressure-temperature' operating limits must be modified pe:'iodically to account for this radiation induced increase in RTNDT by increasing the temper-ature required for a given pressure. The operating limits for a particular operating period are based on the material properties at the end of the operating period. By periodically revising the pressure-temperature limits to account for
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radiation damage, the' stresses and stress intensities A the reactor vessel.
are maintained within acceptable limits.
The magnitude ofcthe shift in RTNDT is proportional to the neutron fluence i
that the materials are. subjected to.
The shift in RT DT can be predicted N
from Regulatory Guide 1.99. To check.the validity of the predicted shift in RTNDT, a reactor vessel material surveillance program is required. Sur-i veillance specimens'~are periodically removed from the vessel and tested. The results of these tests are compared to the predicted shifts in RTNDT, and the pressure-temperature operating. limits are revised accordingly, i
Since the unirradiated RTNDT of the atypical weld metal was determined to be i
high, and it was assumed to be sensitive to radiation damage, the atypical i
weld metal would generally be the limiting vessel material. Therefore, all licensees with vessels that might have been fabricated with atypical weld metal were required.to revise their pressure-temperature operating limits to i
reflect the possibility that atypical material was used in their construction.
Evaluation 1
To resolve the atypical weld issue, B&W has conducted an extensive investigation of records, metallographic examinations, chemical analyses, and fracture mechanics tests on both unirradiated and irradiated atypical weld material. The results of this program are presented in BAW-1556.
j Since 1966, 42 heats of submerged-arc weld wire have been purchased for RV and surveillance specimen fabrication at Mt. Vernon, and, except for the discovery-of the partial-thickness off-chemistry conditions in the second CR-3 surveil-lance block, there is no evidence that atypical weld wire reached the shop floor.
i The results of more than 2000 chemical analyses have been reviewed relative to i
the 42 wire heats.. All, except for the one batch of Crystal River survillance i
material, have been within the normal ranges. These include through-thickness
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i tests from seven RVs fabricated at Mt. Vernon and tests of wire currently in j
inventory.
4 Detailed metallographic examinations'were performed on seven fractured Charpy specimens. Both macro-and micro-examination techniques were employed, as well
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as a fractographic examination with a. scanning electron microscope. Relatively littlepomsity was noted in any of the weld material examined. Examinations.
revealed columnar grains' outlined by proeutectoid. ferrite. The orientation of the grains and the unusually.high amount of proeutectoid ferrite are believed l
- to be the cause of:the high scatter in the Charpy. data.
Numerous chemical analyses.were performed on the atypical weldment. The, bulk of these analyses were obtained-using.a Jarrel-Ash emission. spectrometer. The l
concentrations.of.10 elements were measured.by this technique. X-ray floures -
- cense analysis ~was used to measure the concentrations of. nickel, molybdenum,y and copper.in: irradiated Charpy specimens. Results show that the copper content
- was high, averaging.between 0.4 to 0.5%
The chemistry of atypical material f
is compared-to.typicalimaterialfin Table 2.
' Charpy.V-notch' tests were performed on both unirradiated and irradiated material'.-
The-. irradiated specimens were irradiated in the' Crystal. River 3 reactor vessel.
Dynamic' andestatic fracture toughness : tests.were conducted' on one Linch thick
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. I compact tension ~ specimens at room temperature. Although the dropweight NDT 0
is -20 F, the results of the Charpy tests show that 50 ft-lbs of energy is absorbed at 1500F, therefore the unirradiated value of RTNDT is 90 F.
Using RT DT equal to 900F, the toughness properties obtained from the fracture N
mechanics tests, KIe (static) and kid (dynamic), are conservative (lie above) the KIc curve in ASME Code,Section XI and the KIR curve in ASME Code,Section III respectively. Using an RTNDT of -20 F (the dropweight NDT), the fracture 0
mechanics data fall within the scatter of data on normal material used to ebtain the KIc and KIR curves. This indicates that the RTNDT value of 900F is conservative.
The effect of irradiation on the mechanical properties of atypical material have beenevaluated,usingthetestresultsonCrystalRiver3survgllancqspeci-l mens. These specimens were subjected to a fluence of 1.1 X 10 n/cm'. This J
fluence produced an increase in RTNDT of 350F.
From our review we conclude that the probability that atypical weld metal was used in fabricating the subject vessels is very low. However, we feel that in calculating pressure-temperature operating limits for these vessels, the propertics of atypical material should be considered. As discussed above, we have determined that an initial value of RTNDT of 900F is a very conservative value.
The increase in RT DT due to i radiation should be based on the measured N
0 value of 35 F at a fluence of 1.1 X 10 8 n/cmd and the damage prediction slopes in Regulatory Guide 1.99.
We also find that the administratively applied pressure-temperature oper-ating limits may be removed from these 12 plants. The operating limits in the Technical Specifications for Browns Ferry 1 are presently being reviewed and will include limits based on the atypical weld metal.
The Technical Specifica-tions for Rancho Seco have operating limits based on the atypical material that are more restrictive than they would be if based on the criteria developed from this review. Midland 1 is being reviewed for an Operating License and its Technical Specifications have not been finalized to date. The pressure-temperature limits for Three Mile Island 2 should be revised to reflect the possible use of atypical material in this vessel. This poses no immediate problem since this plant is not currently operational. The Technical Speci-fications of the other nine subject plants contain pressure-temperature operating limits that are in accordance with Appendix G, 10 CFR Part 50 based on both typical and atypical weld metal properties.
The staff will continue to monitor the effects of radiation on the properties of the atypical weld material. Six capsules containin metal are in the Crystal River 3 surveillance program.g the atypical weld One of these capsules has already been removed and tested.
Also, there is enough atypical material in storage at B&W to fabricate fracture toughness specimens _up to 1.0T compact fracture toughness specimen size.
TABLE 1.
LOCATION OF POSSIBLE ATYPICAL WELDS PLANT LOCATION OF WELD
-B&W Oconee 3 Center Circ. Beltline TMI 1 Upper Cire. Beltline Lower Cire. Beltline TM! 2 Dutchman to Lowerhead ANO 1 Head to Flange and Nozzle to-Shell Midland 1 Center Circ. Beltline CR-3 Center Cire. Beltline Rancho Seco Vertical Seam Beltline-WESTINGHOUSE Zion 1 Inter to Lower Cire. Beltline Zion 2 Vertical Seam Beltline 0
(o and 180 )
Turkey Pt. 4 Nozzle Shell to Interm. Circ.
gE, Br. Ferry 1 Shell to Flange and Longitudinal Weld in Beltline Quad Cities 2 Closure Head to Flange Y
o TABLE 2.
ATYPICAL WELD CHEMISTRY C
Mn P
S Si Cr fli Mo CR-3 Weld
.08 1.65.021
.013 1.0
.0/
.10
.45 Mn-Mo-fli
.08 1.6
.018
.015
.5
.07
.60
.40 (Typical)
....