ML17339A371
| ML17339A371 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 11/13/1979 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | Robert E. Uhrig FLORIDA POWER & LIGHT CO. |
| Shared Package | |
| ML17339A372 | List: |
| References | |
| NUDOCS 7912100246 | |
| Download: ML17339A371 (9) | |
Text
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NOVEMBER, 1 3 1STS Docket Nos. 50-250 and 50-251 Dr. Robert E. Uhrig, Vice President Advanced. Systems and Technology Florida Power and Light Company Post Office Box 529100 tatami, Florida 33152
Dear Dr. Uhrig:
RE:
REACTOR VESSEL'TYPICAL WELD tMTERIAL DISTRIBUT Docket Fi s 50-250 and NRC 'PDR (2)
Local PDR NRR:Reading ORBl Reading I8(E (3)
Attorney OELD D. Eisenhut R. Tedesco
,. -W. Gammill'-
B. Grimes
'A. Schwencer tl. 'Grotenhuis C. Parrish TERA NSIC Your August 18, 1979,response to our August 19, 1979 request for information regarding the Turkey Point Plant Unit Nos.
3 and 4 has been reviewed.
No further action is required other than the normal review of your reactor vessel surveillance program.
Our related Safety Evaluati'op is enclosed.
'incerely, Original Signed 8/
Enclosure:
Safety Evaluation, cc:
w/enclosure See neRt page A. Schwencer, Chief Operating Reactors Branch 81.
Division of Operating Reactors Ap3 gg 181 00 CI(V(O', ~
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 November 13, 1979 Docket Hos. 50-250 and 50-251 Dr. Robert E. Uhrig, Vice President Advanced Systems and Technology Florida Power and Light Company Post Office Box 529100 Miami, Florida 33152.
Dear Dr. Uhrig:
RE:
REACTOR VESSEL ATYPICAL WELD MATERIAL Your August 18, 1979 response to our August 19, 1979 request for information regarding the Turkey Point Plant Unit Hos.
3 and 4
has
'been reviewed.
Ho further action is required other than the normal review of your reactor vessel surveillance program.
Our related Safety Evaluation i enclosed.
Sincerely,
Enclosure:
Safety Evaluation cc:
w/enclosure See next page A. Schwencer.,
Chief Operating Reactors Branch 81 Division of Operating Reactors I
V912l 00
Robert E. Uhrig Florida Power and Light Company 2 -
November 13, 1979 cc:
Mr. Robert Lowenstein, Esquire Lowenstein, Newman, Reis and Axelrad 1025 Connecticut
- Avenue, N.W.
Suite 1214 Washington, 0.
C.
20036 Environmental and Urban Affairs Library Florida Inter national University Miami, Florida 33199 Mr. Normari A. Coll, Esquire
- Steel, Hector and Oavis 1400 Southeast First National Bank Building fiiami, Florida 33131 Mr. Henry Yaeger, Plant Manager Turkey Point Plant Florida Power and Light Company P. 0.
Box 013100 Miami, Florida 33101 Mr. Jack Shreve.
Office of the Public Counsel Room 4, Holland Building Tallahassee, Florida 32304'
<gD,R BEGOT Wp.O Cy pO~
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO TURKEY POINT NUCLEAR GENERATING UNIT NOS.
3 AND 4 DOCKET NOS. 50-250 AND 50-251 Introduction By letter dated August 18, 1978, the Florida Power and Light Company (FPL) responded to our request for information dated August 14, 1978 concerning the possible use of atypical weld wire in the reactor vessels of the Turkey Point Plant, Unit Nos.
3 and 4.
Back round During 1978, BN initiated work contracted with the BN 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 (RV) 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, performed by the Mt. Vernon Works gaulity 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 MnMoNi: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 l.
Charpy V-notch tests on the atypical weld metal resulted in a higher than normal value of RTIIDT, 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 120oF 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.
Oiscussion 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'imits 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 (RTNOT) to increase with time.
The pressure-temperature operating limits must be modified. periodically to account for this radiation induced increase in RTNOT 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 radiation'amage, the stresses and stress intensities in the reactor vessel are maintained within acceptable limits.
I The magnitude of the shift in RTNpT is proportional to the neutron fluence that the materials are subjected to.
The shift in RTNOT can be predicted from Regulatory Guide 1.99.
To check the validity of the predicted shift in RT~OT, a reactor vessel material. surveillance program is required.
Sur-veillance specimens are'periodically removed from the vessel and tested.
The results of these tests are compared to the predicted shifts in RTNOT, and the pressure-temperature operating liIIiits are revised accordingly.
Since the unirradiated RTNDT of the atypical. weld metal was determined to be
- high, and it was assumed to be sensitive to radiation damage, the atypical 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 reflect the possibility that atypical material was used in their construction.
Evaluation To resolve the atypical weld issue, BEW has conducted an extensive investigation of records, metallographic examinations, chemical
- analyses, and fracture mechanics tests on both urrirradiated and irradiated atypical weld material.
The results of this program are presented in BAW-1556.
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.
The results of more than 2000 chemical analyses have been reviewed relative to the 42 wire heats.
All, except for the one batch of Crystal River survillance
- material, have been within the norma) ranges.
These include through-thickness tests from seven RYs fabricated at Mt. Vernon and tests of wire currently in inventory; Detailed metallographic examinations were performed on seven fractured Charpy specimens.
Both macro-and micro-examination techniques were employed, as well as
- a. fractographic examination with a scanning electron microscope.
Relatively littleporosity 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 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 quarrel-Ash emission spectrometer.
The concentrations of 10 elements were measured by this technique.
X-ray floures-cense analysis was used to measure the concentrations of nickel, molybdenum, and copper in irradiated Charpy specimens.
Results show that the copper content was high, averaging between Oe4 to 0.5X.
The chemistry of atypical material is compared to typical material in 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 and static, fracture toughness tests were conducted on one inch thick compact tension specimens at room temperature.
Although the dropweight NOT
's
-20oF, the results of the Charpy tests show that 50 ft-lbs of energy is absorbed at 150oF, therefore the unirradiated value of RTNDT is 90 F.
Using RTNDT equal to 90oF, the toughness properties obtained from the fracture mechanics tests, KIc (static) and KId (dynamic), are conservative (lie above) the KIc curve in ASME Code,Section XI and the KIR curve in ASNE Code,Section III respectively.
Using an RTNOT. of -20 F (the dropweight NOT), the fracture mechanics data fall within the scatter of data on normal material used to obtain the KIc and KIR curves.
This indicates that the RTNDT value of 90oF is conservative..
The effect of irradiation on the mechanical properties of atypical material have been evaIuated, using the test results on CrystaI River 3 survpllancg speci-mens.
These specimens were subjected to a fluence of 1.1 X 10 ~ n/cm~.
This fluence produced an increase in RTNDT of 35oF.
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 properties of atypical material should be considered.
As discussed
- above, we have determined that an initial value of RTNDT of 90oF is a very conservative value.
The increase in RTNDT due to iqradiatjon should be based on the measured
'alue of 35 F at a fluence of 1.1 X 10'8 n/cm" and the damage prediction slopes in,Regulatory Guide 1.99.
Me also recommend that the administratively applied pressure-temperature oper-ating limits. be removed from these 12 plants.
The operating limits in the Technical Specifications for Br owns Ferry 1 are presentTy 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 w'ould 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 imnediate 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 containing the atypical. weld metal are in the Crystal River 3 surveillance program..
One of these capsules has already been removed and tested.
-Also, there is enough atypical material in storage at 85/ to fabricate fracture toughness specimens up to 1.OT compact fracture toughness specimen size.
Oetailed information collected from our review, inclduing our calculations, are retained in our Engineering Branch files.
Date:
November 13, 1979
TABLE 1..
LOCATION OF POSSIBLE ATYPICAL MELDS PLANT BN Oconee 3
TMI 1
THI 2 ANO 1
Ni.dland 1
CR-3 Rancho Seco LOCATION OF MELD Center Circ. Beltline Upper Circ. Beltline Lower, Circ. Beltline Dutchman to Lowerhead Head to Flange and Nozzle to Shell Center Circ. Beltline Center Circ. Beltline Vertical Seam Beltline MESTINGHOUSE Zion 1
Zion 2 Turkey Pt.
4 Inter to Lower Circ. Beltline Vertical Seam Beltline
'o and 180o)
Nozzle Shell to Interm. Circ.
GE Br. Ferry 1
guad Cities 2 Shell to Flange and Longitudinal Meld in Beltline Closure Head to Flange
TABLE 2.
ATYPICAL MELD CHEMISTRY CR-3 Held 4
C Mn P
S
.08 1.65
.021
.013 Si Cr 1.0
.0l Ni Mo
.10
.45 Mn-Mo-Ni (Typical)
.08 1.6
.018
.015
.5
.07
.60
.40