ML17229A522
| ML17229A522 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 10/27/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17229A521 | List: |
| References | |
| NUDOCS 9711130317 | |
| Download: ML17229A522 (42) | |
Text
~R Rfgg
+
0 Cy ClO IVl0 Vg
~0
++*++
0
~
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON> D.C. 2055&0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION PRESSURIZED THERMAL SHOCK EVALUATION FLORIDA POWER AND LIGHT COMPANY ST.
LUCIE UNITS 1
AND 2 DOCKET NOS. 50-335 AND 50-389
- 1. 0 INTRODUCTION By letter dated May 14, 1996, the licensee submitted a pressurized thermal shock (PTS) evaluation for St. Lucie Units 1 and 2.
In addition, proprietary and non-proprietary copies of the Combustion Engineering Owners'roup (CEOG) report CEN-405-P, Revision 2, "Application of Reactor Vessel Surveillance Data for Embrittlement Management" were enclosed for NRC review and approval.
By teleconference conducted on August 27, 1996, the Nuclear Regulatory Commission (NRC) staff suggested that the licensee submit new proprietary and non-proprietary reports since data extracted from the power reactor embrittlement database (PR-EDB) are non-proprietary.
New proprietary and non-proprietary versions of CEN-405-P (designated Revision 3) were submitted by letter dated September 23, 1996.
Additional information was provided by letters dated January 14 and May 16, 1997.
The January 14, 1997 letter changed the requested approval date from April 1, 1997 to April 1, 1998 due to scheduling outage changes.
It should be noted that CEN-405-P, Revision 1, was originally submitted to the NRC on December 6,
1991.
On January 29, 1992, the NRC staff issued a request for additional information (RAI).
CEN-405-P, Revision 2, incorporated changes made as a result of the RAI, and was submitted for review and approval on August 6, 1993.
Review of the topical report was given very low priority due to staff resources.
Revision 2 was resubmitted as an attachment to the St.
Lucie PTS evaluation on May 14,
- 1996, and therefore, had a higher priority for review.
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>>s.
The screening 97iii303i7 97i027 PDR ADGCK 05000335 P
PDR Enclosure
criteria are 270'F for plates and axial welds and 300'F for circumferential welds.
The RT>>s is defined as:
RT>>s RTMor<u) + MT>>s + "
where:
(a)
RT, <<> is the initial reference temperature, (b) ZBT,<, is the mean value in the aJgustment in reference temperature caused by irrad>ation, and (c)
H 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 the American Society of Hechanical Engineers (ASHE) Code, Paragraph NB-2331. If measured values are unavailable for the heat of material of
- interest, generic values may be used.
The generic values are based on the data for materials of all heats that were made by the same vendor using similar processes.
The generic values of initial reference temperature for welds are defined in the PTS rule.
The MT>>, depends upon the amount of neutron'irradiation and the amounts of copper and nickel in the material and is calculated as the product of a fluence factor and a chemistry factor (CF).
The fluence factor is calculated from the best estimate neutron fluence at the clad-weld-metal interface on the inside surface of the vessel at the location where the material receives the highest fluence at the end of the period of evaluation.
The CF may be determined using credible surveillance data or from the CF tables in the PTS 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 not well defined statistically, but has normally been interpreted as the mean of the measured values.
The revised PTS rule contains criteria for determining whether surveillance data are credible.
The rule also contains the procedure for calculating the vessel weld CF from the adjusted or measured values of ZBT>>s 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 multiplying them by the ratio of the CF of the vessels 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 by the sum of the squares of the fluence factors.
The resulting CF will give the relationship of ZBT>>> to fluence that fits the plant surveillance data in such a way as to minimize the sum of the squares of the errors.
The margin term is intended to account for variability in initial reference'emperature 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 DISCUSSION 2.1 St. Lucie Unit 1
The St. Lucie Unit 1 (SL-1) reactor vessel beltline includes the intermediate shell plates C-7-1, C-7-2 -and C-7-3, heats A4567-1, B9427-1 and A4567-2 respectively; lower shell plates C-.8-1, C-8-2 and C.-8-3, heats C5935-1, C5935-2 and C5935-3 respectively; intermediate shell axial welds 2-203 A, B, C, heats A8746/34B009; intermediate to lower shell girth welds 9-203, heat 90136 and the material with the greatest amount of embrittlement (limiting material) is the lower shell axial welds 3-203 A, B, C.
The axial weld was fabricated by Combustion Engineering (CE) using the submerged arc weld process with weld wire heat 305424 and Linde 1092 flux, lot number 3889.
Surveillance data for the limiting material are not available in the SL-1 surveillance
- program, however, the data are available in the Beaver Valley Unit 1 (BV-1) surveillance program.
The BV-1 vessel and the surveillance weld were fabricated by CE and designed by Westinghouse (W).
The surveillance weld was fabricated using the submerged arc weld process with weld wire heat 305424 and Linde 1092 flux, lot number 3889 which is the same process that was used to fabricate the SL-1 lower shell axial welds 3-203 A, B, C (the limiting material).
In addition, the BV-1 and SL-1 vessels were both fabricated during the same period by CE in Chattanooga, Tennessee.
The staff evaluated the applicability of the BV-1 surveillance data to the SL-1 vessel in terms of similarity of the irradiation environments.
The staff reviewed the additional information that was provided by letter dated January 14, 1997, which included the displacement rate (dpa/sec) and flux values for the last surveillance capsules removed from both BV-1 and SL-l.
The time weighted average BV-1 and SL-1 cold leg inlet temperatures are 544.7'F and 546.7'F, respectively.
Since the BV-1 surveillance capsules were irradiated at a similar, but slightly lower cold leg temperature as compared to the SL-1 cold leg temperature, the BV-1 surveillance data do not require any temperature correction for use.
In addition, comparison of the ratio of displacement rate to flux,-or the damage ratio, is an indication of differences in the energy distribution of neutrons at the surveillance capsule locations.
When comparing the damage
-ratios of the SL-1 vessel at the critical weld location and the BV-1 and SL-1
surveillance
- capsules, the irradiation behaviors are similar within 9X.
Therefore, in terms of irradiation environment, the BV-1 surveillance data are appropriately applicable to the SL-1 reactor vessel data.
In addition, the licensee provided a statistical analysis in support of the conclusion that no significant bias exists between surveillance data from CE and W designed vessels that were fabricated in the CE Chattanooga facility.
The staff completed an independent statistical analysis to verify the licensee's conclusions.
The results of the staff's analysis are detailed in the statistical analysis section of this SE.
Therefore, based on statistical
- analysis, the BV-1 surveillance weld is considered to be representative of the SL-1 lower shell axial seam welds 3-203 A, B, C.
2.2 St. Lucie Unit 2 The St. Lucie Unit 2 (SL-2) reactor vessel beltline includes. the intermediate shell plates M-605-1 and H-605-3, heats A-8490-2 and A-8490-1 respectively; lower shell plates H-4116-1, H-4116-2 and M-4116-3, heats B-8307-2, A-3131-1 and A3131-2 respectively; intermediate shell axial welds 101-124 A, B, C and 101-124C, heats 83642 and 83637 respectively; intermediate to lower shell girth welds 101-171, heats 3P7317 and 83637 respectively; and lower shell axial welds 101-142 A, B, C:
The material with the greatest amount of embrittlement (limiting material) is the intermediate shell plate H-605-2, heat B-3416-2.
It should be noted that the SL-2 reactor vessel has low cop[er and nickel content, and the limitinp plate has an RT>>s value that is 110 F below the screening criterion of 270 F.
Sufficient surveillance capsule data are not yet available for SL-2, so projections of RT>>, at expiration-of-license (EOL) were made based on initial chemistry values and the projection methodology of 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events."
- 3. 0 STATISTICAL ANALYSIS FOR ST.
LUCI E UNIT 1
As mentioned, CEN-405-P, Revisions 1 and 2, had previously been submitted to the NRC in 1991 and 1993 respectively.
Discrepancies between the licensee's database and the data that the staff identified were outlined in the January 29; 1992 staff RAI.
The main source of the staff's data was the Oak Ridge National Laboratory (ORNL) Report NUREG/CR 4816, "PR-EDB:
Power Reactor Embrittlement Database, Version 2."
The licensee also utilized data from the ORNL report except in cases where data had been omitted, or incorrect data were identified.
Missing values were corrected when additional sources of the data were available.
When incorrect data values were identified, each case was evaluated using original source documents to determine the appropriate
~
value.
Therefore, the licensee's current database represents updated
- values, and the data were utilized in the statistical analysis.
For completeness, analyses were done for both the licensee's reported values (designated as "GEOG data" in the figures and tables),
and the values that were outlined in the staff's 1992 RAI (designated as "NRC data" in Table 2) in
.order to compare the results.
The staff performed the statistical Mann-Whitney Test of Independence on the data provided in Tables A-1 and A-2 of the CEOG Report CEN-405-P, Revision 3, "Application of Reactor Vessel Surveillance Data For Embrittlement Management, September-1996."
The Mann-Whitney Test is a non-parametric test in which the data from two populations are combined, and arranged in ascending order.
The U statistics are determined for each sample population.
Each member of a given population is assigned the number of members from the other population which precede it in the ordered list.
The U statistic is the summation of the numbers assigned to the members of a given population.
The Z statistic, also known as the standardized sampling error, is generated from 1) the calculated U value,
- 2) the product of the number of members in each population divided by two (e.g. n,nz/2),
and 3) the standard deviation of the U statistic.
Table 1 lists data including the predicted minus actual (P-A) values of the increase in the nil ductility transition temperature caused by neutron irradiation (MT >>).
The data are from plate and weld surveillance material that was irradiated in CE and W designed, CE fabricated vessels.
The P-A values for the CE and W data were the parameters used for the staff's and the licensee's statistical analyses.
Additional data with regard to plate orientation are also provided.
Figure 1 shows a schematic of the method used to separate the data according to 1) material type (plate or weld), 2) vessel designer (CE or W), 3) orientation (LT or TL plates only),
and 4) data set (GEOG or NRC) for the statistical analyses.
Mann-Whitney tests were performed on CEOG plate data based on vessel designer and plate orientation.
GEOG weld data were subjected to the Mann-Whitney test with population differentiation based on vessel designer.
NRC values for both plates and welds were also tested.
A total of six (6) tests were performed on the data.
The null hypothesis tested was that there is no difference between the two population distributions that were tested in each permutation.
The null hypothesis was accepted in each case based on the stated decision criteria given in Table 2.
The table summarizes the resul'ts of the Mann-Whitney statistical tests.
Figures 2-7, 9 and 10 present the histograms of the CEOG CE and W plate and weld data.
These figures indicate the frequency of a given value as well as the normal probability distribution based on the calculated mean and standard deviation.
Figures 8 and 11 present the combined CEOG CE fabricated, CE and W
designed plate and weld data, respectively.
Based on the statistical
- analysis, the staff concluded that there is no significant difference or bias between the CE fabricated, CE and W designed surveillance data.
Therefore, surveillance data from CE fabricated, CE and W
designed vessels will, on average, be representative of each other for vessels fabricated in the CE Chattanooga facility.
4.0 INITIAL REFERENCE TEMPERATURE As part of the PTS evaluation, the staff reviewed the basis for the initial reference temperature values for all SL-1 and 2 beltline materials'he results of the review are discussed below.
4.1 St. Lucie Unit 1
The limiting weld in the SL-1 reactor vessel beltline was fabricated from the
'same heat of weld wire (305424) as the BV-1 surveillance weld and wel.ds in the LaSalle 1 (LS-1) reactor vessel beltline.
A full Charpy curve was produced as part of the initial property testing for the BV-1 surveillance program.
In addition, three Charpy tests were performed at +10'F for weld heat 305424 with the same Linde 1092 flux used to fabricate the SL-1 and LS-1 vessel beltline welds.
The three Charpy test data results of 82, 87, and 92 ft-lbs at +10'F were reported for both the SL-1 and LS-1 vessel beltline welds.
Initially, the licensee only used the BV-1 data to conclude that the RT ><<
value for the limiting weld is drop weight controlled.
The staff verif~fed that the drop weight temperature remains.controlling with the inclusion of Charpy data from SL-1 and LS-1.
The resulting initial reference temperature value for the limiting axial welds is -60'F.
4.2 St. Lucie Unit 2 The licensee used a plant specific initial reference temperature value of
+10'F for the limiting plate.
The licensee also reported a plant specific RTgpp
) value of -80'F for the intermediate shell axial welds 101-124 A,B,C in Ne SL-2 reactor vessel.
These welds were fabricated using weld wire heat 83642.
Beaver Valley 2 reported a value of -30'F for a weld that was also fabricated from heat 83642.
The original submittal stated that an RT>>,<
> value of -80'F would be used for the SL-2 weld.
The justification was that these welds are not the limiting
- material, and the RT>>, value is significantly below the screening criterion.
Since there is such a large difference in the two values, the staff issued an RAI by letter dated March 13, 1997.
In response to the RAI, the licensee committed to use the generic value of -56'F with a larger margin term instead of the non-conservative value of -80'F for calculation of the RTp7$ value.
~
~
5.0 BEST-ESTIMATE CHEMICAL COMPOSITION OF THE LIMITING MATERIAL 5.1 St. Lucie Unit 1
The licensee's best-estimate values of the amount of copper and nickel in the limiting"'weld for SL-1 are 0.28X and 0.63X, respectively.
Linear interpolation of the CFs in Table 1 of the PTS rule indicates that the chemistry factor is 191.65'F for welds with these amounts of copper and nickel.
The best estimate values of copper and nickel are mean values of weld deposit data from the CE weld metal qualification (WMg) test and the BV-1 surveillance weld.
'5.2 St. Lucie Unit 2 The licensee's best-estimate values of the amount of copper and nickel in the limiting plate for SL-2 are 0. 13X and 0.62X, respectively.
Linear interpolation of the CFs in Table 1 of the PTS rule indicates that the chemistry factor is 91.5'F for plates with these amounts of copper and nickel.
The best estimate values of copper and nickel are mean values from the plate certification test.
6.0 BEST-ESTIMATE CHEMICAL COMPOSITION OF THE BEAVER VALLEY 1 SURVEILLANCE WELD The licensee's best-estimate of the amount of copper and nickel in the BV-1 surveillance weld are 0.26X and 0.62X, respectively.
Linear interpolation of the CFs in Table 1 of the PTS rule indicates that the chemistry factor is 183.2 F for welds with these amounts of copper and nickel.
The best-estimate of the amount of copper and nickel in the surveillance weld is the mean value of the measurements of these elements from the surveillance weld itself.
7.0 EVALUATION OF SURVEILLANCE DATA The licensee determined the CF for the SL-1 vessel weld using:
(a) the BV-1 surveillance data, (b) the ratio procedure that is recommended in 10 CFR 50'.61 when the chemistry of the surveillance weld is different than the vessel
- weld, and (c) the calculational procedures that are recommended in 10 CFR 50.61.
The best-estimate chemistry of the SL'-1 vessel weld is 0.28X copper.
and 0.63%
The best-estimate chemistry of the BV-1 surveillance weld is 0.26X copper and 0.62X nickel.
The licensee's estimate of the ratio of the CF of the vessel weld to the CF of the surveillance weld was 1.046.
The CF calculated by the licensee was 200. 15 F.
The staff determined the CF for the vessel weld using its best-estimate chemistry for the vessel weld (0.28X copper and 0.63X nickel) and the surveillance weld that was discussed above.
The ratio of the CF of the vessel weld to the CF of the surveillance weld was 1.046.
The CF calculated by the staff was 200.1'F.
'Credibility Criterion (C) in section (c) (2)(i) of 10 CFR 50.61 indicates that the scatter of the measured MT >> values must be less than 17'F for base metal and 28'F for welds.
The licensee determined that the scatter for the BV-'
surveillance weld data is less than 28'F.
Evaluation of this criterion was the basis for the licensee's determination that the BV-1 weld surveillance data met the credibility criteria in 10 CFR 50.61.
The licensee proposed that the calculated CF from the surveillance data (200. 15'F) be used in determination of MT>>$ and RT>>$.
The staff independently evaluated the scatter of the measured ZBT>>$
and determined that the weld surveillance data satisfy Criterion (C) in section (c)(2)(i) of 10 CFR 50.61.
,Hence, the surveillance data are credible and should be used to determine the CF for the vessel weld.
8.0 MARGIN VALUE 8.1 St. Lucie Unit 1
The licensee calculated the margin value in accordance with the methodology in 10 CFR 50.61.
A standard deviation of zero was used for the initial reference temperature (RT>>,
) since, as discussed in section 4, the RT+7( ) is a
measured value. 5 CFR 50.61 recommends that the standard devsa/ion for the adjustment in reference temperature be reduced by half if surveillance.data are credible.
Therefore, a standard deviation of 14'F was used since the surveillance data for the weld were found to be credible.
The licensee calculated a margin value of 28'F.
This value is acceptable since it was calculated in accordance with the methodology in 10 CFR 50.61.
8.2 St.
Lucie Unit 2 The licensee calculated the margin value in accordance with the methodology in 10 CFR 50.61.
A standard deviation of zero was used for the initial reference temperature (RT<, <<>) since the RT>><
> is a measured value.
A standard deviation of 17 V was used for the adjustment in reference temperature for the plate.
The licensee calculated a margin value of 34'F.
This value is acceptable since it was calculated in accordance with the methodology in 10 CFR 50.61.
9.0 PROJECTED RT 7$ VALUE AT EOL
- 9. 1 St. Lucie Unit 1
The RT>>, value calculated by the licensee at EOL is 213'F.
The RT>>$ value calculated by the staff for SL-1 is 212.6'F.
The staff's value is calculated using (a) a measured value of the initial reference temperature, (b) best-estimate values of copper and nickel for the vessel and surveillance welds, (c) a CF calculated from BV-1 surveillance data and adjusted to account for the difference between the best-estimate chemistry of the SL-1 vessel and BV-1
~I 1'
4 1 4
surveillance
- weld, (d) an EOL neutron fluence of 2.27E19n/cm, and (e) a margin value of 28'F.
The slight difference between the staff's and the licensee's RT>>, values is due to round off error.
(213'F calculated by the licensee and 212.6'F calculated by the staff).
Using the BV-1 weld surveillance data for the SL-1 PTS evaluation, indicates that the reactor pressure vessel would be below the PTS screening criteria at the expiration of its license.
9.2 St. Lucie Unit 2 The RT>>, value calculated by the licensee at EOL is 160'F.
The RT>>, value calculated by the staff for SL-2 is 160.3'F.
The staff's value is calculated using (a) a measured value of the initial reference temperature, (b) best-estimate values 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 2.76E19n/cm, and (e) a margin value of 34'F.
The slight difference between the NRC staff's and the licensee's RT>>, values is due to round-off error (160'F calculated by the licensee and 160.3'F calculated by the staff).
The SL-2 reactor pressure vessel PTS evaluation indicates that the reactor pressure vessel would be below the PTS screening criteria at the expiration of its license.
10.0 CEOG re ort CEN-405-P Revision 3
As mentioned, the GEOG submitted report CEN-405-P, Revision 3, "Application of Reactor Vessel Surveillance Data for Embrittlement Management" for review and approval as part of their St. Lucie PTS submittal.
The report presents two approaches for CE owners to apply Regulatory Position 2. 1 of Regulatory Guide (RG) 1.99, Revision 2, when the limi'ting material of the vessel is not in the surveillance
- program, and the surveillance data meet the remaining four RG credibility criteria.
The integrated surveillance approach would use limiting material data from another CE fabricated host vessel after determining the similarity of that vessel to the subject vessel (i.e., similarity of irradiation environment).
The margin reduction approach would use the plant specific surveillance data to reduce the margin to be added to the predicted shift.
The current PTS rule incorporates the five surveillance data credibility criteria of RG 1.99, Revision 2.
The first credibility criterion states that "Materials in the capsules should be those most likely to be controlling with regard to radiation embrittlement according to the recommendations of
[RG 1.99, Revision 2]."
The licensee's approach proposes that the credibility of-the surveillance data be determined using only four of the criteria.
Application of the margin reduction approach would require an exemption from the PTS rule.
0 The integrated surveillance
- approach, would be allowed by the PTS rule.
In order to apply this approach a licensee would need to confirm that the material in the host surveillance program is equivalent to the controlling material in their vess'el.
This method involves several plant specific considerations.
Therefore, approval of a generic topical report for a method that would need to be reviewed on a case-by-case basis could lead to situati'ons where licensees may not be able to effectively reference the report.
Therefore, the staff denies approval of generic topical report CEN-405-P, Revision 3.
11.0 CONCLUSION
(a)
The BV-1 weld surveillance data meet the credibility criteria in 10 CFR 50.61.
The weld data was determined to be acceptable for use in the SL-1 PTS evaluation by comparison of the irradiation environments and by statistical analysis.
(b)
Specifically, since the BV-1 weld surveillance data meet the credibility criteria.of 10 CFR 50.61, the data were used to determine the CF for the limiting SL-1 vessel weld.
(c)
The licensee's and staff's calculated values of RTpyp.for SL-1 at expiration of license (213'F) is well below the 270 F screening criterion specified in 10 CFR 50.61 for axial welds.
(d)
The licensee's and staff's calculated values of RT>>> for SL-2 at, expiration of license (160'F) is well below the 300 F screening criterion specified in 10 CFR 50.61 for plates.
(e)
Since the conclusions in (c) and (d) are dependent upon the available chemistry and surveillance data, they are subject to change when new data become available.
It should also be noted that the licensee for SL-1 must track and assess any changes in the BV-1 data that would effect the SL-1 PTS evaluation.
The NRC staff reserves the right to request a,written assessment of the impact of changes (if any) to the SL-1 PTS evaluation that result from changes in the BV-1 data.
(f)
The NRC staff denies approval of GEOG report CEN-405-P, Revision 3
"Application of Reactor Vessel Surveillance Data for Embrittlement management" since
- 1) application of the margin reduction approach would require an exemption from the PTS rule, and
- 2) application of the integrated surveillance approach would need to be reviewed on a
case-by-case basis.
Attachment:
Embrittlement Database (Extract)
Principal Contributor:
Andrea Lee Date:
October 27, 1997 References 2.
Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Haterials," Revision 2, Hay 1988.
Code of Federal Regulations, Title 10, Part 50, Section 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events."
3.
ASHE Boiler and Pressure Vessel Code,Section III, Appendix G for Nuclear Power Plant Components, Division 1, "Protection Against Nonductile Failure'."
5.
6.
7.
8.
9.
10.
Hay 16, 1997, Letter from J.
A. Stall to USNRC Document Control Desk,
Subject:
St. Lucie Units 1 and 2 Request for Additional Information-
Response
10 CFR 50.61 Pressurized Thermal Shock Evaluation.
January 14, 1997, Letter from J.
A. Stall to USNRC
.Document Control Desk
Subject:
St.
Lucie Units 1 and 2 Request for Additional Information (RAI) Response 10 CFR 50.61 Pressurized Thermal Shock Evaluation.
September 23, 1996 Letter from J.
A. Stall to USNRC Document Control Desk,
Subject:
St. Lucie Units 1 and 2
10 CFR 50.61 - Evaluation of Pressurized Thermal Shock of Reactor Vessel Beltline Haterials-Supplement.
November 15, 1993, Letter from D. A. Sager to USNRC Document Controls Desk,
Subject:
Forwards
Response
to Request for Additional Information Re: Generic Letter, 92-01, Revision 1, "Including New Hean Chemistry Values for Unit 1 Lower Longitudinal Welds."
August 6, 1993, Letter from R.
F. Burski to USNRC Document Control Desk
Subject:
C-E Owners Group Submittal of CEN-405-P, Revision 2, "Application of Reactor Vessel Surveillance Data for Embrittlement Hanagement."
(Enclosure II was formal response to NRC staff, request for additional information.)
D. Lurie and R.
H. Hoore,
- 1994, NUREG-1475: "Applying Statistics, "U.S.
Government Printing Office.
R. Parsons, 1978, "Statistical Analysis," Second Edition, Harper 8
- Row, Publishers, New York, NY.
J. Neter, W. Wasserman, G. A. Whitmore,
- 1978, "Applied Statistics,"
Allyn and Bacon, Inc., Boston, HA.
TABLEI tPAGES OF4)
Plate or Weld Designer Capsule ID Heat ID Orientation From PR-EBD Cu%
NBo Fluence E18 n/crr82 FIvx E10 n/cm"2sec Predicted'hift eF P-A 0
P P
P P
P P
P P
P P
P P
P P
P P
P P
P P
P P
P P.
P P
P P
P P
P P
P P
P P
CE CE CE CE CE CE CE CE CE CE CE CE CE CE CE CE CE CE CE CE W
W W
W W
W W
W W
ANO-2 ANO-2 CC1 CC1 CC2 CC2 FT. CAL FT. CAL FT. CAL FT. CAL MILSTON2.
MILSTON2 MAINEY MAINEY MAINEY MAINEY MAINEY MAINEY PALISAO PAUSAD PALISAO PALISAO ST. LUC 1 ST. LUC 1 ST. LUG2 ST. LUC2 SONGS 2 SONGS2 BV1 BV1 BV1 BV1 BV)
BV1 COOK1 COOK 1 COOK 1 W-97 W-97 W-263 W-263 W-283 W-283 W-225 W-225 W-265 W-265 W-97 W-97 A-25 A-25 A-35 A45 W-263 W-263 A-240
~ A-240 W-290 W-290 W-97 W-97 W-83 W43 W-97 W-97 0
tj V
V W
W T
T T
PAN201 PAN201 PCC103 SHSS01 PCC202 SHSS01 PFC101 SHSS01 PFC101 PFC101 PML201 PML201 PMY01 SHSS01 PMY01 PMYQ1 PMY01 PNIY01 PPAL01 PPAL01 PPAL01 PPAL01 PSL101 PSL1Q1 PSL201 PSL201 PS0201 PS 0201 PBV101 PBV101 PBV101 PBV101 PBV101 PBV101 PCK101 PCK10)
SHSS02 LT TL LT LT LT LT LT LT LT TL LT TL LT LT LT TL LT TL LT TL LT TL LT TL LT Tl.
LT TL'T TL LT TL LT Tl.
LT TL LT 0.08 0.08 0.12 0.18 0.14 0.18 0.10 0.18 0.10 0.10 0.14 0.14 0.15 0.18 0.15 0.15 0.15 0.15 0.25 0.25 0.25 0.25 0.15 0.15 0.11 0.11 0.10 0.10 0.20 D.20 0.20 0.20 0.20 0.20 0.14 0.14 0.14 0.60 0.60 0.64 0.66 0.66 0.66 0.48 0.66 0.48 0.48 0.61 0.61 0.59 0.66 0.59 0.59 0.59 D.59 0.53 0.53 0.53 0.53 0.57 Q.57 0.61 0.61 0.60 0.60 0.54 0.54 0.54 054 0.54 0.54 0.49 0.49 0.68 3.41 341 6.00 5.90 8.06 8.14 5.83 5.83 8.30 8.70 3.75
=
3.67 17.60 17.60 77.30 77.30 5.67 5.67 60.60 60.60 11.00 11.30 5.40 SAO 1.62 1.63 5.07 5.07 6.54 6.54 2.91 2.91 9.49 9.49 2.71 2.71
'.71 6.39 6.24 6.47 6.36 5.58 5.64 5.60 S.BQ 4.78 S.Q1 3.98 3.87 43.00 43.00 B1.40 B1.40 4.70 4.70 B2.00 62.00 7.01 7.20 3.B7 3.67 4.60 4.63 4.80 4.80 5.79 5.79 7.92 7.92 5.11 5.11
.6.79 6.79 6.79 21 50 6Q 88 84 128 60 124 74 70 70 96 120
'l50 185 195 97 93 205 205 175 155 68 70 35 21 55 35 120 135 130 140 150 185 60 70
'0 15
-14 12 29 12 3
-24 7
8
~22 a2 0
34 34
-5 17 18
<<2 18 6
-9
-35 QS
-9 2
6 (TABLE1 CONTINUED)
~l
~I ~
C
~ i
~
~
P P
P P
P P,.
P P
P P
P P
P P
P P
P P
P P
P P
P P
P P
P P
P P
P P
P P
P P
P P
P P
P W
W W
W W
W W
W W
W W
W W
W W
W W
W W
W W
W W
W W
W W
W W
W W
W W
,W W
W W
W COOK 1 COOK 1 COOK 1 CALLA1 CALLA1 HADNFC HAD NEC HADNEC HADNEC DIAB 1 DIAB1 DIAB2 DIAB2 FARLEY1 FARLEY 1 FARLEY 1 FARLEY 1.
FARLEY 1 FARLEY 1 FARLEY2 FARLEY2 FARLEY2 FARLEY2 ROBIN 2 ROBIN 2 ROBIN 2 ROBIN 2 ROBIN 2 ROBIN 2 ROBIN 2 ROBIN 2 IP2
. IP2 IP3 IP3 IP3 IP3 IP3 IP3 IP3 IP3 KEWAUN Y
Y Y
U U
A D
F H
S S
U U
U U
X X
Y Y
U U
W W
S S
S S
T T
V Y
Y T
T Y
Y Z
Z Z
R PCK101 PCK101 SHSS02 PCL101 PCL101 SASTM SASTM SASTM SASTM PDC103 SSHS02.
PDC201 PDC201 PFA101 PFA101 PFA101 PFA101 PFA101 PFA101 PFA201 PFA201 PFA201 PFA201 PHB201 PHB202 PHB203 SASTM PHB203 SASTM PHB202 SASTM PIP203 SASTM PIP301 PIP304 PIP304 PIP304 SHSS02 PIP303 PIP304 PIP304 SHSS02 LT TL LT LT TL LT LT LT LT LT LT LT TL LT TL LT TL LT TL LT TL LT LT LT LT LT LT LT LT TL LT LT LT 0.14 0.14 0.14 0.07 0.07 0.20 0.20 0.20 0.20 0.077 0.14 0.15 0.15 0.14 0.14 0.14 0.14 0.14 0.14 0.20 0,20 0.20 0.20 0.12 0.10 0.09 0.20 0.09 0.20 0.10 0.20 0.14 0.20 0.18 0.24 0.24 0.24 0.14 0.19 0.24 0.24 0.14 0.49 0.49 0.68 0.59 0.59 0.18 0.18 0.18 0.18 0.46 0.68 0.67 0.67 0.55 0.55 0.55 0.55 0.55 0.55 0.60 0.60 0.60 0.60 0.20 0.20 0.20 0.18 0.20 0.18 0.20 0.18 0.57 0.18 0.50 0.52 0.52 0.52 0.68 0.49 0.52 0.52 0.68 13.40 10.60 12.oo 3.27
'.27 3.18 22.20 6.06 20.00 2.98 2.98 3.51 3.51 16.50 16.60 28.30 28.30 5.83 5.83 5.61 5.61 15.40 15.40 3.91 3.91 3.91 3.91 41.10 41.10 7.24 7.24 4.72 4.72 3.23 3.23 3.23 8.05 8.05 10.70 10.70 10.70 20.70 8.59 B.SO
?.69 12.10 12.10 6.04 6.68 7.92 8.37 7.51
?.51 11.20 11.20 17.30 17.30 14.70 14.70 16.30 16.30 16.20 16.20 12.50 12.50 9.29 9.29 9.29 9.29 18.10 18.10 B.90 6.90 B.40 6.4o 7.67 7.87 7.67 8.15 8.15 6.11 6.1'I 6.11 14.50 105 115 110 0
30 85 140 80 127 0
66
~
65 73 115 90 135 105 85 55 103 133 165 165 30 20 15
?0 75 150 45 70 145 70 89 137 118 150 140 150 170 155 140
-17 Q3 30 0
-18 6
33 22
-10 20
-2 28 22
'2 2
20 23 24 4
%7 9
-2
-28
'7
>>17 9
-18 PABLF. 5 COMTINUFO)
1 cP~
C
'I ~
~
P P
P P
P P
P P
P P
P P
P P
P P
P P
P P
P P
P W
W W
W W
W W
W W
W
'W W
W W
W W
W W
W W
. W W
W W
W W
W W
W W
W W
W W
W W
W W
W W
W W
KEWAUN MCG 1 MCG 1 MCG 1 MCG1 SALEM 1 SALEM 1 SALEM 1 SALENI 1 SALEM 1 SALEM 1 SALEM2 SAlEM2 SONGS 1 SONGS)
WOLF 1 WOLF 1 ANO-2 CC1 CC2 FT. CAL FT. CAL MILLSTON2 MAINEY MAINEY MAINEY PALISA0 PALISAO ST. LUC 1 ST. LUG2 SONGS2 Bv 1 Bv1 BV1 COOK 1 COOK 1 V
U U
X X
T T
T T
Y Y
T T
A A
0 0
0 0
F F
0 U
W-97 W-263 W-263 W-225 W-265 W-97 A-25 A45 W-2e3 A-240 W-290 W-97 W43 W-97 U
V W
T Y
8HSS02 PMC101 PMC101 PMC1Q1 PMC101 PSA101 PSA102 PSA103 SHSS02 PSA103 SHSS02 PSA201 PSA201 PSO103 SASTM PSO101 PS0102 PSO1Q3
'ASTM PSO102 SASTM PWC101 PWC101 WAN20 WCC101 WCC201 WFC101 WFC101 WML201 WMYOf WMY01 WMY01 WPAL101 WPAL101 WSL101 WSL201 WSO201 WBV101 WBV101 WBV101 WCKfof WCK101 LT LT TL LT LT LT LT LT LT LT LT TL LT LT LT LT LT LT LT LT TL 0.14 0.09 0.09 0.09 0.09 0.22 0.23 0.22 0.14 0.22 0.14 0.10 0.10 0.18 0.20 0.17 0.18 0.18 0.20 0.18 0.20 0.07 0.07 0.04 0.24 0.20 0.35 0.35 0.30 0.36 0.36 0.36 0.22 0.22 0.23 0.05 0.03 0.28 0.26 0.26 0.27 0.27 0.68 0.60 0.60 0.60 0.53 0.54 0.52 0.68 0.52 0.68 0.61 0.61 0.20 0.18 0.20 0.20 0.20 0.18 0.20 0.18 0.62 0.62, 0.08 0.18 0.04 0.60 0.60 0.08 0.78 0.78 0.78 1.27 1.27 Q.11 0.07 012 0.62 0.62 0.62 0.74 0.74 8.41 4.14 4.14 13.80 13.80 2.84 2.84 2.84 2.84 8.91 8.91 2.56 2.56 28.60 28.60 56.20 56.20 56.20 56.20 57.30 57.30 3.39 3.39 3.34 6.10 7.97 5.83 8.00 3.77 17.60 77.30 5.67 60.60 10.30 5.30 1.62 5.07 B.54 2.91 9.49 2.71 10.60 15.80 14.20 14.20 10.10 10.10 8.29 8.29 8.29 8.29 8.33 8.33 6.99 6.99 49.10 49.10 63.3D 63.3D 63.30 63.30 23.50 23.50 12.00 12.00 6.26 B.58 5.52 5.60 4.61 3.98 43.00 61.40 4.70 62.00 B.56 3.60 4.60 4.80 5.79 7.92 5.11 6.79 6.80 95 45
~
50 45 65 100 1QQ 75 60 110 125 50 70 100 120 140 110 130 150 120 130 30 25 10 59 69 205 221 76 270 345 222 350 290 74 0
15 155 150 185 80 200 C'6 0
4 24 7
18 8
-15 21 1
11 13 1
6 12 44 15
-25
-18 33
-22 16 17 10 6
-29
-5 53 9
(TABLE) CONTINUED)
1' I.
~
~
I W
W W
W W
W W
W W
W W
W W
W W
W W
W W
W W
W W
W W
W W
W W
W W
W W
W
.W W
W W
W W
W W
W W '
W W
W W
W CALLA1 HADNEC HADNEC DIAB 1 DIAB2 FARLEY 1 FARLEY1 FARLEY 1 FARLEY2 FARLEY2 ROBIN 2 ROBIN 2 IP2 IP3 IP3 IP3 KEWAUN KEWAUN MCG 1 INCGf SALEM 3 SALEM2 SONGS)
SONGS 1 WOLF 1 U
A D
S U
0 x
Y U
W T
V Y
.T Y
Z R
V U
X Y
T A
F U
WCL101 WCTY01 Wc'IY01 WDC101 WDC201 WFA1Q1 WFA101 WFAf01 WFA20'I WFA201 WHB201 WH8201 WIP20f WIP301 VNP301 WIP301 WKWE01 WIOtVE01 Mmcfo1 WMC101 WSA101
.WSA20f WSO101 WSO101 WhfCf01 0.06 0.22 0.22 0.21 0.22 0.14 0.1 4 0.14 0.03 0.03 0.34 0.34 0.23 0.15 0.1 5 0.15 0.20 0.20 0.21 0.21 0.18 0.23 0.19 0.19 0.07 0.05 0.05 0.98 0.83 0.19 0.1 9 0.19 0.90 0.90 0.66 0.66 1.06 1.02 1.02 1.02 0.77 0.77 0.88 O.ee 1.26 0.71 0.20 0.20 0.09 3.27 3.16 22.20 2.98 3.51 16.50 28.3Q 5.83 5.61 15.40 41.10 7.24 5.89 323 8.05 10.70 20.70
'.41 4.14 13.80 8.91 2.56 28.60 57.30 3.39 12.1o 6.04 6.68 7.51 11.20 17.30 14.70 16.30
'IB.20 12.50 18.10 B.90 7.99 7.67 8.15 B.f1 14.50 15.80 14.20 10.10 8.33 6.99 49.1o 23.50 12.00 70 95
~
110 110 174'0 100 80 10 10 285 175 195 143 180 220 235 175 160 165 165 155 80 145 20 AB
-27 10 4f
-28 9
0
-14 24 38 0
-24 9
-10 B4 43 47 48 3
@ABLE ) CQNTINUEQ)
SCHEMATIC OF THE METHODUSED TO SEPARATE SURVHLLANCEDATAFROM CE FABRICATED,CE ANDW DESIGNED PLANTS FOR STATISTICALANALYSIS
- Plates CB
/5 NRC CEOG N
NRC CBOG NRC CB G
FIGURE I
~ ~
TABLE2-
SUMMARY
OF RESULTS FROM MANN-WHITNEYSTATISTICALTEST POPULATION TESTED CEOG DATAFOR CE ANDW DESIGNED PLATES CEOG DATAFOR CE ANDW DESIGNED PLATES-OMENTATION ANALYSIS(LTANDTL)
NRC DATAFOR CE ANDW DESIGNED PLATES NRC DATAFOR CE ANDW DESlGNED PLATES-OMENTATION ANALYSlS(LTANDTL)
CEOG DATAFOR CE ANDW DESIGNED WELDS NRC DATAFOR CE ANDW DESIGNED WELDS DECISION CRITERIAFOR Z STATISTIC +
-1.96 s Z z 1.96
-1.96 z Z s 1,96
-1.96 s Z s 1.96
-1.96 s Z s 1.96
-1.96 s Z s 1.96
-1.96 s Z s 1.96 RESULTING Z STATISTIC 1.05 0.83 1.60 0.83 0.88 1.06 Atwo tailed test, employing a critical probability (a) of0.05 results in an acceptable Z range of -1.96 s Z s 1.96. Therefore, the null hypothesis was accepted for each case since the Z statistic was in the acceptable range.
~
~
~ ~
~
~
~
e e o>~g o~~,~w
+l
CE Designed Yessel Plates 0.025 0.02 3
Z P~
i n
c II Ii II g.
4
]
0.015 ~
CC CC C
- 5 ta Ic
- C
'Cg c
- C c
0.005 0 -100
-85
-70
-55
-40 '25
-t0 5
20 35 50 65 'O 85 CVT Shift Difference (Pred-Act),
F
~"-.=-.,:>>;a GEOG data
.Norm Prob Dens Funct r
FIGURE 2-HISTOGRAM OF COMBUSTION ENGlNEERING DESIGNED PLATE DATA
Nestinghouse Designed Yessel Plates
~
~ 15, 0,025 I.
12 0.02 C
C A
(
?l
',c Fp (f
'.C "5
0,015 0.01 0.005 0 -100
-'85
-70
-55. -40
-25
-1 5
00 85 50 85 80 95 CVTShift Difference (Pred-Act), 'F 0
GEOG data
Norm Prob Dens Func FIGURE 3-HISTOGRAM OF WESTINGHOUSE DESIGNED PLATE DATA s ~
~ g,
~
~ ~,
~
tt
~
~
~
1
0
~f V(
CE Designed Vessel Plates Orientation Analysis - LT 5
0.05 0.04 0.03 ~
CO 0
0.02 ~
0.01
-100 45
-70
-55 40
-25
'-10 5
00 35 50 85 80 85 CVT Shift Difference (Pred-Act), 'F 0
"-'-:,-:j GEOG data
Norm Prob Dens Func FIGURE 4 - HISTOGRAM OF CE DESIGNED PLATE DATA(LTORIENTATION)
~
4,
$ ~
Westinghouse Designed Vessel Plates Orientation Analysis - LT 0.025 10
- 0.02 0.015 ~
P 0.01
- C (C
cj
]c C
'4'f
~3'.005
'I
'::I
~
SC r$
I $$
I:jl 0
0
-100 -85
-70
-55 '-40
-25
-'10 5
0 35 50 65 80 95 GVT Shift Difference (Pred-Act), 'F
- ==-~> GEOG data
Norm Prob Dens Func FIGURE 5 - HISTOGRAM OF WESTINGHOUSE DESIGNED PLATE DATA(LT ORIEN.)
CE Designed Vessel Plates Orientation Analysis -TL 0.02 0.015 2
LL 0.01 m
Q.
0.005 0
'.,'."::~ GEOG data I
-100
-85
-70
-55
-40
-25
-10 5
20 35 50 65 80 95 CVT ShN ONerence (Pred-Act),
F
Norm Prob Dens Func 0
FlGURE 6 - HlSTOGRAM OF CE DESlGNED PLATE DATA(TL ORlENTATlON)
'L
es tng ouse estgne esse e
s Orientation Analysis - TL 0.03
- 0.025 0.02
- 2 0.015 c0 CL 0.01 0.005 0 -100 45
-70
-55 40
-25
-10 5
20 35 50 85 80 S5 CVT Shift Difference {Pred-Act), F
'=:.'
.:,,: CEOG data
Norm Prob Dens Fane
.FIGURE 7-HlSTOGRAM OF WESTINGHOUSE. DESIGNED PLATE DATA(TLORIEN.)
~ ~ ~
e
0
~ g
~s I)
40 ALLPLATE DATA(CE ANDN) 35 25 5 20 15 0
00 -8 CVT Shift Difference (Pred-Act),OF Combution Engineering Plates Westinghouse Plates FlGURE 8 - HISTOGRAM OF ALLPLATE DATA (CE ANDW DESIGNED)
~ ~
~ \\
~
0 a<
I
a CE Designed Vessel Nfelds 0.025 0.02 p 3 0.015 0.01 0.005 0
-100
-85
- 0
-55 4
-2
-1 5
0 0
CVT Shift Difference (Pred-Act), 'F 0
'-.:;=,.-.. CEOG data
Norm Prob Dens Func F!GURE 9 - HISTOGRAM OF COMBUSTION ENGlNEERING DESlGNED WELD DATA y
~y
~
0-
~ ~
~ '
Westinghouse Designed Vessel Welds 0.02 0.015 0.01 ca P
LL j
']
0.005 0
~
~
-00 45 -0
-55 4
-2
-1 5
0 CVT Shift Difference (Pred-Act), 'F 0
5 0
0
- .,':.:;. GEOG data
Norm Prob Dens Fvnc FlGURE 10 - HlSTOGRAM OF WESTlNGHOUSE DESlGNED NfELDDATA
e
) 4 I
ALLWELD DATA(GE ANDW) 3 4-2 00 -8
-5 4
-25
-1 5
CVT Shift Difference (Pred-Act), oF Combustion Engineering Welds Westinghouse Welds FIGURE 11 - HISTOGRAM OF ALLWELD DATA (CE ANDW DESIGNED)
0 II I