RS-12-074, Response to Request for Additional Information Regarding Relief Request I4R-19, Corrosion and Flaw Evaluations: Difference between revisions

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| issue date = 04/13/2012
| issue date = 04/13/2012
| title = Response to Request for Additional Information Regarding Relief Request I4R-19, Corrosion and Flaw Evaluations
| title = Response to Request for Additional Information Regarding Relief Request I4R-19, Corrosion and Flaw Evaluations
| author name = Gullott D M
| author name = Gullott D
| author affiliation = Exelon Generation Co, LLC, Exelon Nuclear
| author affiliation = Exelon Generation Co, LLC, Exelon Nuclear
| addressee name =  
| addressee name =  
Line 18: Line 18:


=Text=
=Text=
{{#Wiki_filter:Exelon Generation Company, LLC 4300 Winfield Road www.exeloncorp.com Nuclear Warrenville, IL 60555 RS-12-074 10 CFR 50.55a April 13, 2012 U. S. Nuclear Regulatory Commission A TIN: Document Control Desk Washington, D.C. 20555-0001 Quad Cities Nuclear Power Station, Unit 2 Renewed Facility Operating License No. DPR-30 NRC Docket No. 50-265  
{{#Wiki_filter:Exelon Generation Company, LLC     www.exeloncorp.com 4300 Winfield Road                                                                        Nuclear Warrenville, IL 60555 RS-12-074 10 CFR 50.55a April 13, 2012 U. S. Nuclear Regulatory Commission A TIN: Document Control Desk Washington, D.C. 20555-0001 Quad Cities Nuclear Power Station, Unit 2 Renewed Facility Operating License No. DPR-30 NRC Docket No. 50-265


==Subject:==
==Subject:==
Response to Request for Additional Information Regarding Relief Request 14R-19, Corrosion and Flaw Evaluations  
Response to Request for Additional Information Regarding Relief Request 14R-19, Corrosion and Flaw Evaluations


==References:==
==References:==
: 1) Letter from P. R. Simpson (Exelon Generation Company, LLC) to NRC, "Relief Request 14R-19 Associated with the Reactor Pressure Vessel Nozzle Repairs," dated April 6, 2012 2) Letter from D. M. Gullott (Exelon Generation Company, LLC) to NRC, "Response to Request for Additional Information Regarding Relief Request 14R-19 Associated with the Reactor Pressure Vessel Nozzle Repairs," dated April 12, 2012 3) Email fromJoeIWiebe(NRC)toD.M.Guliott (Exelon Generation Company, LLC), "Quad Cities Unit 2 Relief Request 14R-19 -Followup Question to Response to RAI Question 8," dated April 13, 2012 4) Email fromJoeIWiebe(NRC)toD.M.Guliott (Exelon Generation Company, LLC), "Quad Cities Unit 2 Relief Request 14R-19 -Followup Question to Regarding the Flaw Evaluation," dated April 13, 2012 In Reference 1, in accordance with 10 CFR 50.55a, "Codes and standards," paragraph (a)(3)(i), Exelon Generation Company, LLC (EGC), requested NRC approval of a relief request associated with the Fourth Inservice Inspection (lSI) Interval for Quad Cities Nuclear Power Station (QCNPS), Unit 2. Note that the fourth interval of the QCNPS lSI program complies with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, 1995 Edition with addenda through 1996. During review of the subject relief request, the NRC concluded that additional information would be needed to complete their review. In Reference 2, EGC provided a portion of the requested  
: 1) Letter from P. R. Simpson (Exelon Generation Company, LLC) to NRC, "Relief Request 14R-19 Associated with the Reactor Pressure Vessel Nozzle Repairs,"
dated April 6, 2012
: 2) Letter from D. M. Gullott (Exelon Generation Company, LLC) to NRC, "Response to Request for Additional Information Regarding Relief Request 14R-19 Associated with the Reactor Pressure Vessel Nozzle Repairs," dated April 12, 2012
: 3) Email fromJoeIWiebe(NRC)toD.M.Guliott (Exelon Generation Company, LLC), "Quad Cities Unit 2 Relief Request 14R Followup Question to Response to RAI Question 8," dated April 13, 2012
: 4) Email fromJoeIWiebe(NRC)toD.M.Guliott (Exelon Generation Company, LLC), "Quad Cities Unit 2 Relief Request 14R Followup Question to Regarding the Flaw Evaluation," dated April 13, 2012 In Reference 1, in accordance with 10 CFR 50.55a, "Codes and standards," paragraph (a)(3)(i),
Exelon Generation Company, LLC (EGC), requested NRC approval of a relief request associated with the Fourth Inservice Inspection (lSI) Interval for Quad Cities Nuclear Power Station (QCNPS), Unit 2. Note that the fourth interval of the QCNPS lSI program complies with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, 1995 Edition with addenda through 1996.
During review of the subject relief request, the NRC concluded that additional information would be needed to complete their review. In Reference 2, EGC provided a portion of the requested


U. S. Nuclear Regulatory Commission April 13, 2012 Page 2 information.
U. S. Nuclear Regulatory Commission April 13, 2012 Page 2 information. In Reference 3, the NRC requested follow-up information to supplement the information provided in Reference 2 regarding the Corrosion Evaluation (Le., the information provided in Request for Additional Information (RAI) Question 8). In Reference 4, the NRC requested follow-up information to supplement the information provided in Reference 2 regarding the Flaw Evaluation (Le., the information provided in RAI Question 9). The follow-up information requested in References 3 and 4 is provided in Attachment 1 to this letter.
In Reference 3, the NRC requested follow-up information to supplement the information provided in Reference 2 regarding the Corrosion Evaluation (Le., the information provided in Request for Additional Information (RAI) Question 8). In Reference 4, the NRC requested follow-up information to supplement the information provided in Reference 2 regarding the Flaw Evaluation (Le., the information provided in RAI Question 9). The follow-up information requested in References 3 and 4 is provided in Attachment 1 to this letter. There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Mr. Joseph A. Bauer at (630) 657-2804.
There are no regulatory commitments contained in this letter.
Respectfully, David M. Gullott Manager -Licensing Exelon Generation Company, LLC Attachment 1: Response to Request for Additional Information Regarding Relief Request 14R-19, Corrosion and Flaw Evaluations ATTACHMENT 1 Response to Request for Additional Information Regarding Relief Request 14R-19 Corrosion and Flaw Evaluations In Reference 1, in accordance with 10 CFR 50.55a, "Codes and standards," paragraph (a)(3)(i), Exelon Generation Company, LLC (EGC), requested NRC approval of a relief request associated with the Fourth Inservice Inspection (lSI) Interval for Quad Cities Nuclear Power Station (QCNPS), Unit 2. Note that the fourth interval of the QCNPS lSI program complies with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, 1995 Edition with addenda through 1996. During review of the subject relief request, the NRC concluded that additional information would be needed to complete their review. In Reference 2, EGC provided a portion of the requested information.
Should you have any questions concerning this letter, please contact Mr. Joseph A. Bauer at (630) 657-2804.
In Reference 3, the NRC requested follow-up information to supplement the information provided in Reference 2 regarding the Corrosion Evaluation (Le., the information provided in Request for Additional Information (RAI) Question 8). In Reference 4, the NRC requested follow-up information to supplement the information provided in Reference 2 regarding the Flaw Evaluation (Le., the information provided in RAI Question 9). The follow-up information requested in References 3 and 4 is provided below. Corrosion Evaluation RAI 1 With respect to general and galvanic corrosion of low-alloy steel, the licensee's corrosion evaluation, provided as Attachment 2 to Letter RS-12-069 dated April 12, 2012, references "NECD-21120, "Monticello Feedwater Nozzle Cladding Crack Repair Report," as the basis for low general corrosion rates and the absence of galvanic corrosion.
Respectfully, David M. Gullott Manager - Licensing Exelon Generation Company, LLC : Response to Request for Additional Information Regarding Relief Request 14R-19, Corrosion and Flaw Evaluations
The licensee's corrosion evaluation references "D.C. Vreeland, et. AI., "Corrosion of Carbon and Low-Alloy Steels in Out-of-Pile Boiling Water Reactor Environment," Corrosion, Vol. 17, No.6, 1961, as the basis for minimal crevice corrosion of low alloy steel in a bOiling water reactor (BWR) operating environment.
 
ATTACHMENT 1 Response to Request for Additional Information Regarding Relief Request 14R-19 Corrosion and Flaw Evaluations In Reference 1, in accordance with 10 CFR 50.55a, "Codes and standards," paragraph (a)(3)(i),
Exelon Generation Company, LLC (EGC), requested NRC approval of a relief request associated with the Fourth Inservice Inspection (lSI) Interval for Quad Cities Nuclear Power Station (QCNPS), Unit 2. Note that the fourth interval of the QCNPS lSI program complies with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, 1995 Edition with addenda through 1996.
During review of the subject relief request, the NRC concluded that additional information would be needed to complete their review. In Reference 2, EGC provided a portion of the requested information. In Reference 3, the NRC requested follow-up information to supplement the information provided in Reference 2 regarding the Corrosion Evaluation (Le., the information provided in Request for Additional Information (RAI) Question 8). In Reference 4, the NRC requested follow-up information to supplement the information provided in Reference 2 regarding the Flaw Evaluation (Le., the information provided in RAI Question 9). The follow-up information requested in References 3 and 4 is provided below.
Corrosion Evaluation RAI 1 With respect to general and galvanic corrosion of low-alloy steel, the licensee's corrosion evaluation, provided as Attachment 2 to Letter RS-12-069 dated April 12, 2012, references "NECD-21120, "Monticello Feedwater Nozzle Cladding Crack Repair Report,"
as the basis for low general corrosion rates and the absence of galvanic corrosion. The licensee's corrosion evaluation references "D.C. Vreeland, et. AI., "Corrosion of Carbon and Low-Alloy Steels in Out-of-Pile Boiling Water Reactor Environment," Corrosion, Vol. 17, No.6, 1961, as the basis for minimal crevice corrosion of low alloy steel in a bOiling water reactor (BWR) operating environment.
Requested Information To ensure these two references provide bounding corrosion rates for Quad Cities, Unit 2, compare the chemistry of the steam or water to which the low-alloy steel in the repaired N-11B nozzle will be exposed (in the gap between the nozzle remnant and new nozzle) to the chemistry used in the tests described in the two references discussed above. The specific parameters to be addressed should include the concentration of dissolved oxygen and impurities such as chloride and sulfate, conductivity, hydrogen, electrochemical corrosion potential (ECP), and any other parameters that may affect corrosion rates. Values of the parameters for Quad Cities may be typical or average values or maximum values. If the steam/water environment tests documented in the references do not envelop or bound the actual conditions to which the low-alloy steel in Nozzle N-11 B will be exposed, provide a justification for the use of these reference to support the prediction of minimal corrosion for the exposed low-alloy steel, or provide additional references.
Requested Information To ensure these two references provide bounding corrosion rates for Quad Cities, Unit 2, compare the chemistry of the steam or water to which the low-alloy steel in the repaired N-11B nozzle will be exposed (in the gap between the nozzle remnant and new nozzle) to the chemistry used in the tests described in the two references discussed above. The specific parameters to be addressed should include the concentration of dissolved oxygen and impurities such as chloride and sulfate, conductivity, hydrogen, electrochemical corrosion potential (ECP), and any other parameters that may affect corrosion rates. Values of the parameters for Quad Cities may be typical or average values or maximum values. If the steam/water environment tests documented in the references do not envelop or bound the actual conditions to which the low-alloy steel in Nozzle N-11 B will be exposed, provide a justification for the use of these reference to support the prediction of minimal corrosion for the exposed low-alloy steel, or provide additional references.
Response The corrosion evaluation (51-9180975-001) references NEDC-21120 regarding general and galvanic corrosion.
 
The corrosion rate from NEDC-21120 is based on tests performed with low Page 1 of 4 ATTACHMENT 1 Response to Request for Additional Information Regarding Relief Request 14R-19 Corrosion and Flaw Evaluations alloy steel in contact with BWR water conditions and BWR steam conditions.
===Response===
The corrosion rate used in 51-9180975-001 is 70% faster than the maximum corrosion rate than all of the tests from NEDC-21120.
The corrosion evaluation (51-9180975-001) references NEDC-21120 regarding general and galvanic corrosion. The corrosion rate from NEDC-21120 is based on tests performed with low Page 1 of 4
NEDC-21120 does not provide specific water chemistry parameters.
 
ATTACHMENT 1 Response to Request for Additional Information Regarding Relief Request 14R-19 Corrosion and Flaw Evaluations alloy steel in contact with BWR water conditions and BWR steam conditions. The corrosion rate used in 51-9180975-001 is 70% faster than the maximum corrosion rate than all of the tests from NEDC-21120. NEDC-21120 does not provide specific water chemistry parameters.
NEDC-21120 was written in 1975; therefore, it can be inferred that the BWR water chemistry used in the tests was consistent with the water chemistry requirements for BWRs of 1975 or before. The water chemistry requirements of this time period are known to be much less stringent than today's modern requirements (Le., BWRVIP-130 and more recently BWRVIP-190). BWRVI P-130 was created to better control BWR water chemistry and is significantly more restrictive than the requirements of the 1970s (and before). Quad Cities Unit 2 reactor water meets or exceeds the requirements of BWRVIP-190, which supersedes BWRVIP-130.
NEDC-21120 was written in 1975; therefore, it can be inferred that the BWR water chemistry used in the tests was consistent with the water chemistry requirements for BWRs of 1975 or before. The water chemistry requirements of this time period are known to be much less stringent than today's modern requirements (Le., BWRVIP-130 and more recently BWRVIP-190). BWRVI P-130 was created to better control BWR water chemistry and is significantly more restrictive than the requirements of the 1970s (and before). Quad Cities Unit 2 reactor water meets or exceeds the requirements of BWRVIP-190, which supersedes BWRVIP-130.
BWRVIP-190 has requirements that are equal to or more restrictive than BWRVIP-130.
BWRVIP-190 has requirements that are equal to or more restrictive than BWRVIP-130. Based on this information, it is concluded that the water chemistry at Quad Cities Unit 2 is less aggressive than the water chemistry used in the tests discussed in NEDC-21120.
Based on this information, it is concluded that the water chemistry at Quad Cities Unit 2 is less aggressive than the water chemistry used in the tests discussed in NEDC-21120.
The corrosion evaluation (51-9180975-001) references Vreeland (1961) regarding crevice corrosion. For the same reasoning given above, it is concluded that the water chemistry at Quad Cities Unit 2 is less aggressive than the water chemistry used in the tests discussed in Vreeland (1961).
The corrosion evaluation (51-9180975-001) references Vreeland (1961) regarding crevice corrosion.
Flaw Evaluation RAI-1 The flaw evaluation report states that heatup and cooldown, SRV blowdown, and SCRAM are determined to be controlling for the current flaw evaluations. Please provide the cTiteria and the qualitative assessment that you have performed to determine that SRV blowdown and SCRAM are the controlling emergencyHaulted and normal/upset transients among all design transients.
For the same reasoning given above, it is concluded that the water chemistry at Quad Cities Unit 2 is less aggressive than the water chemistry used in the tests discussed in Vreeland (1961). Flaw Evaluation RAI-1 The flaw evaluation report states that heatup and cooldown, SRV blowdown, and SCRAM are determined to be controlling for the current flaw evaluations.
 
Please provide the cTiteria and the qualitative assessment that you have performed to determine that SRV blowdown and SCRAM are the controlling emergencyHaulted and normal/upset transients among all design transients.
===Response===
Response The bounding transients were determined from the Section III design analysis of record. The transients that have significant temperature excursions, and therefore, significant stress ranges, are heatup and cooldown, SCRAM, and SRV Blowdown.
The bounding transients were determined from the Section III design analysis of record. The transients that have significant temperature excursions, and therefore, significant stress ranges, are heatup and cooldown, SCRAM, and SRV Blowdown. Therefore, these are the transients that lead to the highest stress ranges, and they are controlling for the flaw evaluation.
Therefore, these are the transients that lead to the highest stress ranges, and they are controlling for the flaw evaluation.
Flaw Evaluation RAI-2 Confirm that the crack plane is assumed to exist at a circumferential location with respect to the nozzle hole such that the maximum remote membrane and bending stresses in Reference 5 were used for the subsequent fracture mechanics analysis.
Flaw Evaluation RAI-2 Confirm that the crack plane is assumed to exist at a circumferential location with respect to the nozzle hole such that the maximum remote membrane and bending stresses in Reference 5 were used for the subsequent fracture mechanics analysis.
Justify your selection of the crack plane location if this is not the case. Response Yes, the crack plane is assumed to exist at a circumferential location with respect to the nozzle hole such that the maximum remote membrane and bending stresses are applied. The stresses used are normal to the axial plane of the vessel, where the hoop stresses are at a maximum. Page 2 of 4 ATTACHMENT 1 Response to Request for Additional Information Regarding Relief Request 14R-19 Corrosion and Flaw Evaluations Flaw Evaluation RAI-3 Considering the difference in Young's modulus (E) values of the low alloy steel and the 182 weld and that the "local" weld residual stresses may not be adequately represented by the "averaging" membrane and bending stresses, please demonstrate that your approach will not severely underestimate the effect of residual stresses on the applied stress intensity factor. Response The difference in Young's modulus between the Alloy 182 weld material and the SA-302B (modified) low alloy steel (LAS), is reasonably small, and is not expected to have any significant effect on the material behavior in the elastic regime. Weld residual stresses are steady state secondary stresses, which are generally limited by yield strength.
Justify your selection of the crack plane location if this is not the case.
The Alloy 182 and LAS materials have similar yield strengths and the difference will not Significantly affect the weld residual stress distribution.
 
Since the vessel and 3/16 inch of the original weld were stress relieved during original fabrication, weld residual stresses were significantly relieved in those materials.
===Response===
The rewelding of the J-groove weld produced tensile residual stresses in the weld at roughly the yield level at temperature, and for equilibrium, the stresses in the adjacent LAS material will be slightly compressive and the stresses in the original Alloy 182 weld will be further relieved, possibly to the point of being compressive.
Yes, the crack plane is assumed to exist at a circumferential location with respect to the nozzle hole such that the maximum remote membrane and bending stresses are applied. The stresses used are normal to the axial plane of the vessel, where the hoop stresses are at a maximum.
In the present analysis, the crack tip is assumed to be within the LAS material, so the residual stresses assumed are conservative for the present analysis.
Page 2 of 4
Furthermore, assuming an average compressive stress in the LAS material will underestimate the compressive stresses near the crack front, resulting in a conservative analysis.
 
ATTACHMENT 1 Response to Request for Additional Information Regarding Relief Request 14R-19 Corrosion and Flaw Evaluations Flaw Evaluation RAI-3 Considering the difference in Young's modulus (E) values of the low alloy steel and the 182 weld and that the "local" weld residual stresses may not be adequately represented by the "averaging" membrane and bending stresses, please demonstrate that your approach will not severely underestimate the effect of residual stresses on the applied stress intensity factor.
 
===Response===
The difference in Young's modulus between the Alloy 182 weld material and the SA-302B (modified) low alloy steel (LAS), is reasonably small, and is not expected to have any significant effect on the material behavior in the elastic regime. Weld residual stresses are steady state secondary stresses, which are generally limited by yield strength. The Alloy 182 and LAS materials have similar yield strengths and the difference will not Significantly affect the weld residual stress distribution. Since the vessel and 3/16 inch of the original weld were stress relieved during original fabrication, weld residual stresses were significantly relieved in those materials. The rewelding of the J-groove weld produced tensile residual stresses in the weld at roughly the yield level at temperature, and for equilibrium, the stresses in the adjacent LAS material will be slightly compressive and the stresses in the original Alloy 182 weld will be further relieved, possibly to the point of being compressive. In the present analysis, the crack tip is assumed to be within the LAS material, so the residual stresses assumed are conservative for the present analysis. Furthermore, assuming an average compressive stress in the LAS material will underestimate the compressive stresses near the crack front, resulting in a conservative analysis.
Flaw Evaluation RAI-4 Section 2.1 of the flaw evaluation report described the Raju-Newman stress intensity factor model, but Section 3.3's description is not clear: "The SIF solution used is based on a corner flaw in a flat plate under remote bending and tension." Please confirm that the flaw evaluation is based on the Raju-Newman model without any further simplifications.
Flaw Evaluation RAI-4 Section 2.1 of the flaw evaluation report described the Raju-Newman stress intensity factor model, but Section 3.3's description is not clear: "The SIF solution used is based on a corner flaw in a flat plate under remote bending and tension." Please confirm that the flaw evaluation is based on the Raju-Newman model without any further simplifications.
Response Yes, the flaw evaluation is based on the Raju-Newman model without any further simplifications.
 
Additionally, stresses due to weld residual stresses are estimated and superimposed with the stresses due to operating transients.
===Response===
The analysis also considers the applicable pressure on the crack face at the particular transient time point being evaluated.
Yes, the flaw evaluation is based on the Raju-Newman model without any further simplifications.
Page 3 of 4 ATTACHMENT 1 Response to Request for Additional Information Regarding Relief Request 14R-19 Corrosion and Flaw Evaluations Flaw Evaluation RAI-5 Section 4.2 specified a certain crack depth increment to account for any potential flaw growth without performing an actual flaw growth analysis.
Additionally, stresses due to weld residual stresses are estimated and superimposed with the stresses due to operating transients. The analysis also considers the applicable pressure on the crack face at the particular transient time point being evaluated.
Please justify this assumed crack depth increment by providing an estimated crack growth within one cycle using appropriate fatigue crack and stress corrosion cracking growth rates (BWRVIP-60), albeit it is a small value. Response The stress corrosion cracking (SCC) growth rate in the steam space region where the instrument nozzle N-11B is located is expected to be negligible as discussed on Page 5-5 of BWRVIP-60A.
Page 3 of 4
Fatigue crack growth is also considered to be negligible over a two year period. Conservatively, per BWRVIP-60A, Page 8-1, the upper bound crack growth rate in the low alloy steel shell provided is 2.83 x 10-6 in/hr. This rate is applicable for the BWR water environment at 550 OF which corresponds to the steady state operating condition.
 
For the next two year fuel cycle period, this corresponds to 0.0496 inches of stress corrosion crack growth. Hence, a value of 1/16" or 0.0625 inches is conservative and allows for the potential growth rate due to SCC and fatigue in the low alloy steel vessel material.
ATTACHMENT 1 Response to Request for Additional Information Regarding Relief Request 14R-19 Corrosion and Flaw Evaluations Flaw Evaluation RAI-5 Section 4.2 specified a certain crack depth increment to account for any potential flaw growth without performing an actual flaw growth analysis. Please justify this assumed crack depth increment by providing an estimated crack growth within one cycle using appropriate fatigue crack and stress corrosion cracking growth rates (BWRVIP-60), albeit it is a small value.
REFERENCES  
 
: 1. Letter from P. R. Simpson (Exelon Generation Company, LLC) to NRC, "Relief Request 14R-19 Associated with the Reactor Pressure Vessel Nozzle Repairs," dated April 6, 2012 2. Letter from D. M. Gullott (Exelon Generation Company, LLC) to NRC, "Response to Request for Additional Information Regarding Relief Request 14R-19 Associated with the Reactor Pressure Vessel Nozzle Repairs," dated April 12, 2012 3. Email fromJoeIWiebe(NRC)toD.M.Guliott (Exelon Generation Company, LLC), "Quad Cities Unit 2 Relief Request 14R-19 -Followup Question to Response to RAI Question 8," dated April 13, 2012 4. Email from Joel Wiebe (NRC) to D. M.Guliott (Exelon Generation Company, LLC), "Quad Cities Unit 2 Relief Request 14R-19 -Followup Question to Regarding the Flaw Evaluation," dated April 13,2012 Page 4 of 4}}
===Response===
The stress corrosion cracking (SCC) growth rate in the steam space region where the instrument nozzle N-11B is located is expected to be negligible as discussed on Page 5-5 of BWRVIP-60A. Fatigue crack growth is also considered to be negligible over a two year period.
Conservatively, per BWRVIP-60A, Page 8-1, the upper bound crack growth rate in the low alloy steel shell provided is 2.83 x 10-6 in/hr. This rate is applicable for the BWR water environment at 550 OF which corresponds to the steady state operating condition. For the next two year fuel cycle period, this corresponds to 0.0496 inches of stress corrosion crack growth. Hence, a value of 1/16" or 0.0625 inches is conservative and allows for the potential growth rate due to SCC and fatigue in the low alloy steel vessel material.
REFERENCES
: 1. Letter from P. R. Simpson (Exelon Generation Company, LLC) to NRC, "Relief Request 14R-19 Associated with the Reactor Pressure Vessel Nozzle Repairs," dated April 6, 2012
: 2. Letter from D. M. Gullott (Exelon Generation Company, LLC) to NRC, "Response to Request for Additional Information Regarding Relief Request 14R-19 Associated with the Reactor Pressure Vessel Nozzle Repairs," dated April 12, 2012
: 3. Email fromJoeIWiebe(NRC)toD.M.Guliott (Exelon Generation Company, LLC), "Quad Cities Unit 2 Relief Request 14R Followup Question to Response to RAI Question 8,"
dated April 13, 2012
: 4. Email from Joel Wiebe (NRC) to D. M.Guliott (Exelon Generation Company, LLC), "Quad Cities Unit 2 Relief Request 14R Followup Question to Regarding the Flaw Evaluation,"
dated April 13,2012 Page 4 of 4}}

Latest revision as of 05:55, 12 November 2019

Response to Request for Additional Information Regarding Relief Request I4R-19, Corrosion and Flaw Evaluations
ML12107A361
Person / Time
Site: Quad Cities Constellation icon.png
Issue date: 04/13/2012
From: Gullott D
Exelon Generation Co, Exelon Nuclear
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-12-074
Download: ML12107A361 (6)
v • d • e


Text

Exelon Generation Company, LLC www.exeloncorp.com 4300 Winfield Road Nuclear Warrenville, IL 60555 RS-12-074 10 CFR 50.55a April 13, 2012 U. S. Nuclear Regulatory Commission A TIN: Document Control Desk Washington, D.C. 20555-0001 Quad Cities Nuclear Power Station, Unit 2 Renewed Facility Operating License No. DPR-30 NRC Docket No. 50-265

Subject:

Response to Request for Additional Information Regarding Relief Request 14R-19, Corrosion and Flaw Evaluations

References:

1) Letter from P. R. Simpson (Exelon Generation Company, LLC) to NRC, "Relief Request 14R-19 Associated with the Reactor Pressure Vessel Nozzle Repairs,"

dated April 6, 2012

2) Letter from D. M. Gullott (Exelon Generation Company, LLC) to NRC, "Response to Request for Additional Information Regarding Relief Request 14R-19 Associated with the Reactor Pressure Vessel Nozzle Repairs," dated April 12, 2012
3) Email fromJoeIWiebe(NRC)toD.M.Guliott (Exelon Generation Company, LLC), "Quad Cities Unit 2 Relief Request 14R Followup Question to Response to RAI Question 8," dated April 13, 2012
4) Email fromJoeIWiebe(NRC)toD.M.Guliott (Exelon Generation Company, LLC), "Quad Cities Unit 2 Relief Request 14R Followup Question to Regarding the Flaw Evaluation," dated April 13, 2012 In Reference 1, in accordance with 10 CFR 50.55a, "Codes and standards," paragraph (a)(3)(i),

Exelon Generation Company, LLC (EGC), requested NRC approval of a relief request associated with the Fourth Inservice Inspection (lSI) Interval for Quad Cities Nuclear Power Station (QCNPS), Unit 2. Note that the fourth interval of the QCNPS lSI program complies with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, 1995 Edition with addenda through 1996.

During review of the subject relief request, the NRC concluded that additional information would be needed to complete their review. In Reference 2, EGC provided a portion of the requested

U. S. Nuclear Regulatory Commission April 13, 2012 Page 2 information. In Reference 3, the NRC requested follow-up information to supplement the information provided in Reference 2 regarding the Corrosion Evaluation (Le., the information provided in Request for Additional Information (RAI) Question 8). In Reference 4, the NRC requested follow-up information to supplement the information provided in Reference 2 regarding the Flaw Evaluation (Le., the information provided in RAI Question 9). The follow-up information requested in References 3 and 4 is provided in Attachment 1 to this letter.

There are no regulatory commitments contained in this letter.

Should you have any questions concerning this letter, please contact Mr. Joseph A. Bauer at (630) 657-2804.

Respectfully, David M. Gullott Manager - Licensing Exelon Generation Company, LLC : Response to Request for Additional Information Regarding Relief Request 14R-19, Corrosion and Flaw Evaluations

ATTACHMENT 1 Response to Request for Additional Information Regarding Relief Request 14R-19 Corrosion and Flaw Evaluations In Reference 1, in accordance with 10 CFR 50.55a, "Codes and standards," paragraph (a)(3)(i),

Exelon Generation Company, LLC (EGC), requested NRC approval of a relief request associated with the Fourth Inservice Inspection (lSI) Interval for Quad Cities Nuclear Power Station (QCNPS), Unit 2. Note that the fourth interval of the QCNPS lSI program complies with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, 1995 Edition with addenda through 1996.

During review of the subject relief request, the NRC concluded that additional information would be needed to complete their review. In Reference 2, EGC provided a portion of the requested information. In Reference 3, the NRC requested follow-up information to supplement the information provided in Reference 2 regarding the Corrosion Evaluation (Le., the information provided in Request for Additional Information (RAI) Question 8). In Reference 4, the NRC requested follow-up information to supplement the information provided in Reference 2 regarding the Flaw Evaluation (Le., the information provided in RAI Question 9). The follow-up information requested in References 3 and 4 is provided below.

Corrosion Evaluation RAI 1 With respect to general and galvanic corrosion of low-alloy steel, the licensee's corrosion evaluation, provided as Attachment 2 to Letter RS-12-069 dated April 12, 2012, references "NECD-21120, "Monticello Feedwater Nozzle Cladding Crack Repair Report,"

as the basis for low general corrosion rates and the absence of galvanic corrosion. The licensee's corrosion evaluation references "D.C. Vreeland, et. AI., "Corrosion of Carbon and Low-Alloy Steels in Out-of-Pile Boiling Water Reactor Environment," Corrosion, Vol. 17, No.6, 1961, as the basis for minimal crevice corrosion of low alloy steel in a bOiling water reactor (BWR) operating environment.

Requested Information To ensure these two references provide bounding corrosion rates for Quad Cities, Unit 2, compare the chemistry of the steam or water to which the low-alloy steel in the repaired N-11B nozzle will be exposed (in the gap between the nozzle remnant and new nozzle) to the chemistry used in the tests described in the two references discussed above. The specific parameters to be addressed should include the concentration of dissolved oxygen and impurities such as chloride and sulfate, conductivity, hydrogen, electrochemical corrosion potential (ECP), and any other parameters that may affect corrosion rates. Values of the parameters for Quad Cities may be typical or average values or maximum values. If the steam/water environment tests documented in the references do not envelop or bound the actual conditions to which the low-alloy steel in Nozzle N-11 B will be exposed, provide a justification for the use of these reference to support the prediction of minimal corrosion for the exposed low-alloy steel, or provide additional references.

Response

The corrosion evaluation (51-9180975-001) references NEDC-21120 regarding general and galvanic corrosion. The corrosion rate from NEDC-21120 is based on tests performed with low Page 1 of 4

ATTACHMENT 1 Response to Request for Additional Information Regarding Relief Request 14R-19 Corrosion and Flaw Evaluations alloy steel in contact with BWR water conditions and BWR steam conditions. The corrosion rate used in 51-9180975-001 is 70% faster than the maximum corrosion rate than all of the tests from NEDC-21120. NEDC-21120 does not provide specific water chemistry parameters.

NEDC-21120 was written in 1975; therefore, it can be inferred that the BWR water chemistry used in the tests was consistent with the water chemistry requirements for BWRs of 1975 or before. The water chemistry requirements of this time period are known to be much less stringent than today's modern requirements (Le., BWRVIP-130 and more recently BWRVIP-190). BWRVI P-130 was created to better control BWR water chemistry and is significantly more restrictive than the requirements of the 1970s (and before). Quad Cities Unit 2 reactor water meets or exceeds the requirements of BWRVIP-190, which supersedes BWRVIP-130.

BWRVIP-190 has requirements that are equal to or more restrictive than BWRVIP-130. Based on this information, it is concluded that the water chemistry at Quad Cities Unit 2 is less aggressive than the water chemistry used in the tests discussed in NEDC-21120.

The corrosion evaluation (51-9180975-001) references Vreeland (1961) regarding crevice corrosion. For the same reasoning given above, it is concluded that the water chemistry at Quad Cities Unit 2 is less aggressive than the water chemistry used in the tests discussed in Vreeland (1961).

Flaw Evaluation RAI-1 The flaw evaluation report states that heatup and cooldown, SRV blowdown, and SCRAM are determined to be controlling for the current flaw evaluations. Please provide the cTiteria and the qualitative assessment that you have performed to determine that SRV blowdown and SCRAM are the controlling emergencyHaulted and normal/upset transients among all design transients.

Response

The bounding transients were determined from the Section III design analysis of record. The transients that have significant temperature excursions, and therefore, significant stress ranges, are heatup and cooldown, SCRAM, and SRV Blowdown. Therefore, these are the transients that lead to the highest stress ranges, and they are controlling for the flaw evaluation.

Flaw Evaluation RAI-2 Confirm that the crack plane is assumed to exist at a circumferential location with respect to the nozzle hole such that the maximum remote membrane and bending stresses in Reference 5 were used for the subsequent fracture mechanics analysis.

Justify your selection of the crack plane location if this is not the case.

Response

Yes, the crack plane is assumed to exist at a circumferential location with respect to the nozzle hole such that the maximum remote membrane and bending stresses are applied. The stresses used are normal to the axial plane of the vessel, where the hoop stresses are at a maximum.

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ATTACHMENT 1 Response to Request for Additional Information Regarding Relief Request 14R-19 Corrosion and Flaw Evaluations Flaw Evaluation RAI-3 Considering the difference in Young's modulus (E) values of the low alloy steel and the 182 weld and that the "local" weld residual stresses may not be adequately represented by the "averaging" membrane and bending stresses, please demonstrate that your approach will not severely underestimate the effect of residual stresses on the applied stress intensity factor.

Response

The difference in Young's modulus between the Alloy 182 weld material and the SA-302B (modified) low alloy steel (LAS), is reasonably small, and is not expected to have any significant effect on the material behavior in the elastic regime. Weld residual stresses are steady state secondary stresses, which are generally limited by yield strength. The Alloy 182 and LAS materials have similar yield strengths and the difference will not Significantly affect the weld residual stress distribution. Since the vessel and 3/16 inch of the original weld were stress relieved during original fabrication, weld residual stresses were significantly relieved in those materials. The rewelding of the J-groove weld produced tensile residual stresses in the weld at roughly the yield level at temperature, and for equilibrium, the stresses in the adjacent LAS material will be slightly compressive and the stresses in the original Alloy 182 weld will be further relieved, possibly to the point of being compressive. In the present analysis, the crack tip is assumed to be within the LAS material, so the residual stresses assumed are conservative for the present analysis. Furthermore, assuming an average compressive stress in the LAS material will underestimate the compressive stresses near the crack front, resulting in a conservative analysis.

Flaw Evaluation RAI-4 Section 2.1 of the flaw evaluation report described the Raju-Newman stress intensity factor model, but Section 3.3's description is not clear: "The SIF solution used is based on a corner flaw in a flat plate under remote bending and tension." Please confirm that the flaw evaluation is based on the Raju-Newman model without any further simplifications.

Response

Yes, the flaw evaluation is based on the Raju-Newman model without any further simplifications.

Additionally, stresses due to weld residual stresses are estimated and superimposed with the stresses due to operating transients. The analysis also considers the applicable pressure on the crack face at the particular transient time point being evaluated.

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ATTACHMENT 1 Response to Request for Additional Information Regarding Relief Request 14R-19 Corrosion and Flaw Evaluations Flaw Evaluation RAI-5 Section 4.2 specified a certain crack depth increment to account for any potential flaw growth without performing an actual flaw growth analysis. Please justify this assumed crack depth increment by providing an estimated crack growth within one cycle using appropriate fatigue crack and stress corrosion cracking growth rates (BWRVIP-60), albeit it is a small value.

Response

The stress corrosion cracking (SCC) growth rate in the steam space region where the instrument nozzle N-11B is located is expected to be negligible as discussed on Page 5-5 of BWRVIP-60A. Fatigue crack growth is also considered to be negligible over a two year period.

Conservatively, per BWRVIP-60A, Page 8-1, the upper bound crack growth rate in the low alloy steel shell provided is 2.83 x 10-6 in/hr. This rate is applicable for the BWR water environment at 550 OF which corresponds to the steady state operating condition. For the next two year fuel cycle period, this corresponds to 0.0496 inches of stress corrosion crack growth. Hence, a value of 1/16" or 0.0625 inches is conservative and allows for the potential growth rate due to SCC and fatigue in the low alloy steel vessel material.

REFERENCES

1. Letter from P. R. Simpson (Exelon Generation Company, LLC) to NRC, "Relief Request 14R-19 Associated with the Reactor Pressure Vessel Nozzle Repairs," dated April 6, 2012
2. Letter from D. M. Gullott (Exelon Generation Company, LLC) to NRC, "Response to Request for Additional Information Regarding Relief Request 14R-19 Associated with the Reactor Pressure Vessel Nozzle Repairs," dated April 12, 2012
3. Email fromJoeIWiebe(NRC)toD.M.Guliott (Exelon Generation Company, LLC), "Quad Cities Unit 2 Relief Request 14R Followup Question to Response to RAI Question 8,"

dated April 13, 2012

4. Email from Joel Wiebe (NRC) to D. M.Guliott (Exelon Generation Company, LLC), "Quad Cities Unit 2 Relief Request 14R Followup Question to Regarding the Flaw Evaluation,"

dated April 13,2012 Page 4 of 4

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