ML20199F464
| ML20199F464 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 11/13/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20199F436 | List: |
| References | |
| NUDOCS 9711240223 | |
| Download: ML20199F464 (5) | |
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1 UNITED STATES s
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NUCLEAR REGULATORY COMMIS810N J'
WASHINGTON, D.C. m =1
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l
RELATED TO THE REACTOR PRES $URE VESSEL I
PRESSUR17ED THERMAL SHOCK ASSESSMENT DUQUESNE LIGHT COMPANY BEAVER VALLEY POWER STATION. UNIT 2 DdKET Np. 50-412
1.0 INTRODUCTION
By letter dated February 13, 1997, the Duquesne Light Company submitted WCAP-14784, Revision 2. " Evaluation of Pressurized Thermal Shock for Beaver Valley Unit 2," which provided the Beaver Valley Power Station, Unit No. 2 (BVPS-2) 10 CFR 50.61 pressurized thermal shock (PTS) assest'nent regarding the reactor vessel integrity.
This submittal was provided in response to Item 3 of an NRC request for additional information issued October 18, 1996, concerning WCAP-14484, " Analysis of Capsule V from the Duquesne Light Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program."
The PTS rule adopted on July 23, 1985, and revised on May 15, 1991, and December 19, 1995, established screening criteria that are a measure of a limiting level of reactor vessel material embrittlement beyond which operation cannot continue without further plant-specific evaluation.
The screening criteria are given in terms of reference temperature RT The screening criteria are 270'f for plates and axial welds and 300,F l'o,,r circumferential welds.
The RT,,, is defined as:
RT,,, - RT.,m + MT,,, + M where: (a) RT is the initial reference temperature, (b) MT is the mean value in the,akY!stment in reference temperature caused by irraM'ation, and (c) M is the margin to be added to cover uncertainties in the initial reference temperature, copper and nickel contents, fluence, and calculational procedures.
The initial reference temperature is the measured unirradiated value as defined in ASME Code, Paragraph NB-2331.
If a measured value is unavailable for the heat of material of interest, a generic value may be used. Generic values are based on the data for materials of all heats that were made by the same vendor using similar processes.
Generic values of initial reference temperature for welds are defined in the PTS rule.
ENCLOSURE 9711240223 971113 DR ADOCK 0500 4 2
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l The MT,,,d n1ckel in the material and is calculated as the product of a depends upon the amount of neutron irradiation and the amounts of copper an fluence factor and a chemistry factor (CF). The fluence factor is calculated i
f rom ~the best estimate neutron fluence at the clad-weld-metal interface on the inside surface of the vessel at the location where the akterial receives i
the highest fluence at the end of the period of evaluation. The CF may be i
determined using credible surveillance data or from the CF tables in the PTS 1
rule.
The CFs in the tables are dependent upon the best-estimate values of the amount of copper and nickel in the material. The term "best-estimate" is p
not well defined statistically, but has nomally been interpreted as the mean of the measured values.
The revised PTS rule contains criteria for determining whether surveillance data is credible.
The rule also contains the procedure for calculating the vessel weld CF from the adjusted or measured values of MT,,.
Specifically, the rule states that if there is clear evidence that the copper and nickel content of the surveillance weld differs from that of the vessel weld, the measured values of MT,, should be adjusted by mitiplying them by the ratio of the CF of the vessel weld to that of the surveillance weld.
The CF is calculated by multiplying each adjusted or measured value of MT, by its corresponding fluence factor, summing the products, and dividing y the sum of the squares of the fluence factors. The resulting CF will give t e t
relationship of MT to fluence that fits the plant surveillance data in such a way as to miNeize the sum of the squares of the errors.
The margin term is intended to account for variability in initial reference temperature and the adjustment in reference temperature caused by irradiation.
The value of the margin term is dependent upon whether the initial reference temperature was a measured or generic value and whether the adjustment in reference temperature was determined from credible surveillance data or from the CF-tables in the PTS rule.
2.0 NRC STAFF'S EVALUATION The BVPS-2 reactor vessel beltline includes the intermediate shell plates 89004-1 and 89004-2 (heat identification numbers C0544-1 and C0544-2, respectively), the lower shell plates 89005-1 and 89005-2 (heat identification numbers C1408-2 and C1408-1, respectively), the intermediate and lower shell i
vertical seams, and the girth weld between the intermediate and lower shell plates (all welds were fabricated using weld wire heat identification number 83642).
The material with the greatest amount of embrittlement (limiting material) is plate B9004-1.
The limiting plate in the BVPS-2 reactor vessel beltline, 89004-1, has an initial reference temperature value of 60'F. The licensee's best estimate values of the amount of copper and nickel in the limiting slate
'for BVPS-2 are 0.07%-and 0.55%, respectively. Linear interpolation of tie CFs in Table 1 of the PTS rule indicates that the CF-is 44 degrees F.
The best estimate values of copper and nickel are mean values from the plate certification test and the Combustion Engineering analysis from a sample obtained.at 1/4 thickness of the plate.
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- The licensee calculated the margin value for the limiting plate in accordance with the methodology in 10 CFR 50.61. 1he standard deviation of the initial reference temperature (RT is a measured value.
The standard deviation fo.,b.,) is zero since the RTeadjustmentinthereTr'encetempera un r
plate is 17 degreer F.
The licensee calculated a margin value of 34 degrees F.
This value is acceptable since it was calculated in accordance with the methodology in 10 CFR 50.61.
The RT,,,(EOL) is 149 degrees F.value calculated by the licensee for the limiting pla license The RT value calculated by the staff for BVPS-2 is 148.48 degrees F.
The staff',s,,value is calculated using (a) a measured value of the initfal reference temperature, (b) best-estimate value of copper and nickel for the vessel plate, (c) a CF determined from the CF '
table for plates in 10 CFR 50.61, (d) an EOL neutron fluence of 3.85E19n/cm,
and (e) a margin value of 34 degrees F.
The slight difference between the staff's and the licensee's RT values is due to round off errot (149 degrees F calculated by the licensee,a,n,d 148.48 degrees calculated by the staff).
The intermediate shell 11 ate, B9004-2, is the surveillance program base metal material for BVPS-2.
Te.e chemistry data for B9004-2 was obtained from mterial test certifications from the original fabrication.
Tho' staff W/Jependently verified that the surveillance' data for this base metal was credible.
This is explained in detail, below.
In addition, the staff verified that the RT,h,an the RT,,, value for the limiting plate.
value for the intermediate shell pla.te, B9004-2, is significantly lower t The chemistry data for the best-estimate chemistry for the beltline welds were obtained from their surveillance welds, material test certifications from the original fabriration, as well as the two additional material chemistry tests performed for St. Lucie Unit 2 and Almaraz Unit 2.
The BVPS-2 surveillance weld and the St. Lucie Unit 2 and Almaraz Unit 2 welds were made of the same weld wire heat identification number and flux type as the intermediate and lower shell vertical seams and the girth weld between the intermediate and lower shell plates for BVPS-2 (3/16-inch diameter weld wire type B-4, heat number 83642, Linde 0091 flux, lot number 3536).
The licensca revised the best-estimate chemistry for the beltline welds in WCAP-14784, Revision 2.
The licensee added two new data points of r.opper/ nickel values for the weld metal (0.04% Cu/0.06% Ni ano s.03% Cu/0.06%
N1) and removed the last data point of the weld metal copper / nickel values (0.04% Cu/0.07% NI) that had been previously reported.
T1e revised values were indicated in Attachment 2 of the Florida Power and Light Company letter, JPN-PSL-SESP-93-47 Revision 0, " Table 1: St. Lucie Unit 2 Reactor Vessel Beltline Weld Material."
As a result of the revised chemistry data, the licensee's best-estimate chemistry for the beltline welds and surveillance welds are 0.05% Cu and 0.07%
Ni. According to Regulatory Guide (RG) 1.99, Revision 2, the chemistry factor for welds with this chemistry is 34.1 degrees F.
The wald samples h:ve low Cu i
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. and Ni values and the RT, values for the welo metal is significantly lower than the RT,,, value for,t$e limiting plate. Therefore, the changes in the tu and Ni values are not significant in this assessment.
The Itcensee found the base metal and weld surveillance data to be credible.
However, the method of determining credibility did not satisfy Regulatory Guide 1.99, Revision 2, because the "least squares" method for curve fitting of the best-fit line did not go through the origin for the BVPS-2 surveillance data. As a result, the staff independently performed the "least squares"
> method for curve fitting with the best-fit line going through the origin.
Credibility Criterion (C) d ART,,, valu(c)(2)(1)be less than 17 degroes F for in section of 10 CFR 50.61 indicates that the scatter of the measure es must base metals and 28 degrees F for welds.
Evaluation of this critarion was the basis for the licensee's determination that the BVPS-2 base metal and weld surveillance data met the credibility criteria in 10 CFR 50.61.
As stated above, the staff inde)endently evaluated the scatter of the measured ART,,,,
and determined that tle base metal and weld surveillance data satisfy Criterion (C)lculated CFs are in agreement to the staff's calculated CFs in section (c)(2)(1) of 10 CFR 50.61.
In addition, the licensee's ca (approximat.ely 35.3 degrees F for plates and 15.2 degrees F for welds).
Hence, the surveillance data are credible and should be used to determine the CF for the base metal and vessel weld.
The licensee's chemistry values for the weld heat identification number 83642 are 0.05% Cu and 0.07% Ni, The chemistry values for this heat in the Combustion Engineering Owners Group Final Report, CE NPSD-1039, Revision 02, are 0.046% Cu and 0.086% N1.
Because the difference in the chemistry values is small, the change in the Ri,, value for the subject weld is not p
significant.
The staff verified that all of the beltline region materials in the BVPS-2 reactor vessel have E0L RT,,, values well below 149 degrees F at E0L (32 EFPY).
3.0 CONCLUSION
The licensee and staff assessments indicate that the reactor pressure vessel would be below the PTS screening criteria at the expiration of the BVPS-2 facility operating license.
Principal Contributor:
M. Khanna Date:
November 13, 1997
4.0 REFERENCES
1.
Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials, Revision 2, May 1988.
2.
NUREG-0800, Standard Review Plan, Section 5.3.2:
Pressure-Temperature Limits.
3.
Code of Federal Regulations Title 10, Part 50, Appendix G, Fracture Toughness Requirements.
4.
Generic Letter 68-11, NRC Position on "adiation Embrittlement of Reactor Vessel Waterials and its Impact on P',ent Operations, July 12, 1988.
5.
ASME Boiler and Pressure Vessel Code, Section !!I, Appendix G for Nuclear Power Plant Components, Division 1
" Protection Against Non-ductile Failure."
6.
WCAP-9615 Revision 1, 'Duquesne Light Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program," S. E. Yanichko, et al.,
September 1989.
7.
WCAP-14484, " Analysis of Capsule V from the Duquesne Light Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program,"
P. A. Grendys, et al., February 1996.
- 8. to JPN-PSL-SESP-93-47 Revision 0, " Table 1:
St. Lucie Unit 2 Reactor Vessel Beltline Weld Material."
9.
Letter Report MT/ SMART-210(88), " Response to U.S. Nuclear Regulatory Commission Generic Letter 88-11 for the Beaver Valley Unit 2 Reactor Vessel," N. K. Ray, et al., November 1988.
- 10. WCAP-9228, " Central Nuclear de Almaraz, Almaraz Unit No. 2 Reactor Vessel i
Radiation Surveillance Program," P. J. Fields, et al., December 1977.
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