AEP-NRC-2015-25, Response to RAI Regarding License Amendment Request to Revise Technical Specification 5.5.14, Containment Leakage Rate Testing Program, to Use NEI-94-01, Revision 3-A as Regulatory Guidance Versus Current NEI-94-01, Rev. 0
| ML15055A048 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 02/20/2015 |
| From: | Gebbie J Indiana Michigan Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| AEP-NRC-2015-25 | |
| Download: ML15055A048 (9) | |
Text
z INDIANA Indiana Michigan Power MICHIGAN Cook Nuclear Plant POWER One Cook Place Bridgman, MI 49106 A unit of American Electric Power Indiana Michigan Power.com February 20, 2015 AEP-NRC-2015-25 10 CFR 50.90 Docket Nos.: 50-315 50-316 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Donald C. Cook Nuclear Plant Units 1 and 2 Response to a Request for Additional Information Regarding the License Amendment Request to Revise Technical Specification 5.5.14, "Containment Leakage Rate Testing Program," to Use NEI-94-01, Revision 3-A as Regulatory Guidance Versus the Current NEI 94-01, Revision 0
References:
- 1. Letter from J. P. Gebbie, Indiana Michigan Power Company (I&M), to U. S. Nuclear Regulatory Commission (NRC), "Donald C. Cook Nuclear Plant Units 1 and 2, License Amendment Request to Revise Technical Specification 5.5.14, 'Containment Leakage Rate Testing Program,"' AEP-NRC-2014-09, dated March 7, 2014, Agencywide Documents Access and Management System (ADAMS) Accession No. ML14071A435.
- 2. Email from T. A. Beltz, NRC, to H. L. Etheridge, I&M, "Draft Requests for Additional Information Mechanical & Civil Engineering Branch of the Office of Nuclear Reactor Regulation Regarding a License Amendment Request for the Donald C. Cook Nuclear Plant, Units 1 and 2, to Revise Technical Specification Section 5.5.14, "Containment Leakage Rate Testing Program," Indiana Michigan Power Company Docket Nos. 50-315 and 50-316 (TAC Nos. MF3568 and MF3569),"
dated August 25, 2014.
- 3. Email from T. A. Beltz, NRC, to H. L. Etheridge, I&M, "Draft Requests for Additional Information Mechanical & Civil Engineering Branch of the Office of Nuclear Reactor Regulation Regarding a License Amendment Request for the Donald C. Cook Nuclear Plant, Units 1 and 2, to Revise Technical Specification Section 5.5.14, "Containment Leakage Rate Testing Program" Indiana Michigan Power Company Docket Nos. 50-315 and 50-316 (TAC Nos. MF3568 and MF3569),"
dated August 26, 2014.
- 4. Letter from J. P. Gebbie, I&M, to NRC, "Donald C. Cook Nuclear Plant Units 1 and 2 Response to a Request for Additional Information Regarding the License Amendment Request to Revise Technical Specification 5.5.14, "Containment Leakage Rate Testing Program," to use NEI-94-01, Revision 3-A as Regulatory Guidance Versus the Current NEI 94-01, Revision 0,"
AEP-NRC-2014-73, dated September 30, 2014, ADAMS Accession No. ML14275A454.
U. S. Nuclear Regulatory Commission AEP-NRC-2015-25 Page 2
- 5. Letter from J. P. Gebbie, I&M, to NRC, "Donald C. Cook Nuclear Plant Units 1 and 2 Supplemental Response to a Request for Additional Information Regarding the License Amendment Request to Revise Technical Specification 5.5.14, "Containment Leakage Rate Testing Program," to use NEI-94-01, Revision 3-A as Regulatory Guidance Versus the Current NEI 94-01, Revision 0," AEP-NRC-2014-100, dated December 16, 2014, ADAMS Accession No. ML14352A232.
- 6. Email from M. L. Chawla, NRC, to H. L. Etheridge, I&M, "Request for Additional Information -
Donald C. Cook Units 1 and 2 - LAR - To Revise Technical Specification 5.5.14, "Containment Leakage Rate Testing Program" - MF3568 and MF3569," dated November 20, 2014.
- 7. Letter from J. P. Gebbie, I&M, to NRC, "Donald C. Cook Nuclear Plant Units 1 and 2 Response to a Request for Additional Information Regarding the License Amendment Request to Revise Technical Specification 5.5.14, "Containment Leakage Rate Testing Program," to use NEI-94-01, Revision 3-A as Regulatory Guidance Versus the Current NEI 94-01, Revision 0,"
AEP-NRC-2015-07, dated January 15, 2015, ADAMS Accession No. ML15020A662.
- 8. Email from M. L. Chawla, NRC, to H. L. Etheridge, I&M, "Request for Additional Information (RAI) - Donald C. Cook Nuclear Plant Units 1 and 2 - LAR to Revise TS 5.5.14, "Containment Integrated Leak Rate Testing" - MF3568/69," dated February 3, 2015.
This letter provides Indiana Michigan Power Company's (I&M), the licensee for Donald C. Cook Nuclear Plant Units 1 and 2, response to a Request for Additional Information (RAI) by the U. S. Nuclear Regulatory Commission (NRC) regarding a license amendment request to change Technical Specification (TS) Section 5.5.14, "Containment Leakage Rate Testing Program."
By Reference 1, I&M submitted a request to amend the TS to Facility Operating Licenses DPR-58 and DPR-74. I&M proposes to change TS 5.5.14, "Containment Leakage Rate Testing Program,"
to use Nuclear Energy Institute (NEI) 94-01, Revision 3-A as regulatory guidance versus the current NEI 94-01, Revision 0. By Reference 2, the NRC transmitted RAIs (RAI-EMCB-1,-2,-3, and -4) regarding the proposed amendment.
By Reference 3, the NRC transmitted an additional RAI (RAI-SCVB-1) regarding the proposed amendment. References 4 and 5 provided I&M's response to References 2 and 3.
By Reference 6, the NRC transmitted an additional RAI regarding the proposed amendment. Reference 7 provided I&M's response to Reference 6. By Reference 8, the NRC transmitted an additional RAI regarding the proposed amendment. This letter provides I&M's response to Reference 8. Enclosure 1 to this letter provides an affirmation statement. Enclosure 2 to this letter provides I&M's response to the NRC's RAI contained in Reference 8.
Copies of this letter and its enclosures are being transmitted to the Michigan Public Service Commission and Michigan Department of Environmental Quality, in accordance with the requirements of 10 CFR 50.91. There are no new or revised commitments made in this letter.
U. S. Nuclear Regulatory Commission AEP-NRC-2015-25 Page 3 Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Manager, at (269) 466-2649.
Sincerely, Joel P. Gebbie Site Vice President JMT/amp
Enclosures:
- 1. Affirmation
- 2. Response to Request for Additional Information Regarding Technical Specification Section 5.5.14, "Containment Leakage Rate Testing Program" c:
M. L. Chawla, NRC Washington, D.C.
J. T. King, MPSC MDEQ - RMD/RPS NRC Resident Inspector C. D. Pederson, NRC Region III A. J. Williamson, AEP Ft. Wayne, w/o enclosures
Enclosure I to AEP-NRC-2015-25 AFFIRMATION I, Joel P. Gebbie, being duly sworn, state that I am the Site Vice President of Indiana Michigan Power Company (I&M), that I am authorized to sign and file this request with the U. S. Nuclear Regulatory Commission on behalf of I&M, and that the statements made and the matters set forth herein pertaining to I&M are true and correct to the best of my knowledge, information, and belief.
Indiana Michigan Power Company Joel P. Gebbie Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME THIS'&)
DAY OF *--2015 My CommisNotanr--blic My Conmmission Expires Q)
- Q._*%
- A'-)\\*
DANIELLE BURGOYNE Notary Public, State of Michigan County of Berrien My Commission Expires 04-04-2018 Acting in the county of *'I2
- -..\\
to AEP-NRC-2015-25 Response to Request for Additional Information Regarding Technical Specification Section 5.5.14, "Containment Leakage Rate Testing Program" By letter dated March 7, 2014 (Agencywide Documents Access and Management System (ADAMS)
Accession No.
as supplemented by letters dated September 30, 2014 (ADAMS Accession No. ML14275A454), December 16, 2014 (ADAMS Accession No. ML14352A232), and January 15, 2015 (ADAMS Accession No. ML15020A662),
Indiana Michigan Power Company (I&M), submitted a license amendment request (LAR) for a permanent extension to the Type A (Integrated Leak Rate Test (ILRT)) and Type C (isolation valve Local Leakage Rate Test (LLRT)) by revising Technical Specification (TS) 5.5.14, for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2. The proposed change would permit the existing Containment ILRT (Type A) frequency to be extended from 10 years to 15 years and Type C testing frequency to be extended up to 75 months on a permanent basis.
The U. S. Nuclear Regulatory Commission (NRC) staff has reviewed the LAR and has identified areas where additional information is needed to complete its review. There are two sets of requests for additional information (RAI) here. The first RAI is from Containment and Ventilation Branch (SCVB) and the second RAI is from probabilistic risk assessments (PRA) licensing branch.
By electronic mail dated February 3, 2015, the NRC transmitted a RAI regarding the March 7, 2014 LAR. The text of the RAIs and I&M's responses are provided below.
SCVB-RAI-1 In the licensee's submittal (March 7, 2014) it states, in section 4.3 "Type B and Type C Testing Programs", that U2C20 As-Found Minimum Pathway Leakage Rate had exceed the TS performance criterion value of 0.6 La (66,131 sccm) for the combined Type B and Type C test as found minimum pathway total. The reasoning given was that:
In that case, an evaluation was conducted and work was performed to replace the hinge ring retaining dowel pins which were determined to be too short.
The staff would like additional information regarding:
a) The as-found minimum pathway leakage rate for U2C20 (April 2012).
b) The component that resulted in the exceeding the combined Type B and Type C leakage TS value of (66,131 sccm).
c) Whether the containment past operability had been determined; if so, please explain the basis for the conclusion.
d) Whether this component falls under the Containment Maintenance Rule function (10 CFR. 50.65) and if so what impact did the failure have.
e) The component's previous failures, if any, and any actions taken to prevent recurrence.
to AEP-NRC-2015-25 Page 2 I&M Response to SCVB-RAI-1 a) The as-found minimum pathway leakage rate for U2C20 (April 2012). The total containment "As-Found" Minimum Path Leak Rate (MNPLR) for U2C20 was 11,286 standard cubic centimeters per minute (sccm) excluding penetration 2-CPN-1 5.
Penetration 2-CPN-15 containment barriers consist of a closed system outside containment and a check valve (2-SI-189) inside containment.
Only the check valve containment barrier on this penetration is LLRT tested. The results of the check valve, 2-SI-189, LLRT are considered both the MNPLR and Maximum Path Leak Rate. Since valve 2-SI-1 89 did not fully pressurize to accident pressure (12 pounds per square inch gage (psig) for CNP), the LLRT result is greater than leakage allowable (La).
This results in the total containment "As-Found" MNPLR for U2C20 being 11,286 sccm + leakage for 2-SI-189 (which was > La).
La for each CNP unit is 110,219 sccm.
b) The component that resulted in the exceeding the combined Type B and Type C leakage TS value of (66,131 sccm). 2-SI-189, "ECCS Safety Valves Discharge Header to Pressurizer Relief Tank Containment Isolation Check Valve," is the component that resulted in exceeding the combined Type B and Type C leakage TS value.
Valve 2-SI-189 is downstream of six safety valves located outside containment which discharge to the Pressurizer Relief Tank, which is located inside containment. The six reliefs consist of the following 1:
2-SV-56, Charging Pumps suction safety 2-SV-104W, residual heat removal (RHR) heat exchanger (HX) outlet safety 2-SV-104E, RHR HX outlet safety 2-SV-98S, South Safety Injection Pump discharge safety 2-SV-98N, North Safety Injection Pump discharge safety 2-SV-96, Safety Injection Pumps suction safety c) Whether the containment past operability had been determined; if so, please explain the basis for the conclusion. The past operability of containment was documented in the CNP corrective action program in Action Request 2012-5255. The conclusion was that containment remained OPERABLE during all modes of applicability over the timeframe of concern (mode 1-4 between U2C19 and U2C20). The basis for the conclusion was documented as follows:
"Since the ECCS safety valves and the closed system piping to containment penetration 2-CPN-15 will remain intact during design bases accidents, including seismic events, the potential containment atmosphere leakage through this penetration would have been prevented.
Containment remained OPERABLE during all modes of applicability over the entire time frame of concern."
The basis for the conclusion was that the second barrier on 2-CPN-15 remained intact and was not breached during the timeframe that containment integrity was required.
The CNP work control system was searched for any deficiencies impacting the components of the closed system on 2-CPN-15 and no deficiencies were identified, 1 See Figure 1 included at the end of Enclosure 2.
to AEP-NRC-2015-25 Page 3 which supports this determination.
In accordance with CNP engineering specification ES-CIV-0306-QCN, "Containment Isolation System Licensing
/
Design Basis Requirements" and Updated Final Safety Analysis Report Section 5.4, Table 5.4-1, 2-CPN-15 has two barriers. One barrier is the closed piping system and the second barrier is the containment isolation valve 2-SI-189. We only perform an LLRT on valve 2-SI-189 and this test result is conservatively assumed to be equal to the MNPLR per the Unit 2 Primary Containment Leak Rate Running Total procedure.
d) Whether this component falls under the Containment Maintenance Rule function (10 CFR. 50.65) and if so what impact did the failure have. When this failure occurred in 2012, a Maintenance Rule Evaluation was documented in the CNP corrective action program, Action Request 2012-5255. The conclusion of the evaluation was based on the program criteria that the functional failure definition includes Type B testing only. Consequently, the LLRT failure of valve 2-SI-189 (Type C leak test) was outside the scope of Maintenance Rule function for containment, which addresses Type B leak tests. Therefore, this leakage did not impact the Containment Maintenance Rule function and Expert Panel (a)(1) consideration was not necessary. Since 2012, CNP has revised the Containment Structure / System Maintenance Rule Scoping Document.
The current revision states that the Appendix J program monitors penetration leakage testing in all modes when containment integrity is required.
The Appendix J program requires investigation and correction of any failures to meet the leakage acceptance criteria in accordance with the CNP corrective action program. Any Maintenance Rule functional failure definitions for penetration leakage are considered redundant to the Appendix J requirements.
Therefore, Appendix J leakage is not monitored by Maintenance Rule criteria.
e) The component's previous failures, if any, and any actions taken to prevent recurrence. Valve 2-SI-1 89 is LLRT tested each outage. The only previous failure was in the 2006 Unit 2 refueling outage (U2C16) when 2-SI-189 failed to pressurize during LLRT.
During U2C16 an "As-Found" LLRT condition was not obtained for valve 2-SI-189 because the check valve was incorrectly opened and inspected prior to the work task for the "As-Found" LLRT due to a scheduling error. The "As-Found" LLRT failure was documented in the CNP corrective action program in Action Request 06100045. The valve internals were removed and reinstalled as part of the open and inspect work order task. After the open and inspect task and the subsequent "As-Found" LLRT failure, the valve was reopened and it was found that the seat contact was unacceptable with feeler gage measurement.
In 2006, it was documented in the maintenance work order task that it appeared that the dowel pin should be longer to ensure proper engagement. Subsequently, it was identified that the vendor for the valve had revised the design to recommend the use of longer dowel pins. This anti-rotation dowel pin length was corrected in 2012 following the second failure of valve 2-SI-189 for the "As-Found" LLRT.
APLA-RAI-I In response to the request for additional information dated January 15, 2015 (ADAMS Accession No. ML15020A662), I&M performed an assessment of the PRA against the PRA standard ASME RA-Sa-2003 and indicated that eleven supporting requirements (SR) were to AEP-NRC-2015-25 Page 4 determined to not be met at the Capability Category (CC) I level according to the 2003 ASME PRA Standard.
- a. I&M provided justification of no impact for ten of the eleven SRs not met at CC I level.
Clarify if there are any missing requirements not listed in the RAI response that do not meet CC I and, if so, justify why not meeting each CC I requirement will have no impact on the ILRT extension application.
- b. If PRA is assessed against Revision 2 of Regulatory Guide (RG) 1.200 and the 2009 PRA standard ASME/ANS RA-Sa-2009, explain whether there are any additional SRs that are not met at CC I level that would impact the ILRT extension application.
I&M Response to APLA-RAI-1
- a. I&M provided justification of no impact for ten of the eleven SRs not met at CC I level. Clarify if there are any missing requirements not listed in the RAI response that do not meet CC I and, if so, justify why not meeting each CC I requirement will have no impact on the ILRT extension application. I&M's January 15, 2015 response to the NRC RAI concluded that there were 11 Summary Reports (SR) not met at Capability Category (CC) I, and provided 10 paragraphs discussing why these issues were of no impact to the ILRT extension application. One of these paragraphs contained a combined explanation for two (SR SC-C2 & SC-C3) of the 11 SRs.
- b. If PRA is assessed against Revision 2 of Regulatory Guide (RG) 1.200 and the 2009 PRA standard ASME/ANS RA-Sa-2009, explain whether there are any additional SRs that are not met at CC I level that would impact the ILRT extension application. I&M did not directly review the CNP PRA model against the 2003 American Society of Mechanical Engineers (ASME) PRA Standard (ASME-RA-Sa-2003 as described in Regulatory Guide (RG) 1.200, Revision 1) CC I SRs.
The January 15, 2015 I&M response was based on a gap assessment of the CNP PRA model against ASME/American Nuclear Society RA-Sa-2009 Standard SRs for CC II capability, including RG 1.200 Revision 2 clarifications. Based on this gap assessment if the model had no gap to the 2009 Standard CC II for an SR it was considered acceptable with regard to the comparable 2003 Standard SR CC I. SRs identified with gaps to the 2009 CC II requirements were evaluated against the comparable 2003 SR CC I requirements.
Those 11 SRs which did not meet the 2003 Standard CC I requirements were provided in the I&M January 15, 2015, response, along with justifications of no impact on the ILRT extension application.
to AEP-NRC-2015-25 Page 5 HEAT EXCH. SAFETY VALVE I 2-SV-104E and 2-SV-104W Inside Containment I
EMERG. CORE COQLINq
-CHARGIRN4 PUMP SUCTION SAFETY VALVE DISCH."
I 6*V-56 FROM RESIDUAL 4 LOOP SEE DWG 51!
I FISOLATION r lie VALVE TEST 4
CONECTION Figure 1:
Excerpt from Operating Flow Diagram with valve numbers added