ML20129F952
| ML20129F952 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 09/25/1996 |
| From: | Hosmer J COMMONWEALTH EDISON CO. |
| To: | NRC (Affiliation Not Assigned), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML19355E418 | List: |
| References | |
| NUDOCS 9610020126 | |
| Download: ML20129F952 (14) | |
Text
Commonweahh FAison Company 1400 Opus Place Downers Grove, P.60515-5701 September 25,1996 Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 Attn: Document Control Desk
Subject:
Byron and Braidwood Station Units 1 & 2 Response to Request for Additional Information Pertaining to Application of Leak-Before-Break Methodology to the Primary Reactor Coolant System (RCS)
NRC Docket Numbers: 50: 454. 455. 456 and 417
References:
1.
mer letter to the Nuclear Regulatory Commission dated Ap.i 30,1996, transmitting Application of Leak-Beibre-Break Methodology to the Primary Reactor Coolant System
- 2. R. Assa letter to I. Johnson dated August 26,1996, transmitting Request for Additional Information Pertaining to Application of Leak-Before-Break Methodology to the Primary Reactor Coolant System Reference 1 transmitted the Commonwealth Edison Company's (Comed) application of Leak-Before-Break Methodology to the Primary Reactor Coolant System for use at Byron and Braidwood Station Units 1 and 2. Reference 2 transmitted the Nuclear Regulatory Commission's Request for Additional Information (R AI). Attached is Comed's response to this RAI. Specifically, Attachment A contains the proprietary versions of the RAI response, l
Attachment B contains the non-proprietary version of the RAI response, and i
A: chment 1 is applicable to the non-proprietary and proprietary versions.
Please note that the attachment contains informstion proprietary to Westinghouse Corporation and is supported by an aflidavit signed by Westinghouse, the owner of the information. The aflidavk ( Attachment C) sets for1h the basis on which the information may be withheld from public disciocre by the Commission and addresses with specificity the considerations listed in Paragraph (b)(4) of htsn 2.790 of the Commission's regulations. Accordingly, it is respectfully requested that tne ;c.fc mation which is proprietary to Westinghouse be withheld from public disclosure in accordance wth 19 CFR 2.790 of the Commission's regulations.
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U.S. Document Control Desk September 25,1996 Correspondence with respect to the proprietary aspects of the items should be addressed to N.J. Liparulo, Manager Nuclear Safety Regulatory and Licensing Activities, Westinghouse Corporation Box 355 Pittsburgh, Pennsylvania. 15230-0355.
Comed is requesting approval of this analysis by October 15,1996, to support design work relating to the steam generator replacement project.
If you have any questions concerning this correspondence, please contact Denise Saccomando at (630) 663-7283.
Sincerely, d /be John B. Hosmer Engineering Vice President Attachments cc:
G. Dick, Byron Project Manager-NRR R. Assa, Braidwood Project Manager-NRR C. Phillips, Senior Resident inspector-Braidwood H. Peterson, Senior Resident Inspector-Byron Office ofNuclear Safety-IDNS i
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Attachment B Non-Proprietary Version of the RAI Response i
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i NON-PROPRIETARY INFORMATION ATTACHMENTB i
- 1. From what type or grade of weld material are the stainless steel field welds in the primary loop manufactured? Submit any information available regarding the chemical composition, delta-ferrite percentage, tensile behavior (unaged and aged), and fracture toughness data (unaged and aged) for the Byron /Braidwood welds.
Response: Field welds are made of Tungsten Ined Gas (TIG) and Submerged Manual Arc Welding (SMAW) process combination.
Byron /Braidwood weld information (Certified Material Test Reports-CMTRs) is only available for the unaged condition and is provided in non-proprietary version Attachment 1.
With respect to weld material properties for the aged conditions, page 4-3 of WCAP-14559 Revision.1 statesin part that, '. Available data on aged stainless steel welds (References 4-2 and 4-3) indicate that Jic values for the worst case welds are of the same order as the aged material..
Therefore, weld regions are less limiting than the cast material.. "
- 2. In Appendix B, Tables 8-1, B-3, and B-4, chemical composition data is provided for 10.0-inch ID cast components. Since this LBB analysis was only analyzed for the fracture behavior of the 27 to 33 inch primary loop piping, why is this data included?
It appears that no mechanical property data from the aforementioned heats of material are included in the tables of Section 4.0. Confirm that no test results for the heats of 10.0-inch ID cast sections have been included in the average values presented in Table 4-9.
Response: In Appendix B, Tables B-1, B-2, B-3, and B-4, chemical composition data for 10.0 inch ID cast components was included for reference only. The 10.0 inch ID cast components were not usedin the primary loop piping and the ass.>ciated data was not used for the Byron Units 1 and 2 and Braidwood Units 1 and 2 primary loop LBB analysis.
The mechanicalproperties data for the 10.0 inch ID cast components was not consideredin Section 4 norin Table 4-9
NON-PROPRIETARY INFORMATION ATTACHMENT B
- 3. The stress-strain curve in WCAP-14559, Revision 1, Figure 4-1 for SA351 CF8A has been provided without sufficient information to determine its applicability to the LBB analysis.
Provide the following:
(a) copies of the pages from the source document and any necessary calculations to show how this curve was "obtained",
(b) identification of the condition (aged or unaged) of the material that is characterized by Figure 4-1, (c) if aged, aging conditions, (d)if available, values of the Ramberg-Osgood parameters which fit to the stress-strain data for the leakage crack opening determination, and (e) justification for how this curve, as used in the LBB calculations, conservatively represents the material properties of SA351 CF8A until end-of-life (EOL).
Response
(a)
Westinghouse utilized the methodology developed in Reference 4-1 (Nuclear Systems Materials Handbook, Part I-Structural Materials, Group 1 - High Alloy Steels, Section 2, ERDA Report TID 26666, November 1975) of WCAP-14559, Revision 1 to determine tne appropriate stress-strain curves for the stainless steel materials being analyzed for Byron and Braidwood. The work of Reference 4-1 was sponsored by the Energy Research and Development Agency (ERDA) at the Hanford Engineering Development Laboratory (HEDL) in 1975. The methodology of Reference 4-1 was developed through extensive analyses of actual stress-strain behaviors from many experimental results. This methodology has been used by Westinghouse for allprevious LBB submittals and has been accepted by the NRC.
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NON-PROPRIETARY INFORMATION Response to Question 3a continued: The specific equations from Reference 4-1 used to determine the stress-curves are provided below:
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NON-PROPRIETARY INFORMATION ATTACHMENT B i
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NON-PROPRIETARY INFORMATION ATTACHMENTB Insert calc 4
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NON-PROPRIETARY INFORMATION ATTACHMENT B (b)
Yield strength information from Byron /Braidwood primary loop piping certified material test reports was used for the unaged condition. The materialyield strength increases slightly due to aging; therefore, the use of the unaged material conditions is more conseNative and appropriate. Figure 4-1 was generated using the unaged lower bound materialyield strength of the Byron /Braidwood primary loop piping at operating temperature.
(c)
As indicated in the response to question 3(b), unaged material conditions were used for Figure 4-1.
(d)
Figure 4-1 was generated using the minimum yield strength. Figure 4-1 is not applicable for the leakage crack determination. For the leakage crack determination, the average yield strength was used as indicated in Section 6.4 on page 6-2 of WCAP-14559 Revision 1. For the average yield strength, no stress-strain cuwe was generated and therefore Ramberg-Osgood parameters are not available.
(e)
Figure 4-1 shows the lower bound stress-strain cune. Since the cume was generated using the lowest yield strength from all the heats of Byron /Braidwood Units 1 and 2 primary loop piping cast materials (SA351-CF8A), this cume is a consewative representation. As indicated earlier, the materialyield strength does not reduce due to aging and therefore the yield strength used to generate Figure 4-1 is representative through end-of-life (EOL).
- 4. Provide the stress intensity factor range (K. and K,nn) associated with each of the transients listed in Table 8-1 for the fatigue crack growth analysis.
Response: The calculated fatigue crack growth analysis for the primary j
loop piping, regardless oflocation, was demonstrated to be very small per a previous LBD submittal for the Byron and Braidwood units. This submittal (WCAP-10553, May 1984) was reviewed and approved by the NRC in the Safety Evaluation dated October 28,1985. The analysis presented in WCAP-14559 Revision 1 is consistent with WCAP-10553.
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NON-PROPRIETARY INFORMATION ATTACHMENT B l
- 5. A significant difference is given in Table 9-1 regarding the critical flaw size calculated for location 3 between the Z-factor modified limit load [
]
and J-integral [
] approaches. Due to differences in the methodology, some variation in the results would be expected. However, provide a rationale for the wide variation shown in Table 9-1.
Response: Location 3 is the criticallocation for the cast elbows. Limit load and J-integral results differ because of the nature of the analysis methodology, For the thermally aged cast material, J-integral analysis results govern. Limit load analysis results forlocation 3 were included as additionalinformation.
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- 6. It is stated in Section 4.3 that it is conservative to assume EOL toughness properties for the various piping materials as equal to those for [
). In reviewing WCAP-10456, November 1983, it is noted that a conclusion is drawn on page 2-12 that, [
).
Explain now the values given for J,, and T,n. in Table 4.1-of the current submittal are consistent with theses reported in WCAP-10456.
Response: The French data was mentioned for comparison purposes only. The basicintent was to show that an independent data source found very similar values. As statedin WCAP-10456 page 2-10,[
] ** The French data also specified[
]**
[
J '" data in WCAP-10456 is based on a more representative Westinghouse testing of[
]**
] ** fracture toughness data was based on the testing of multiple specimens from which many data points were obtained. The Westinghouse [
J '", which falls within this range specified in WCAP-10456, is considered to represent the best set of cunservative values for use in these analyses. These values are consistent w,th those l
previously approved in the Safety Evaluation dated December 22,1986, l
for Northem States Power
NON-PROPRIETARY INFORMATION ATTACHMENT B
- 7. As a followup to question #6, provided information on Jmo as derived from the tests by [
). Does this data support the Jmo value given for
[
] in Table 4-10 of the current submittal?
Response: The French data did not specifically report Ja,x values. The Westinghouse [
J '** material Jmu value (reference WCAP-10456' and WCAP-10931, Revision 1), was obtained when the maximum load was applied during testing and has bee, utilized in previous Westinghouse LBS submittals. The Jma utilizedin the Byron /Braidwood i
analyses carry appropriate conservatisms.
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i (a) details of the fabrication process for these welds, (b) data on the weld materials' tensile behavior, (c) data on these welds' chemical composition, and
)
(d) fracture toughness data which have been acquired for the weld material and heat affected zones.
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Response
(a)
Rcactor vessel nozzles were buttered with Inconel material SPA-5.11,E-Ni-Cr-Fe-3 prior to post weld heat treatment (PWHT) being performed on the reactor vessel. The stainless steel (SS) safe-ends were then welded to the Inconel buttering with Inconel SFA-5.11.
Safe-end to piping weld material was 309SS.
(b)
SFA-5.11 E-Ni-Cr-Fe-3 material requires tensile tests to.:
performed and have a minimum tensile strength of 80,000 psi.
(c)
Chemical composition of SFA-5.11 E-Ni-Cr-Fe-3 is as follows:
C =0.10, Mn=5.0 to 9.5, Fe=10.0, P=0.03, S=0.015, Si=1.0, Cu=0.50, Ni=59.0 min., Ti=1.0, Cr=13.0 to 17.0, and Cb plus Ta=1.0 to 2.5. (Note: weight percent)
(d)
Neither the ASME Code nor the Westinghouse equipment specification require fracture toughness testing. As a result, none was performed.
NON-PROPRIETARY INFORMATION ATTACHMENT B i
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- 9. Provide a tensile stress-strain curve which is applicable for the SA376 Grade l
304N piping of the Byron /Braidwood primary loop piping. If available, also provide values of Ramberg-Osgood parameters fit to the stress-strain data l
for use in leakage crack determination.
Response: A stress-strain curve for the straight pipe material SA376 i
Grade 304N for the Byron /Braidwood prime"? loop piping was not generated for the leakage crack open'ng determination. [
.] '".
Note - Coding "' associated with the brackets sets forth the basis on which the i
information is considered proprietary. These codes are listed with their meanings in WCAP-7211 (Reference 1-11).
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Attachment C Affidavit for Withholding K:nla'bybwd\\sgrp\\llbrai