ML20140B623
| ML20140B623 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 01/21/1986 |
| From: | Schnell D UNION ELECTRIC CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR ULNRC-1244, NUDOCS 8601270061 | |
| Download: ML20140B623 (9) | |
Text
-
aaa UNION ELECTRIC COM%NY 1901 Gratiot Street St. Louis January 21, 1986 Donald E Schnell Mce Presdent Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Denton:
ULNRC-124 4 DOCKET NUMBER 50-483 CALLAWAY PLANT PRESSURIZED THERMAL SHOCK
Reference:
NRC 10 CFR Part 50.61 titled " Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events" Per NRC requirements, Union Electric is transmitting the Pressurized Thermal Shock (PTS) study for the Callaway Reactor vessel.
Attachment I is a description of the study and a table listing the projected valtes of RT for the Callaway Reactor Vesselascalculatedbythemethodh$ovidedinthereference.
p The submittal of this study satisfies the requirements of 10 CFR 50.61.
If you have any questions or require additional information, please contact us.
Very truly yours, Donald F.
Schnell WEK/lw Attachment i
AD - J. Knight (Itr only) 1 B601270061 060 403 h. EB (BALLARD) l PDR ADOCK O PDR fr'* EICSB (ROSA)
P e
g PSB (GAMMILL) s RSB (BERLINGER)
FOB (BENAROYA)
Mailing A. ness: RO Box 149, St. Lou 6s MO 63166
__~_
i STATE OF MISSOURI )
)
SS l!
CITY OF ST. LOUIS )
,i i
e Donald F.
Schnell, of lawful age, being first duly sworn j
upon oath says that he is Vice President-Muclear and an officer of Union Electric Company; that he has read the foregoing document and i
knows the content thereof; that he has executed the same for and on behalf of said company with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.
I By Donald F. Schnell Vice President Nuclear SUBSCRIBED and sworn to before me this M/[ day of -
, 198 b.
i l
l AA' V~f (/ f BARBARA J. PFAFF NOTARY PUBUC, STATE OF Mis 100m MY COMmSSION EXPlRES APRit 22. 1989 ST. LOUIS COUNTY
}
i t
cc:
Gerald Charnoff, Esq.
Shaw, Pittman, Potts & Trowbridge 1800 M.
- Street, N.W.
Washington, D.C.
20036 Nicholas A.
Petrick Executive Director SNUPPS 5 Choke Cherry Road Rockville, Maryland 20850 G. C. Wright Division of Projects and Resident Programs, Chief, Section lA U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137 Bruce Little Callaway Resident Office U.S. Nuclear Regulatory Commission RR01 Steedman, Missouri 65077 Paul O'Connor Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 316 7920 Norfolk Avenue Bethesda, MD 20014 Ron Kucera, Deputy Director Department of Natural Resources P.O. Box 176 Jefferson City, Missouri 65102
l l
Attachment I RESPONSE TO NRC REQUIREMENT 10 CFR PART 50.61
" FRACTURE TOUGHNESS REQUIREMENTS FOR PROTECTION AGAINST PRESSURIZED THERMAL STHOCK EVENTS" I
January 21 1986
J
~
PROJECTED VALUES OF RT FOR PTS PTS SCREENING CRITERION A.
PURPOSE 1.
To determine projected values of reference nil ductility transition temperature (RT at the inner vessel surf ace of reactor vessel beltline materiNh)for the Callaway Nuclear Plant.
2.
To compare these values to the pressurized thermal shock (PTS) screening criteria as stated in NRC 10 CFR Part 50.61.
B.
BACKGROUND In the Safety Evaluation Report for Callaway (reference 4), the NRC identifled the Unresolved Safety Issue A-49 " Pressurized Thermal Shock."
The staff concluded that the Callaway Plant can be operated before complete resolution of this issue and completion of the proposed rule-making without undue risk to the health and safety of the public.
In July of 1985, the NRC issued the requirement 10 CFR 50.61 (reference 1) stating:
"For each pressurized water nuclear power reactor for which an operating license has been issued, the licensee shall face)ofreactorvesselbeltlNhm(aterialsbygivingvalues submit projected values of RT at the inner vessel sur-from the time of submittal to the expiration date of the operating license. The assessment must specify the bases for the projection, including the assumptions regarding core loading patterns. This assessment must be submitted by January 23, 1986, and must be updated whenever changes in core loadings, surveillance measurements, or other infor-mation indicate a significant change in projected values."
This report is written to specifically address 10 CFR 50.61 for the Callaway Nuclear Plant Unit 1.
Westinghouse has performed a number of programs for the Westinghouse Owners Group (WOG) to evaluate the impact of PTS on reactor vessel integrity.
In WCAP-10019 (reference 5), Westinghouse presented generic l
analyses of reactor vessel integrity during accident scenarios that have a potential to cause brittle fracture.
In a report to WOG in December 1982 (references 2 and 3), Westinghouse provided RT calcu-lations for the Callaway Plant utilizing the NRC-prescribed khods in SECY-82-465. WCAP-10319 (reference 6) described the probabilistic methodology developed by Westinghouse for treating the PTS issue in detail. This report is referencing the methodology provided in the Westinghouse reports as a basis for demonstrating the acceptability of the Callaway Plant RTNDT values at the end-of-life conditions contained herein.
O 2
t C.
METHODOLOGY l
1.
The value of RT at the inside surface of the reactor vessel for each materiN is the lower of the results given by Equations I and 2 (reference 1).
PTS = I + M + [-10 + 470 Cu + 350 Cu x Ni] f.270 0
f Equation 1: RT 0 l94 Equation 2: RTPTS
- I + H + 203 #
The definitions stated in 10 CFR Part 50.61 apply.
I j
2.
The core loading pattern assumed is the plant design basis or first j
core load for Callaway Unit 1.
The azimuthal peak, core mid-height (E > 1.0 MeV) value given in the FSAR Table 4.3-6 (reference 7) is used for all material as a conservative enveloping value. An addi-tional factor was added to account for uncertainties in the calculated i
}
ofLegendre'spolynomialisusedforthescatteringcros$ expansion value (20%), for the estimated increase in flux if the P section instead of the P1 expansion (20%), and for future core design 4
changes (20%). This factor is 1.73 (1.2 x 1.2 x 1.2).
It is antici-i pated that use of the new Optimized Fuel Assemblies at Callaway will
{
make the fluence value used even more conservative.
3.
Based on a calculated axial flux distribution, the initial reference j
temperature, and the material chemistry for the reactor vessel, plate i
R2708-1 of the lower vessel shell is the most limiting with respect to the screening criteria. Based on the calculated axial flux distribu-l tion, the nozzle to intermediate shell weld and the lower shell bottom weld are in an area of relatively low flux and do not need to be
]
compared to the screening criteria.
However, the nozzle to interme-diate shell weld is included to verify that this is not a concern.
Hence, the following vessel beltline materials were analyzed:
Plates Welds Intermediate shell R2707-1 Nozzle to intermediate shell
}
Intermediate shell R2707-2 Intermediate shell longitudinal Intermediate shell R2707-3 Intermediate to lower shell Lower shell R2708-1 Lower shell longitudinal I
Lower shell R2708-2 Surveillance capsule l
Lower shell R2708-3 1
i 4.
The PTS screening criteria are (reference 1):
270*F--for plate and axial (longitudinal) weld materials
}
300*F--for circumferential weld materials j
5.
In January 1986, the vessel at Callaway will have been in operation i
for approximately one (1) effective full power year (EFPY). Cal-culations were made for a forty (40) year vessel life assuming i
an eighty (80) percent use factor. Hence, calculations were made for one (1), three (3), fif teen (15), and thirty-two (32) EFPYs.
t I
4
c D.
DATA INPUTS 1.
The initial reference temperatures of the plate materials are actual values determined by the ASME Code Paragraph NB-2331 (reference 8) and the Materials Certification Reports. The initial reference temperatures of tne weld materials are actual values from the WOG Reactor Vessel Materials Data Base (reference 9).
2.
The weight percent of copper and nickel for the plate material is from the Materials Certification Reports.
The weight percent for the welds is from the WOG Data Base. The actual values for the Callaway welds were compared to the average of all the weld values in the data base with the same wire heat number. The larger of the two values was used for conservatism.
3.
Thefastneutronfluxabove1HeVisusedatall1gationgandis 9
i conservativ9fyestjmatedtobe1.73 times 2.08X10 n/cm -sec or 3.60 X10 n/cm -sec.
E.
CONCLUSION The results of the RT calculations for the Callaway Nuclear Plant are given in Table 1.
AlfThrojectedvaluesarewellbelowthecurrentNRC screening criteria for the full 40-year plant lifetime. Because of the conservatisms used in the calculations, it is not anticipated that addi-tional calculations will be required for the current licensed life of the plant.
i 1
I i
REFERENCES 1.
NRC 10 CFR Part 50.61 entitled, " Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events."
Westinghouse Owners Group Report, " Calculation of RT V
Westinghouse Domestic Operating Licensed Reactor VesI0Is,alues for 2.
" November 1982.
Westinghouse Owners Group Addendum, " Calculation of RT Values for 3.
Westinghouse Domestic Near Term Operating Licensed Rea$kdr Vessels,"
l December 1982.
4.
NUREG-0830 Supplement No. 3, Safety Evaluation Report Related to the Operation of Callaway Plant, Unit No. 1, May 1984.
5.
WCAP-10019, Summary Report on Reactor Vessel Integrity for Westinghouse Operating Plants, Figure 11.2-10, December 1981.
6.
WCAP-10319, A Generic Assessment of Significant Flaw Extension, including Stagnant Loop Considerations, From Pressurized Thermal Shock of Reactor Vessels on Westinghouse Nucleae Power Plants, December 1983.
I 7.
SNUPPS Final Safety Analysis Report, Copy S-3, Table 4.3-6, Rev. O.
8.
American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section !!!, Paragraph N8-2331.
9.
Westinghouse Owners Group Reactor Vessel Materials Data Base, January 198b.
1 i
TABLE 1 Naterial RT,7, (*F)
Component 1 Cu 1 Ni ' "RT 1 EFPY 3 FFPY 15 EfPY '32 Ef PY Intermediate Shell
.06
.58
+40'F 105 111 123 131 Plate R2707-1 Intermediate Shell
.07
.62
+10*F 79 86 102 112 Plate R2707-2 Intermediate Shell
.06
.62
-10*F 55 61 74 82 Plate R2707-3 Lower Shell Plate
.08
.60
+50*F 123 131 149 161 R2708-1 Lo*4er Shell Plate
.06
.56
+10*F 75 80 93 100 R2708-2 Lower Shell Plate
.08
.64
+20*F 93 102 121 132 R2708-3 Nozzle to Inter-0.037 0.087 -60'F
-7
-6
-2
+1 mediate Shell Weld Intermediate Shell 0.040 0.060 -60*F
-7
-5
-1
+2 Longitudinal Weld Intermediate to Lower 0.040 0.050 -60'F
-7
-b
-1
+1 Shell Weld Lower Shell 0.040 0.060 -60'F
-7
-5
-1
+2 Longitudinal Weld Surveillance Capsule 0.040 0.050 -60*F
-7
-5
-1
+1 Weld