ML23228A025
| ML23228A025 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 09/25/2023 |
| From: | Mahesh Chawla Plant Licensing Branch IV |
| To: | Diya F Ameren Missouri, Union Electric Co |
| Chawla M | |
| References | |
| EPID L-2022-LLA-0165 | |
| Download: ML23228A025 (1) | |
Text
September 25, 2023 Mr. Fadi Diya Senior Vice President and Chief Nuclear Officer Ameren Missouri Callaway Energy Center 8315 County Road 459 Steedman, MO 65077
SUBJECT:
CALLAWAY PLANT, UNIT NO. 1 - ISSUANCE OF AMENDMENT NO. 234 TO REVISE TECHNICAL SPECIFICATION 5.5.16, CONTAINMENT LEAKAGE RATE TESTING PROGRAM, FOR PERMANENT EXTENSION OF TYPE A AND TYPE C LEAK RATE TEST FREQUENCIES (EPID L-2022-LLA-0165)
Dear Mr. Diya:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 234 to Renewed Facility Operating License No. NPF-30 for the Callaway Plant, Unit No. 1 (Callaway). The amendment consists of changes to the technical specifications (TSs) in response to your application dated November 3, 2022, as supplemented by letters dated June 5, July 25, and August 15, 2023.
The amendment revises Callaway TS 5.5.16, Containment Leakage Rate Testing Program, by replacing the existing reference to Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, with a reference to Nuclear Energy Institute (NEI) Topical Report (TR) 94 01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR [Title 10 of the Code of Federal Regulations] Part 50, Appendix J [Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors], dated July 2012, and the limitations and conditions specified in TR NEI 94-01, Revision 2-A, dated October 2008, as the documents used to implement the performance-based containment leakage testing program in accordance with Option B of 10 CFR Part 50, Appendix J.
A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Mahesh L. Chawla, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-483
Enclosures:
- 1. Amendment No. 234 to NPF-30
- 2. Safety Evaluation cc: Listserv
UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT NO. 1 DOCKET NO. 50-483 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 234 License No. NPF-30
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Union Electric Company (UE, the licensee),
dated November 3, 2022, as supplemented by letters dated June 5, July 25, and August 15, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-30 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan*
The Technical Specifications contained in Appendix A, as revised through Amendment No. 234 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
This amendment is effective as of its date of issuance and shall be implemented within 90 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Jennifer L. Dixon-Herrity, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to Renewed Facility Operating License No. NPF-30 and the Technical Specifications Date of Issuance: September 25, 2023 Jennifer L.
Dixon-Herrity Digitally signed by Jennifer L. Dixon-Herrity Date: 2023.09.25 14:15:23 -04'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 234 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-30 CALLAWAY PLANT, UNIT NO. 1 DOCKET NO. 50-483 Replace the following pages of Renewed Facility Operating License No. NPF-30 and the Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Renewed Facility Operating License REMOVE INSERT Technical Specifications REMOVE INSERT 5.0-19 5.0-19 5.0-20 5.0-20
Renewed License No. NPF-30 Amendment No. 234 (3)
UE, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)
UE, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source of special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)
UE, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This renewed license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level UE is authorized to operate the facility at reactor core power levels not in excess of 3565 megawatts thermal (100% power) in accordance with the conditions specified herein.
(2)
Technical Specifications and Environmental Protection Plan*
The Technical Specifications contained in Appendix A, as revised through Amendment No. 234 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
Environmental Qualification (Section 3.11, SSER #3)**
Deleted per Amendment No. 169.
Amendments 133, 134, & 135 were effective as of April 30, 2000 however these amendments were implemented on April 1, 2000.
The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.
Programs and Manuals 5.5 5.5 Programs and Manuals CALLAWAY PLANT 5.0-19 Amendment No.
5.5.15 Safety Function Determination Program (SFDP) (continued) c.
Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and d.
Other appropriate limitations and remedial or compensatory actions.
A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:
a.
A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or b.
A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or c.
A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.
The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.
5.5.16 Containment Leakage Rate Testing Program (continued) a.
A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Nuclear Energy Institute (NEI) Topical Report (TR) NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J,"
Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008, as modified by the following exceptions:
234
Programs and Manuals 5.5 5.5 Programs and Manuals CALLAWAY PLANT 5.0-20 5.5.16 Containment Leakage Rate Testing Program (continued) b.
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 48.1 psig.
c.
The maximum allowable containment leakage rate, La, at Pa, shall be 0.20% of the containment air weight per day.
d.
Leakage rate acceptance criteria are:
1.
Containment leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and C tests and 0.75 La for Type A tests; 2.
Air lock testing acceptance criteria are:
a)
Overall air lock leakage rate is 0.05 La when tested at Pa; b)
For each door, leakage rate is 0.005 La when pressurized to 10 psig.
e.
The provisions of Technical Specification SR 3.0.2 do not apply to the test frequencies in the Containment Leakage Rate Testing Program.
f.
The provisions of Technical Specification SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.
(continued) 1.
The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.
2.
The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.
Amendment No. 234
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 234 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-30 UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT NO. 1 DOCKET NO. 50-483
1.0 INTRODUCTION
By application dated November 3, 2022 (Reference 1), as supplemented by letters dated June 5, July 25, and August 15, 2023 (References 2, 3, and 4, respectively), Union Electric Company, doing business as (dba) Ameren Missouri (the licensee), requested changes to the technical specifications (TSs) for the Callaway Plant, Unit No. 1 (Callaway). The license amendment request (LAR) proposes changes to Callaway TS 5.5.16, Containment Leakage Rate Testing Program, to allow for the permanent extension of the Type A integrated leakage rate test (ILRT) and Type C leak rate test frequencies based on the guidance in Nuclear Energy Institute (NEI) topical report (TR) NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR [Title 10 of the Code of Federal Regulations] Part 50, Appendix J
[Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors],
Revision 3-A, dated July 2012 (Reference 5).
Specifically, the proposed change would revise Callaway TS 5.5.16 by replacing the existing reference to Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program, dated September 1995 (Reference 6) with a reference to NEI 94-01, Revision 3-A, and the limitations and conditions specified in NEI 94-01, Revision 2-A, dated October 2008 (Reference 7), as the documents used to implement the performance-based containment leakage testing program in accordance with Option B, Performance-Based Requirements, of 10 CFR Part 50, Appendix J.
The supplemental letters dated June 5, July 25, and August 15, 2023, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC, the Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register (FR) on January 24, 2023 (88 FR 4218).
2.0 REGULATORY EVALUATION
2.1 Licensees Proposed Changes The LARs proposed change to Callaway TS chapter 5.0, Administrative Controls, section 5.5.16 will replace RG 1.163 with NEI 94-01, Revision 3-A and the conditions and limitations specified in NEI 94-01, Revision 2-A.
Additionally, the licensee proposed the deletion of TS 5.5.16.a, exceptions 3 and 4. Exception 3 addresses post-modification ILRT associated with the steam generator (SG) replacement that occurred in 2005. Exception 4 addresses the performance of the Callaway Type A test that occurred in 2014. The exceptions are for activities that already took place and are no longer applicable.
2.2 Regulatory Requirements The NRC regulatory requirements related to the content of the TSs are contained in 10 CFR 50.36, Technical specifications. The regulations, in 10 CFR 50.36(c)(5),
Administrative controls, state, in part:
Administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.
The regulations in 10 CFR 50.54(o) require that the primary reactor containments for water-cooled power reactors be subject to the requirements set forth in 10 CFR Part 50, Appendix J.
Appendix J to 10 CFR Part 50 specifies containment leakage testing requirements, including the types required to ensure the leak-tight integrity of the primary reactor containment and systems and components which penetrate the containment. In addition, Appendix J to 10 CFR Part 50 discusses leakage rate acceptance criteria, test methodology, frequency of testing, and reporting requirements for each type of test. The testing requirements in Appendix J to 10 CFR Part 50 ensure that (a) leakage through containments or systems and components penetrating containments does not exceed allowable leakage rates specified in the TSs, and (b) integrity of the containment structure is maintained during its service life.
The regulations in 10 CFR 50.55a, Codes and standards, contain the containment inservice inspection (ISI) requirements, which, in conjunction with the requirements of 10 CFR Part 50, Appendix J, ensure the continued leak-tight and structural integrity of the containment during its service life.
Section V.B.3 of 10 CFR Part 50, Appendix J, Option B, states that the RG or other implementation document used by a licensee to develop a performance-based leakage-testing program must be included, by general reference, in the plant TSs. The submittal for TS revisions must contain justification, including supporting analyses, if the licensee deviates from methods approved by the NRC and endorsed in an RG.
Option B of Appendix J to 10 CFR Part 50, specifies performance-based requirements and criteria for preoperational and subsequent leakage rate testing. These requirements are met by performing (1) Type A tests to measure the containment system overall integrated leakage rate, (2) Type B pneumatic tests to detect and measure local leakage rates across pressure retaining
leakage-limiting boundaries such as penetrations, and (3) Type C pneumatic tests to measure containment isolation valve leakage rates.
After the containment system has been completed and is ready for operation, Type A tests are conducted at periodic intervals based on the historical performance of the overall containment system to measure the overall integrated leakage rate. The leakage rate test results must not exceed the maximum allowable leakage (La) at design-basis loss-of-coolant accident (LOCA) pressure (Pa) with margin, as specified in the TSs. Option B also requires that a general visual inspection for structural deterioration of the accessible interior and exterior surfaces of the containment system be conducted prior to each Type A test and at a periodic interval between tests based on the performance of the containment system. A general visual inspection is necessary as structural deterioration of the surfaces of the containment system may affect the containments leak-tight integrity.
Type B and Type C tests are performed based on the safety significance and historical performance of each boundary and isolation valve to ensure integrity of the overall containment system as a barrier to fission product release.
The adoption of the Option B performance-based containment leakage rate testing for Types A, B, and C testing did not alter the basic method by which Appendix J to 10 CFR Part 50 leakage rate testing is performed; however, it did alter the frequency at which Types A, B, and C containment leakage tests must be performed. Under the performance-based option of 10 CFR Part 50, Appendix J, the test frequency is based upon an evaluation that reviewed as-found leakage history to determine the frequency for leakage testing which provides assurance that leakage limits will be maintained. The change to the Type A test frequency did not directly result in an increase in containment leakage. Similarly, the proposed change to the Type C test frequencies would not directly result in an increase in containment leakage.
2.3 Regulatory Guidance NEI 94-01, Revision 0, dated July 1995 (Reference 8), provides methods for complying with the provisions of 10 CFR Part 50, Appendix J, Option B, and includes provisions that address the extension of the performance-based Type A test interval for up to 10 years, based upon two consecutive successful tests. It is part of the current Callaway licensing basis.
NEI 94-01, Revision 2-A, incorporates the regulatory positions stated in RG 1.163 and delineates a performance-based approach for determining Types A, B, and C containment leakage rate testing frequencies. It also includes provisions for extending Type A ILRT intervals to up to 15 years. This approach uses industry performance, plant-specific data, and risk insights in determining the appropriate testing frequency, and discusses the performance factors that licensees must consider in determining test intervals. In a letter dated June 25, 2008 (Reference 9), the NRC published a safety evaluation (SE) with limitations and conditions for NEI 94-01, Revision 2 and EPRI Report No. 1009325 (Reference 10).
NEI 94-01, Revision 3-A, provides guidance for extending Type C local leakage rate test (LLRT) intervals beyond 60 months. The NRC published an SE with limitations and conditions for NEI 94-01, Revision 3, by letter dated June 8, 2012 (Reference 11). In the SE, the NRC concluded that NEI 94-01, Revision 3, describes an acceptable approach for implementing the optional performance-based requirements of 10 CFR Part 50, Appendix J, and is acceptable for referencing by licensees proposing to amend their containment leakage rate testing TSs,
subject to two conditions. The SE was incorporated into Revision 3 and subsequently issued as NEI 94-01, Revision 3-A, in July 2012.
EPRI Report No. 1009325, Revision 2-A1, provides a generic assessment of the risks associated with a permanent extension of the ILRT surveillance interval to 15 years, and a risk-informed methodology to be used to confirm the risk impact of the ILRT extension on a plant-specific basis. Probabilistic risk assessment (PRA) methods are used, in combination with ILRT performance data and other considerations, to justify the extension of the ILRT surveillance interval. This is consistent with guidance provided in RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (Reference 12), and RG 1.177, Revision 1, An Approach for Plant-Specific, Risk-Informed Decision-making: Technical Specifications (Reference 13), to support changes to surveillance test intervals.
3.0 TECHNICAL EVALUATION
3.1 ILRT History (Type A Testing)
At Callaway, a Type A ILRT is currently required to be performed once every 10 years. The LAR proposes extending the maximum Type A test interval to 15 years.
Per Callaway TS 5.5.16.c, the licensee specified a maximum allowable containment leakage rate, La, of 0.20 percent of the containment air weight per day at the calculated peak pressure, Pa. TS 5.5.16.b states that the peak calculated containment internal pressure for the design-basis LOCA, Pa, is 48.1 pounds per square inch gauge (psig). Footnote 1 to LAR table 3.3.4-1, Periodic Type A ILRT Results for Callaway Unit 1, states, in part, that The containment design pressure is 60 psig.
There have been five ILRTs performed on the Callaway containment since January 1984 (Callaway startup). In the LAR, the licensee provided the results of these ILRTs, which show that substantial margin has been maintained relative to the performance criterion for the most recent Type A tests, such that the extended interval would be allowed by program guidance for Callaway. The results of these ILRTs are documented in LAR section 3.3.4, Integrated Leakage Rate Testing History (ILRT). The results are summarized in table 3.1.1 below.
1 EPRI Report 1018243 is also identified as EPRI Report 1009325, Revision 2-A. This report is publicly available and can be found at www.epri.com by typing 1018243 in the search box.
TABLE 3.1.1 Callaway Type A ILRT History Test Date Pa (psig)
(1)
Pt Test Pressure (psig)
(2)
Pd Containment Design Pressure (psig) 95% Upper Confidence Limit (UCL)
(wt%/day)
As-Found Acceptance Criteria (wt%/day)
As-Left Acceptance Criteria (wt%/day)
Jan 1984*
48.1 50.05 60.0 0.047 0.20 (1.0 La) 0.15 (0.75 La)
April 1987 48.1 49.7 60.0 0.044 0.20 (1.0 La) 0.15 (0.75 La)
Feb 1990 48.1 50.3 60.0 0.066 0.20 (1.0 La) 0.15 (0.75 La)
Oct 1999 48.1 47.7 60.0 0.0445 0.20 (1.0 La) 0.15 (0.75 La)
Oct 2014 48.1 48.6 60.0 0.1038 0.20 (1.0 La) 0.15 (0.75 La)
Table 3.1.1 Notes:
(1) Pa - As defined in Callaway TS 5.5.16.b (2) Pt - Final test pressure (psig) - minimum allowable Pt is 0.96Pa psig = 46.2 psig (Reference ANSI/ANS 56.8-1994, American National Standard for Containment System Leakage Testing Requirements, section 3.2.11) (Reference 14)
(*) Preoperational ILRT performed in conjunction with a structural integrity test.
Section 9.1.2, Test Interval, of NEI 94-01, Revision 3-A, states, in part, that The elapsed time between the first and the last tests in a series of consecutive passing tests used to determine performance shall be at least 24 months. As shown in table 3.1.1 of this SE, this requirement of NEI 94-01, Revision 3-A, section 9.1.2 has been satisfied.
Callaway TS 5.5.16.a currently references RG 1.163. Section C, Regulatory Position, of RG 1.163 states, in part, that NEI 94-01, Revision 0, provides methods acceptable to the NRC staff for complying with the provisions of Option B in Appendix J to 10 CFR Part 50.
Section 9.2.3, Extended Test Intervals, of NEI 94-01, Revision 0, states, in part, that:
In reviewing past performance history, Type A test results may have been calculated and reported using computational techniques other than the Mass Point method from ANSI/ANS 56.8-1994 (e.g., Total Time or Point-to-Point).
Reported test results from these previously acceptable Type A tests can be used to establish the performance history. Additionally, a licensee may recalculate past Type A UCL (using the same test intervals as reported) in accordance with ANSI/ANS 56.8-1994 Mass Point methodology and its adjoining Termination criteria in order to determine acceptable performance history.
NEI 94-01, Revision 3-A, includes substantively identical language except for referring to ANSI/ANS 56.8-2002, Containment System Leakage Testing Requirements (Reference 15),
instead of to ANSI/ANS 56.8-1994.
The NRC staff notes that this language does not mandate that a licensee recalculate past Type A test results to demonstrate conformance with the definition of performance leakage rate contained in NEI 94-01, Revision 3-A. The NRC staff also notes that the Callaway ILRT
results since January 1984 (Callaway startup) demonstrate ample margin (i.e., greater than (>)
48 percent) between each UCL leakage value and La.
Callaway TS 5.5.16.d.1 establishes the maximum limit for the as-left leakage rate for startup following completion of Type A testing at less than or equal to () 0.75 La, which currently equals 0.15 percent of containment air weight per day.
Callaway TS 5.5.16.c specifies a leakage rate La not to exceed 0.20 percent of containment air weight per day at the calculated peak pressure, Pa. As displayed in table 3.1.1 of this SE, there has been adequate margin to the performance limit as described in TS 5.5.16.c of La for the historical ILRTs since startup for Callaway.
The past ILRT results for Callaway since startup have confirmed that the primary containment leakage rates are acceptable with respect to the design criterion leakage of containment air weight (La) per day. Since the last two Type A tests for Callaway had as-found test results well within the current maximum allowable containment leakage rate specified in TS 5.5.16.c of 0.20 weight-percent/day (1.0 La), a test frequency of 15 years in accordance with NEI 94-01, Revision 3-A, and the conditions and limitations of NEI 94-01, Revision 2-A, would be acceptable.
The NRC staff concludes that the Callaway ILRT test results provide reasonable assurance that containment overall leakage will be maintained below the design-basis leak rate, consistent with the requirements in TS 5.5.16, and will fulfill the requirements of 10 CFR 50, Appendix J, Option B, with the test frequency of 15 years.
3.2 Type B and C Testing The LAR proposes extending the containment isolation valve (CIV) leakage rate testing (Type C) frequency from the 60 months currently permitted by 10 CFR Part 50, Appendix J, Option B, to a 75-month frequency for Type C leakage rate testing of selected components, in accordance with NEI 94-01, Revision 3-A.
The Callaway 10 CFR Part 50, Appendix J, Types B and C leakage rate test program requires testing of electrical penetrations, airlocks, hatches, bellows, and valves within the scope of the program as required by 10 CFR Part 50, Appendix J, Option B and TS 5.5.16. There are 65 Type B penetrations for Callaway, which include two airlocks (i.e., one personnel hatch and one auxiliary hatch); an equipment hatch; the fuel transfer tube; 56 electrical penetrations; penetration 34 (i.e., containment pressurization line); penetration 51 (i.e., ILRT pressurization pressure sensing line); and penetrations 36, 50, and 68 (i.e., maintenance spare air and electrical access penetrations). There are 109 Type C test components for Callaway, which include the 72 General Design Criteria (GDC) 55 and 56 valve penetrations listed in Callaway Final Safety Analysis Report (FSAR) table 6.2.4-1 (Reference 16).
The NRC staff reviewed the local leak rate summaries contained in LAR section 3.5.5, Containment Leakage Rate Testing Program - Type B and Type C Testing Program. For Callaway, LAR section 3.5.5 indicates that the combined Types B and C leakage acceptance criterion of 0.60 La equals 252,028 standard cubic centimeters per minute (sccm) and La equals 420,046.7 sccm. NEI 94-01, Revision 3-A, section 10.2, Type B and Type C Testing Frequencies, indicates that this criterion is to be compared to the combined Types B and C as-found minimum pathway test total as a restart permissive criterion to the combined Types B and C as-left maximum pathway test total.
Request for Additional Information (RAI)-SCPB-3 During its review, the NRC staff noted that LAR section 3.6.7, Local Leak Rate Testing Program Effectiveness, Example #3 details a long-lived and ongoing problem with essential service water primary containment penetrations 28, 29, 71, and 73. The NRC staff reviewed LAR table 3.5.6-1, Callaway Types B and C LLRT Combined As-Left Trend Summary, and noted that the problem existed as far back as refueling (RF) outage 17 (i.e., 2010). LAR section 3.6.6, Containment Isolation System (SM) Exceeded the Maintenance Rule (MR)
Performance Criteria - RF20 (2014), reflects that during RF20 (i.e., 2014), the containment isolation system exceeded the maintenance rule performance criteria. Subsequently, during RF21 (i.e., 2016), the as-left maximum pathway leakage rate (MXPLR) was dramatically reduced from 170,116.74 sccm to 53,139.63 sccm for reasons not explained in the LAR.
However, LAR section 3.6.7 concludes that Table 3.5.6-1, Callaway Types B and C LLRT Combined As-Left Trend Summary, shows the effectiveness of the corrective actions taken to address LLRT Program Effectiveness. Based on these observations, in RAI-SCPB-3 dated April 5, 2023 (Reference 17), the NRC staff requested that the licensee provide additional information to adequately support this conclusion.
In its response to RAI-SCPB-3, dated June 5, 2023 (Reference 2), the license stated, in part:
Penetrations 28, 29, 71, and 73 have had issues with corrosion and performance of the valves associated with these penetrations. Those issues have impacted ILRT, LLRT testing and required repairs. In RF6 (fall of 1993), Callaway implemented a 1991 approved design change, i.e., modification package MP 91-1004, which installed double offset stainless steel valves. This eliminated the corrosion issues with the containment isolation valves and helped with the seat leakage. However, the modification did not completely resolve the seat leakage issues since the issue had more to do with the water and residual carbon steel piping than the valves themselves.
During RF21, the licensee installed 316 stainless steel (SS) replacement valves for EJV0343, 344, 345, and 346. This enhanced the ability to prioritize the valves to be worked during RF outages for containment penetrations 28, 29, 71, and 73. The plant enhancement led to significant improvement in the as-left MXPLR performance summation for penetrations 28, 29, 71, and 73. This improvement is reflected in table RAI-SCPB-3, Callaway Type C LLRT As-Left MXPLR Trend Summary for Penetrations 28, 29, 71, and 73. A comparison of the as-left MXPLR for penetrations 28, 29, 71, and 73 to the as-left Types B and C MXPLR for each penetration following the completion of the RF21 repair activities shows the effectiveness of the corrective actions taken to address LLRT program effectiveness.
The licensee also stated in its response to RAI-SCPB-3 that [s]ince installation, stainless steel replacement valves have generally performed satisfactorily, but the raw water continues to be a challenge with respect to degradation of the seats. The challenges to valve sealing integrity consist of the internal environmental factors of raw water, entrained silt, debris, and corrosion.
As a result of the described operating history, the subject containment isolation valves for penetrations 28, 29, 71, and 73 have been placed on a test frequency such that they are tested every RF outage. If an adverse step change is identified or the administrative limit is exceeded, a condition report and a repair job to perform the required maintenance is initiated. The
performance bases eligibility report states that these penetrations cannot be placed on extended test intervals.
Since the overall leak test performance for the containment isolation valves associated with penetrations 28, 29, 71, and 73 has displayed dramatic and sustained improvement since the completion of RF21, the NRC staff concludes that the licensee has a corrective action program that appropriately addresses poor performing valves and penetrations.
RAI-SCPB-2 During its review, the NRC staff noted that LAR table 3.5.6-1 did not include the results for the as-found minimum pathway leakage rates (MNPLRs) for the subject RF outages nor results for the percentage of 0.6 La (percent/24 hours). The staff noted that as-found MNPLRs from the current RF outage (i.e., when combined with as-left MXPLRs from the previous two or three RF outages, corrected for understatement where appropriate) were the true measure of the effectiveness of the licensees Types B and C penetration maintenance program. The staff requested that the licensee provide the as-found MNPLRs for the subject RF outages and the corresponding results for the percentage of 0.6 La. In response to RAI-SCPB-2, the licensee revised LAR section 3.5.6, Type B and Type C Local Leak Rate Testing Program Implementation Review, and table 3.5.6-1 to incorporate the MNPLR results for each respective RF outage.
As documented in the licensees response to RAI-SCPB-2, the as-found MNPLR, as-left MNPLR, and as-left MXPLR test values for Callaway following the completion of the RF21 repair activities are summarized as follows:
As-found MNPLR leak rate shows an average of 13.1 percent of 1.0 La with a high of 27.9 percent of 1.0 La.
As-left MNPLR leak rate shows an average of 9.94 percent of 1.0 La with a high of 13.0 percent of 1.0 La.
As-left MXPLR shows an average of 15.4 percent of 1.0 La with a high of 18.6 percent of 1.0 La.
The NRC staff notes that since the completion of the RF21 repair activities (e.g., containment penetrations 28, 29, 71, and 73) the Types B and C test results show a large amount of margin between the actual as-found and the as-left outage summations and the respective TS leakage rate acceptance criteria.
Of the 65 Type B penetrations, as documented in the response to RAI-SCPB-1, 90 percent of these penetrations are currently on extended test frequencies. Of the 109 Type C components, 59 percent of these penetrations are currently on extended test frequencies. The number of penetrations on extended frequency is adjusted periodically based on valve performance and other plant testing requirements as previously discussed.
Based on the NRC staffs review of the historical information provided in (1) LAR section 3.5.5, (2) LAR section 3.5.6, and (3) the licensees responses to RAI-SCPB-1, RAI-SCPB-2, and RAI-SCPB-3, the NRC staff observed that the licensee is successfully implementing the requirements of its 10 CFR Part 50, Appendix J, Option B performance-based testing program.
In summary, the licensee provided an adequate explanation of the cause of failure for the LLRT Type C penetration failures experienced during the most recent RF outages for Callaway.
Furthermore, based on the review of LAR sections 3.5.5 and 3.5.6, as supplemented by the RAI responses, the NRC staff concluded that the aggregate leakage rate results of the as-found minimum pathway for all Callaway Types B and C tests from the last five RF outages since the completion of RF21 (i.e., 2016) demonstrate a history of adequate maintenance. These aggregate test results at the end of each operating cycle were well below (i.e., > 21.7 percent margin) the Types B and C test TS leakage rate acceptance criteria of less than (<) 0.60 La.
Based on its review of the information in LAR sections 3.5.5 and 3.5.6 and the RAI responses, the NRC staff concludes that there is reasonable assurance that the licensee has been compliant with the guidance of both section 10.2.1, Type B Test Intervals, and section 10.2.3, Type C Test Interval, of NEI 94-01, Revision 0.
Types B and C Test Program Assessment - Callaway In summary, the NRC staff determined that:
The licensee has been compliant with the guidance of RG 1.163 and NEI 94-01, Revision 0; The recent historical combined total Types B and C test results are substantially below the acceptance limit of TS 5.5.16; and The licensee has a corrective action program that appropriately addresses poor performing valves and penetrations.
The NRC staff finds that the licensee is effectively implementing the Callaway Type B and C leakage rate test program, as required by Option B of 10 CFR Part 50, Appendix J. Therefore, extending the CIV leakage rate testing (Type C) frequency from the 60 months currently permitted by 10 CFR [Part] 50, Appendix J, Option B, to a 75-month frequency for Type C leakage rate testing of selected components, in accordance with NEI 94-01, Revision 3-A, is acceptable.
3.3 Containment Inspection 3.3.1 Containment Inservice Inspection Programs The ISI Programs are summarized below for subsections IWE and IWL to the requirements of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (ASME Code),Section XI, Inservice Inspection of Nuclear Power Plant Components, the 2007 Edition with the 2008 Addenda and with the applicable conditions in 10 CFR 50.55a(b)(2)(ix), and for Service Level I (SLI) Containment Coating and Assessment Program to the requirements of American Society for Testing and Materials (ASTM) D5163-08, Standard Guide for Establishing a Program for Condition Assessment of Coating Service Level I Coating Systems in Nuclear Power Plants (Reference 18).
Service Level I Containment Coating and Assessment Program In LAR section 3.5.1, Containment Building Coatings, the licensee described containment building coatings procedure responsibilities and actions of engineering, work control, quality
control, and painters for the surface preparation of substrates to be painted, qualification of painters and inspection personnel, and the application and inspection of coatings and surfacers to equipment, systems, and structures for use in containment. The Callaway FSAR programs/activities also credit the implementation of the protective coatings monitoring and maintenance aging management program for Service Level I (SLI) coatings as described in LAR section 3.7, License Renewal Aging Management.
At Callaway, the licensee inspects coated surfaces at each refueling outage in accordance with ASTM D5163-08. The inspections include a general visual inspection on all readily accessible coated surfaces during a walkthrough. Thorough visual inspections are performed on previously designated areas and on areas noted as deficient during the walkthrough.
Based on the above, the NRC staff finds that the program is consistent with the requirements of ASTM D5163-08 and that the licensee has demonstrated that it is meeting the program by implementing adequate general visual inspections of the accessible coated surfaces and documenting deficiencies found in the Containment Coating Deficiencies Log in the Containment Building Coating Condition Assessment.
Summary of ASME Section XI, Subsections IWE and IWL Containment ISI Programs The containment leak-tight integrity is verified through periodic ISIs conducted in accordance with the requirements of the ASME Code,Section XI, subsection IWE that provides the rules and requirements for the ISI of Class metal containment (MC) pressure-retaining components and their integral attachments. NRC regulations in 10 CFR 50.55a(b)(2)(ix)(E) [2018] require licensees to conduct a general visual examination once each period as required by subsection IWE. ASME Section XI requires visual examinations 3 times within a 10-year interval for ASME Class MC components and their integral attachments (subsection IWE) and 2 times within a 10-year interval for ASME Class concrete containment (CC) components and their integral attachments (subsection IWL). Furthermore, NEI 94-01 requires that visual examinations be conducted during at least three other outages, and in the outage during which the ILRT is being conducted, and these requirements are not changed because of the extended ILRT interval. In addition, Appendix J to 10 CFR Part 50, Type B local leak tests performed to verify the leak-tight integrity of containment penetration bellows, airlocks, seals, and gaskets are also not affected by the change to the Type A test frequency.
In LAR section 3.5.2, Containment Pressure Boundary Inservice Inspection Program, the licensee outlined the requirements for the non-destructive examinations (NDEs) of Callaways containment pressure boundary and related components as specified by 10 CFR 50.55a. The licensee developed the IWE program in accordance with the requirements of ASME Section XI, subsection IWE, 2007 Edition with the 2008 Addenda and the applicable conditions in 10 CFR 50.55a(b)(2)(ix),Section XI condition: Metal containment examinations. The licensee provided the dates of inspection periods in the containment pressure boundary inspection intervals, as follows:
The three inspection periods during the third containment inspection interval:
First Period (36 months):
December 1, 2017, to November 30, 2020 Second Period (36 months): December 1, 2020, to November 30, 2023 Third Period (36 months):
December 1, 2023, to November 30, 2026
The three proposed inspection periods during the fourth containment inspection interval:
(The fourth interval plan has not been finalized yet):
First Period (36 months):
December 1, 2026, to November 30, 2030 Second Period (36 months): December 1, 2030, to November 30, 2033 Third Period (36 months):
December 1, 2033, to November 30, 2036 The licensee described that ASME Section XI, IWA-2430 allows the inspection interval to be increased or decreased by 12 months to coincide with Callaway RF outages. In addition, the licensee is implementing Code Case N-765, Alternative to Inspection Interval Scheduling Requirements of IWA-2430,Section XI, Division 1, which modifies IWA-2430(c), providing additional scheduling flexibility.
In its supplemental letter dated June 5, 2023, the licensee referenced Callaway procedure ESP-ZZ-01016, ASME Section XI IWE Containment Pressure Boundary Inspection, which provides the examination acceptance criteria for the IWE program.
In LAR section 3.5.3, Containment Exterior Concrete and Tendon Inspection Program, the licensee outlined the requirements for the non-destructive examination (NDE) and testing of Callaways concrete containment and post-tensioning system as specified by 10 CFR 50.55a.
The licensee developed the IWL program in accordance with the requirements of ASME Section XI, subsection IWL, 2007 Edition with the 2008 Addenda and the applicable conditions in 10 CFR 50.55a(b)(2)(viii),Section XI condition: Concrete containment examinations. The licensee provided the dates of inspection periods for the containment concrete and tendon inspection intervals, as follows:
The two inspection periods during the third containment inspection interval:
First Period (5 years):
September 9, 2016, to September 8, 2021 (35th Year Inspection)
Second Period (5 years):
September 9, 2021, to September 8, 2026 (40th Year Inspection).
The two inspection periods during the fourth containment inspection interval:
First Period (5 years):
September 9, 2026, to September 8, 2031 (45th Year Inspection)
Second Period (5 years):
September 9, 2031, to September 8, 2036 (50th Year Inspection)
The licensee also described that ASME Section XI, IWL-2410 and 2420 allow the extension of the concrete and post-tensioning system examination by 12 months, if necessary.
As stated in its supplemental letter dated June 5, 2023, Callaway procedure ESP-ZZ-01012, Containment Post-Tensioning System Inspection, which references to Specification C-1003, Specification of Inservice Inspection of the Containment Building Post-Tensioning System and Exterior Concrete Shell, provides the acceptability criteria of concrete surface condition in accordance with sections 5.2 and 5.3 of American Concrete Institute (ACI) 349.3R, Report on Evaluation and Repair of Existing Nuclear Safety-Related Concrete Structures.
Summary of the Containment Inservice Inspection Examination Results:
In LAR table 3.5.2-1, Subsection IWE Summary Table, column, Number of Components, the licensee listed the population of metallic components potentially subject to examinations. The number of components examined during the inspection interval will be based upon the ASME Code requirements for the subject item number. The examination acceptance criteria for IWE and examinations are specified in Callaway procedure ESP-ZZ-01016.
The license stated that during the RF24 outage, Zones 1 through 28 and 31 through 34 (32 total) were visually inspected with no defects noted. During the IWE general inspection of the incore pit, several areas of coating distress were noted. However, after cleaning and scraping it was revealed that the liner was free of defects and, therefore, was acceptable.
During the RF25 outage, seven containment penetrations were visually inspected (VT-1) under the IWE activities. All seven inspections were noted to be satisfactorily completed, and no degradation was identified.
During the 35th year, IWL L-B L2.20 testing of post-tension tendon V39 resulted in the tendon tested tensile strength of 235.3 thousand pounds per square inch (ksi), which is less than the ultimate tensile strength of 240 ksi. The licensee reviewed the original containment design and stated that the tendon system was designed using an average wire stress of 156.9 ksi.
Furthermore, all six samples tested reached a minimum yield stress of 213.2 ksi. This is greater than the maximum design yield stress limit of 192 ksi. The licensee noted that the lift-off forces for all surveyed tendons were within or above the projected acceptance band at the 35th year surveillance point and determined that each of the tendons is operable by meeting its overall design function.
The licensee provided LAR tables 3.5.1-1, Unqualified Coatings Log - Post RF25 Update 06/20/2022, and 3.5.1-2, Unqualified Coatings Log - Post RF24 Update 01/26/2021, for RF25 and RF24, respectively, that quantified the change in unqualified coating areas or areas confirmed to be design-basis accident qualified SLI coatings. During the RF24 inspections, the licensee discovered several areas of coating distress in the incore pit. The areas were cleaned and scraped to expose the liner for inspection and the liner was free of defects and accepted.
Based on the above, the NRC staff finds that the licensee has an adequate containment ISI program in place as demonstrated by the implementation of overlapping inspection activities performed as part of the IWE/IWL programs and activities developed to support renewal of the original operating license, and inspections of SLI protective coatings. These programs periodically examine, monitor, and manage structural deterioration and aging degradation of the Callaway containment pressure boundary such that the primary containment can perform its intended function as a leak-tight barrier consistent with the guidance contained in NEI 94-01.
3.3.2 Operating Experience In LAR section 3.6, Operating Experience (OE), the licensee described the impacts of the following selected site-specific and industry operating experiences.
In LAR section 3.6.1, IN 1992-20, Inadequate Local Leak Rate Testing, the licensee reviewed for applicability to Callaway NRC Information Notice (IN) 1992-20, Inadequate Local Leak Rate Testing, dated March 3, 1992 (Reference 19). The NRC issued IN 1992-20 to alert licensees to problems with the local leak rate testing of two-ply stainless-steel bellows used on piping
penetrations at some plants. The licensee stated that there are no pressure retaining bellows used on containment penetrations at Callaway.
In LAR section 3.6.2, IN 2004-09, Corrosion of Steel Containment and Containment Liner, the licensee reviewed for applicability to Callaway containment steel liner-plate NRC IN 2004-09, Corrosion of Steel Containment and Containment Liner, dated April 27, 2004 (Reference 20).
The NRC issued IN 2004-09 to alert licensees on how corrosion of the liner plate could reduce safety factors and could change the failure threshold of the containment during extreme conditions. The licensee stated that at Callaway ASME Section XI, subsection IWE visual inspections of 100 percent of all accessible areas are performed during every other RF outage.
During RF12 (2002), corroded areas of liner plate were cleaned and painted, and new SS plates were installed over existing carbon steel plates at containment normal sumps. Additionally, during RF12, a 1-foot band of liner plate above the 2000 ft elevation floor was painted along the entire perimeter of the reactor building. The licensee stated that the plant does not have a moisture barrier seal that adjoins the concrete floor to the liner plate because a 1-foot-thick concrete slab is typically installed over the liner plates. A walkdown of the fill slab and liner plate interface performed during RF13 (2004) confirmed that the interface is in good condition.
In LAR section 3.6.3, IN 2010-12, Containment Liner Corrosion, the licensee reviewed for applicability to Callaway containment steel liner-plate NRC IN 2010-12, Containment Liner Corrosion, dated June 18, 2010 (Reference 21). The NRC issued IN 2010-12 to inform licensees of issues concerning the degradation (corrosion) of the containment liner that could affect the leak-tightness of the containment structure. The licensee stated that at Callaway ASME Section XI, subsection IWE inspections are conducted every 3 years during the RF outage. The entire containment boundary and liner plate are inspected during each 3-year period. Additionally, Callaway Engineering conducts a containment coatings survey during every RF outage to inspect for loose or damaged coatings.
In LAR section 3.6.4, IN 2014-07, Degradation of Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner, the licensee reviewed for applicability to Callaway leak-chase systems NRC IN 2014-07, Degradation of Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner, dated May 5, 2014 (Reference 22). The NRC issued IN 2014-07 to inform licensees of issues concerning degradation of floor weld leak-chase channel systems that could affect leak-tightness and aging management of the containment structure. The leak-chase system provides a pathway for potential intrusion of moisture that could cause corrosion degradation of inaccessible areas of the pressure-retaining containment shell. The licensee stated that at Callaway the leak-chase system does not have access points at the concrete floor level. The accessible portions of the leak-chase system are inspected under the ASME Section XI, subsection IWE program, and no presence of degradation have been noted that require further evaluation. The licensee also stated that at Callaway the steel liner to concrete floor interface is inspected every 18 months during the RF outage. The latest such inspection during RF20 identified no issues related to IN 2014-07.
In LAR section 3.6.5, RIS 2016-07 Containment Shell or Liner Moisture Barrier Inspection, the licensee reviewed for applicability to Callaway containment moisture barrier inspection NRC Regulatory Issue Summary (RIS) 2016-07, Containment Shell or Liner Moisture Barrier Inspection, dated May 9, 2016 (Reference 23). The licensee stated, in part, that Callaway does not have a moisture barrier seal that adjoins the concrete floor to the liner plate. In its response dated June 5, 2023, to an NRC staff RAI, the licensee included detailed fabrication
drawings showing the interface between the steel liner-plate to the concrete floor in the containment building. The licensee stated, in part:
The steel liner to concrete floor interface is also inspected every 18 months (every refuel) under the Coatings Aging Management Program and every 4.5 years (every 3 refuels) under the Structural Monitoring Aging Management Program.
Currently, Callaway has not identified any Subsection IWE components that require augmented examination.
Given the results of these inspections, the licensees evaluation concluded that no new actions were required to address RIS 2016-07.
Based on the NRC staffs review of the historical information provided in LAR sections 3.5.5 and 3.5.6 and the supplemental information provided in the licensees responses to RAI-SCPB-1, RAI-SCPB-2, and RAI-SCPB-3, the NRC staff concludes that there is reasonable assurance that the Callaway performance-based local leak rate testing program will be effective in satisfying the requirements of 10 CFR Part 50, Appendix J, Option B.
3.3.3 Containment Accident Pressure on Containment Spray System and Emergency Core Cooling System Performance NRC Generic Letter 2004-02 In LAR section 3.2.1, Resolution of NRC Generic Letter 2004-02, the licensee reviewed the resolution to Generic Safety Issue (GSI)-191, Assessment of Debris Accumulation on PWR
[Pressurized Water Reactor] Sump Performance, in Generic Letter (GL) 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized Water Reactors, and stated that the Callaway containment accident pressure (CAP) of 1.7 pounds per square inch (psi) is credited to the available net positive suction head (NPSH) during the phase of a large break LOCA (LBLOCA) when containment temperature is above 212 degrees Fahrenheit (°F) to ensure that no flashing to steam would occur in the debris bed at the top of the sump strainers. The CAP credit (1.7 psi), which is approximately 10 percent of the available containment pressure, was approved by the NRC staff in a letter dated October 21, 2022 (Reference 24), in response to an LAR submitted by letter dated March 31, 2021 (Reference 25). This LAR addressed the resolution to GSI-191 and provided a response to GL 2004-02.
The NRC staff finds that the CAP credit of 1.7 psi used to prevent boiling within the sump strainer debris bed is not changed and is consistent with the current licensing basis.
NPSH Analysis In LAR section 3.2.2, Containment Spray System (CSS) Residual Heat Removal (RHR)
NPSH, the licensee reviewed the available NPSH for the containment spray (CS) and RHR pumps that draw water from the containment sump during the recirculation phase of a LBLOCA.
The licensee stated that sufficient water (394,000 gallons) is available in the refueling water storage tank (RWST) for the emergency core cooling system (ECCS) to rapidly reduce LOCA containment pressure and temperature. The licensee also stated that the water in the RWST will allow sufficient water volume in the containment sump to permit recirculation flow between the
core and the containment and meet the NPSH requirements of the CS and RHR pumps.
Additionally, the RWST water volume is sufficient to allow manual switchover of the CS pumps at the beginning of the LOCA recirculation phase.
Callaway FSAR section 6.2.1.5, Minimum Containment Pressure Analysis for Performance Capability Studies on Emergency Core Cooling System, states that the LOCA containment pressure is conservatively minimized in the NPSH analysis for the pumps that draw water from the sump in the LOCA recirculation phase. The available NPSH calculation conservatively assumes that the vapor pressure of the liquid in the sump is equal to the containment ambient pressure so that the available NPSH is equal to the sump static head minus the head loss in the pump suction piping. The NPSH discussion in Callaway FSAR section 6.3.2.2, Equipment and Component Descriptions, does not credit CAP for CS and RHR pumps. The licensee provided the values of the available and required NPSH for the CS and the RHR pumps, which are consistent with Callaway FSAR table 6.2.2-7, Input and Results of NPSH Analysis. There is at least 10 percent NPSH margin in the available NPSH from the required NPSH.
Based on the above, the NRC finds that the licensee will maintain the current licensing basis NPSH analysis without additional CAP credit.
3.4 NEI 94-01, Revision 2-A, Section 4.0, Limitations and Conditions In the SE of NEI 94-01, Revision 2, issued by the NRC dated June 25, 2008, it was concluded that the methodology is acceptable for referencing by licensees proposing to amend their TSs to permanently extend the Type A surveillance test interval to 15 years, subject to the limitations and conditions noted within the SE. In LAR table 3.8.1-1, NEI 94-01 Revision 2-A Limitations and Conditions, the licensee provided a response to each of these limitations and conditions.
3.4.1 NEI 94-01, Revision 2-A, Condition 1 Condition 1 of NEI 94-01, Revision 2-A, states:
For calculating the Type A leakage rate, the licensee should use the definition in the NEI TR 94-01, Revision 2, in lieu of that in ANSI/ANS-56.8-2002. (Refer to SE Section 3.1.1.1).
The licensees response to Condition 1 states:
Callaway will utilize the definition in NEI 94-01 Revision 3-A, Section 5.0. This definition has remained unchanged from Revision 2-A to Revision 3-A of NEI 94-01.
NRC Staff Assessment of Licensees Response to Condition 1 Section 3.2.9, Type A test performance criterion, of ANSI/ANS-56.8-2002 defines the performance leakage rate and reads, in part:
The performance criterion for a Type A test is met if the performance leakage rate is less than La. The performance leakage rate is equal to the sum of the measured Type A test UCL and the total as-left [minimum pathway leakage rate]
of all Type B or Type C pathways isolated during performance of the Type A test.
NRC SE section 3.1.1.1, Type A Performance Leakage Rate, of NEI 94-01 Revision 2, states, in part:
Section 5.0 of NEI TR 94-01, Revision 2, uses a definition of performance leakage rate for Type A tests that is different from that of ANSI/ANS-56.8-2002. The definition contained in NEI TR 94-01, Revision 2, is more inclusive because it considers excessive leakage in the performance determination. In defining the minimum pathway leakage rate, NEI TR 94-01, Revision 2, includes the leakage rate for all Type B and Type C pathways that were in service, isolated, or not lined up in their test position prior to the performance of the Type A test. Additionally, the NEI TR 94-01, Revision 2, definition of performance leakage rate requires consideration of the leakage pathways that were isolated during performance of the test because of excessive leakage in the performance determination. The NRC staff finds this modification of the definition of performance leakage rate used for Type A tests to be acceptable.
Section 5.0, Definitions, of NEI 94-01, Revision 3-A, states, in part:
The performance leakage rate is calculated as the sum of the Type A upper confidence limit (UCL) and as-left minimum pathway leakage rate (MNPLR) leakage rate for all Type B and Type C pathways that were in service, isolated, or not lined up in their test position (i.e., drained and vented to containment atmosphere) prior to performing the Type A test. In addition, leakage pathways that were isolated during performance of the test because of excessive leakage must be factored into the performance determination. The performance criterion for Type A tests is a performance leak rate of less than 1.0La.
The NRC staff reviewed the definitions of performance leakage rate contained in NEI 94-01, Revision 2 and Revision 3-A and concluded that the definitions contained in both documents are substantively identical.
Therefore, the NRC staff concludes that the licensee will use the definition found in section 5.0 of NEI 94-01, Revision 3-A for calculating the Type A leakage rate in the Callaway Containment Leakage Rate Testing Program and that this adequately addresses Condition 1.
3.4.2 NEI 94-01, Revision 2-A, Condition 2 Condition 2 of NEI 94-01, Revision 2-A, states:
The licensee submits a schedule of containment inspections to be performed prior to and between Type A tests. (Refer to SE Section 3.1.1.3).
The licensees response to Condition 2 states:
Reference Section 3.5.2, Table 3.5.2-1 and Section 3.5.3, Tables 3.5.3-1, 3.5.3-2, and 3.5.3-3, of this submittal.
NRC Staff Assessment of Licensees Response to Condition 2 NRC SE section 3.1.1.3, Adequacy of Pre-Test Inspections (Visual Examinations), of NEI 94-01, Revision 2, states, in part:
NEI TR 94-01, Revision 2, Section 9.2.3.2, states that: To provide continuing supplemental means of identifying potential containment degradation, a general visual examination of accessible interior and exterior surfaces of the containment for structural deterioration that may affect the containment leak-tight integrity must be conducted prior to each Type A test and during at least three other outages before the next Type A test if the interval for the Type A test has been extended to 15 years. NEI TR 94-01, Revision 2, recommends that these inspections be performed in conjunction or coordinated with the examinations required by ASME Code,Section XI, Subsections IWE and IWL. The NRC staff finds that these visual examination provisions, which are consistent with the provisions of regulatory position C.3 of RG 1.163, are acceptable considering the longer 15-year interval. Regulatory Position C.3 of RG 1.163 recommends that such examination be performed at least two more times in the period of 10 years.
The NRC staff agrees that as the Type A test interval is changed to 15 years, the schedule of visual inspections should also be revised. Section 9.2.3.2 in NEI TR 94-01, Revision 2, addresses the supplemental inspection requirements that are acceptable to the NRC staff.
Section 9.2.1, Pretest Inspection and Test Methodology, of NEI 94-01, Revision 3-A, states, in part:
Prior to initiating a Type-A test, a visual examination shall be conducted of accessible interior and exterior surfaces of the containment system for structural problems that may affect either the containment structure leakage integrity or the performance of the Type A test. This inspection should be a general visual inspection of accessible interior and exterior surfaces of the primary containment and components. It is recommended that these inspections be performed in conjunction or coordinated with the ASME Boiler and Pressure Vessel Code,Section XI, Subsection IWE/IWL required examinations.
Section 9.2.3.2, Supplemental Inspection Requirements, of NEI 94-01, Revision 3-A, states:
To provide continuing supplemental means of identifying potential containment degradation, a general visual examination of accessible interior and exterior surfaces of the containment for structural deterioration that may affect the containment leak-tight integrity must be conducted prior to each Type A test and during at least three other outages before the next Type A test if the interval for the Type A test has been extended to 15 years. It is recommended that these inspections be performed in conjunction or coordinated with the ASME Boiler and Pressure Vessel Code,Section XI, Subsection IWE/IWL required examinations.
The NRC staff reviewed LAR section 3.5.2 and table 3.5.2-1. Table 3.5.2-1 is a descriptive table of all components that require ASME Code Section XI, subsection IWE and pre-ILRT inspections. In addition, the NRC staff reviewed LAR section 3.5.3 and tables 3.5.3-1, Subsection IWL Inservice Inspection Summary Table, 3.5.3-2, List of Adopted Code Cases, and 3.5.3-3, List of Technical Approach and Positions. Review of these LAR sections and tables supports the conclusion that the licensee performed and continues to perform general visual inspection of the accessible interior and exterior surfaces of the primary containment and components prior to Type A tests. Based on this review, the NRC staff determined that the scheduled IWE and the IWL inspection requirements and the scheduled pre-ILRT primary
containment inspection requirement of NRC staff SE section 3.1.1.3 for NEI 94-01, Revision 2-A, are satisfied for Callaway.
Based on the above, the NRC staff determined that the licensee intends to comply with the guidance contained in NEI 94-01, Revision 3-A, sections 9.2.1 and 9.2.3.2 and to satisfy the provisions contained in NRC staff SE section 3.1.1.3. Therefore, the licensee has adequately addressed Condition 2.
3.4.3 NEI 94-01, Revision 2-A, Condition 3 Condition 3 of NEI 94-01, Revision 2-A, states:
The licensee addresses the areas of the containment structure potentially subjected to degradation. (Refer to SE Section 3.1.3).
The licensees response to Condition 3 states:
Reference Sections 3.5.2 and 3.5.3 of this submittal.
NRC Staff Assessment of Licensees Response to Condition 3 As described in its SE of NEI 94-01, Revision 2-A, section 3.1.3, Type A Test (ILRT), Type B and Type C Tests (LLRTs), and Containment In-Service Inspections (ISIs), the NRC staff identified areas that need to be specifically addressed during the ASME Code Section XI, subsection IWE and IWL inspections. These areas include containment pressure-retaining boundary components (e.g., seals and gaskets of mechanical and electrical penetrations, bolting, penetration bellows) and accessible and inaccessible areas of the containment structures (e.g., moisture barriers, steel shells, liners backed by concrete, inaccessible areas of ice-condenser containments that are subject to potential corrosions). Furthermore, the licensee should also explore inaccessible degradation-susceptible areas in plant inspections using viable NDE methods.
The ASME Code Section XI, subsection IWE and IWL programs periodically examine, monitor, and manage structural deterioration and aging degradation of the Callaway containment pressure boundary per ASME Code Section XI tables IWE-2500-1 and IWL-2500-1 such that the primary containment can perform its intended function as a leak-tight barrier consistent with the guidance contained in NEI 94-01, Revision 2-A. In the LAR, the licensee provided tables 3.5.2-1 and 3.5.3-1, which summarize IWE and IWL examination requirements applicable to Callaway.
Based on the above, the NRC staff finds that the licensee provided an acceptable level of information regarding recent ASME Code Section XI, subsection IWE and IWL inspections that were evaluated as acceptable and performed in accordance with the ASME Code. Therefore, the licensee has adequately addressed Condition 3.
3.4.4 NEI 94-01, Revision 2-A, Condition 4 Condition 4 of NEI 94-01, Revision 2-A, states:
The licensee addresses any tests and inspections performed following major modifications to the containment structure, as applicable. (Refer to SE Section 3.1.4).
The licensees response to Condition 4 states:
Steam Generator replacements were performed using the installed equipment hatch.
NRC Staff Assessment of Licensees Response to Condition 4 As described in LAR section 3.1.10, the licensee performed the SG replacements using the installed equipment hatch during RF14 in the fall of 2005. Therefore, there has been no major containment repairs or modifications performed on the Callaway containment structure.
Furthermore, as summarized in LAR section 3.1.11, the NRC accepted the licensees proposed change to TS 5.5.16 that allowed plant operation to resume after the SG replacements without performing an ILRT.
Based on the above, the NRC staff finds that the licensee demonstrated that there have been no major containment repairs or modifications performed on the Callaway containment structure. Therefore, the licensee has adequately addressed Condition 4.
3.4.5 NEI 94-01, Revision 2-A, Condition 5 Condition 5 of NEI 94-01, Revision 2-A, states:
The normal Type A test interval should be less than 15 years. If a licensee has to utilize the provision of Section 9.1 of NEI 94-01, Revision 2, related to extending the ILRT interval beyond 15 years, the licensee must demonstrate to the NRC staff that it is an unforeseen emergent condition. (Refer to SE Section 3.1.1.2).
The licensees response to Condition 5 states:
Callaway will follow the requirements of NEI 94-01 Revision 3-A, Section 9.1.
This requirement has remained unchanged from Revision 2-A to Revision 3-A of NEI 94-01.
In accordance with the requirements of NEI 94-01, Revision 2-A, SER [safety evaluation report] Section 3.1.1.2, Callaway will also demonstrate to the NRC staff that an unforeseen emergent condition exists in the event an extension beyond the 15-year interval is required.
NRC Staff Assessment of Licensees Response to Condition 5 The licensees response indicates acknowledgement and acceptance of this NRC staff position; therefore, the licensee has adequately addressed Condition 5.
3.4.6 NEI 94-01, Revision 2-A, Condition 6 Condition 6 of NEI 94-01, Revision 2-A, states:
For plants licensed under 10 CFR Part 52, applications requesting a permanent extension of the ILRT surveillance interval to 15 years should be deferred until after the construction and testing of containments for that design have been completed and applicants have confirmed the applicability of NEI TR 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2, including the use of past containment ILRT data.
The licensees response to Condition 6 states:
Not applicable. Callaway was not licensed under 10 CFR Part 52.
NRC Staff Assessment of Licensees Response to Condition 6 Condition 6 applies only to plants licensed under 10 CFR Part 52. The Callaway license was issued under 10 CFR Part 50 and, therefore, this condition is not applicable.
3.4.7 Conclusion Related to the Six Conditions Listed in NEI 94-01, Revision 2-A, Section 4.1, of the NRC SE The NRC staff evaluated each of the six conditions listed in NEI 94-01, Revision 2-A, section 4.1, of the NRC SE and determined that the licensee adequately addressed each of them. Therefore, the NRC staff finds it acceptable for the licensee to adopt the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008 as part of the implementation document listed in Callaway TS 5.5.16.
3.5 NEI 94-01, Revision 3-A, Conditions The NRC published an SE with limitations and conditions for NEI 94-01, Revision 3, by letter dated June 8, 2012. In that SE, the NRC concluded that NEI 94-01, Revision 3, describes an acceptable approach for implementing the optional performance-based requirements of 10 CFR Part 50, Appendix J, and is acceptable for reference by licensees proposing to amend their containment leakage rate testing TSs, subject to two conditions. The SE was incorporated into Revision 3 and subsequently issued as NEI 94-01, Revision 3-A, on July 31, 2012.
The LAR proposes to use NEI 94-01, Revision 3-A, as the implementation document for the leak-rate testing program. Accordingly, the licensee will be adopting at Callaway, in part, the testing criteria of ANSI/ANS 56.8-2002 as part of its licensing basis. As stated in NEI 94-01, Revision 3-A, section 2.0, Purpose and Scope, where technical guidance overlaps between NEI 94-01, Revision 3-A, and ANSI/ANS 56.8-2002, the guidance in NEI 94-01, Revision 3-A, takes precedence.
3.5.1 NEI 94-01, Revision 3-A, Condition 1 NEI 94-01, Revision 3-A, Condition 1 states:
NEI TR 94-01, Revision 3-A, is requesting that the allowable extended interval for Type C LLRTs be increased to 75 months, with a permissible extension (for non-routine emergent conditions) of nine months (84 months total). The staff is allowing the extended interval for Type C LLRTs be increased to 75 months with the requirement that a licensee's post-outage report include the margin between the Type B and Type C leakage rate summation and its regulatory limit. In addition, a corrective action plan shall be developed to restore the margin to an acceptable level. The staff is also allowing the non-routine emergent extension out to 84-months as applied to Type C valves at a site, with some exceptions that must be detailed in NEI 94-01, Revision 3. At no time shall an extension be allowed for Type C valves that are restricted categorically (e.g., BWR [boiling-water reactor] MSIVs [main steam isolation valves]), and those valves with a history of leakage, or any valves held to either a less than maximum interval or to the base refueling cycle interval. Only non-routine emergent conditions allow an extension to 84 months.
Condition 1 identifies three issues that are required to be addressed:
(1)
The allowance of an extended interval for Type C LLRTs of 75 months requires that a licensees post-outage report include the margin between the Type B and Type C leakage rate summation and its regulatory limit; (2)
A corrective action plan is to be developed to restore the margin to an acceptable level; and (3)
Use of the allowed 9-month extension for eligible Type C valves is only allowed for non-routine emergent conditions, but not for valves categorically restricted and other excepted valves.
The licensees response to Condition 1, Issue 1 states:
The post-outage report shall include the margin between the Type B and Type C MNPLR summation value, as adjusted to include the estimate of applicable Type C leakage understatement, and its regulatory limit of 0.6 La.
The licensees response to Condition 1, Issue 2 states:
When the potential leakage understatement adjusted Types B and C MNPLR total is greater than the Callaway administrative leakage summation limit of 0.5 La, but less than the regulatory limit of 0.6 La, then an analysis and determination of a corrective action plan shall be prepared to restore the leakage summation margin to less than the Callaway leakage limit. The corrective action plan shall focus on those components which have contributed the most to the increase in the leakage summation value and what manner of timely corrective action, as deemed appropriate, best focuses on the prevention of future component leakage performance issues so as to maintain an acceptable level of margin.
The licensees response to Condition 1, Issue 3 states:
Callaway will apply the 9-month allowable interval extension period only to eligible Type C components and only for non-routine emergent conditions. Such occurrences will be documented in the record of tests.
The NRC staff has reviewed the requirements of NEI 94-01, Revision 3-A, against the licensees responses for Issues (1), (2), and (3) of Condition 1. The licensees response indicates that following approval of the requested amendment, the licensees actions will be consistent with the guidance of NEI 94-01, Revision 3-A. The NRC staff notes that revised guidance contained in NEI 94-01, Revision 3-A, section 10.1, Introduction, section 10.2.3.4, Corrective Action, section 11.3.2, Programmatic Controls, and section 12.1, Report Requirements, reflects the NRC staffs SE input pertaining to Issues (1), (2), and (3). The NRC staff concludes that the licensee has acknowledged all the requirements of Condition 1, and that the licensee has established its intent for Callaway to comply with these requirements; therefore, the licensee has adequately addressed Condition 1.
3.5.2 NEI 94-01, Revision 3-A, Condition 2 NEI 94-01, Revision 3-A, Condition 2 states:
The basis for acceptability of extending the ILRT interval out to once per 15 years was the enhanced and robust primary containment inspection program and the local leakage rate testing of penetrations. Most of the primary containment leakage experienced has been attributed to penetration leakage and penetrations are thought to be the most likely location of most containment leakage at any time. The containment leakage condition monitoring regime involves a portion of the penetrations being tested each RF outage, nearly all LLRTs being performed during plant outages. For the purposes of assessing and monitoring or trending overall containment leakage potential, the as-found minimum pathway leakage rates for the just tested penetrations are summed with the as-left minimum pathway leakage rates for penetrations tested during the previous 1 or 2 or even 3 RF outages. Type C tests involve valves which, in the aggregate, will show increasing leakage potential due to normal wear and tear, some predictable and some not so predictable. Routine and appropriate maintenance may extend this increasing leakage potential. Allowing for longer intervals between LLRTs means that more leakage rate test results from farther back in time are summed with fewer just tested penetrations and that total used to assess the current containment leakage potential. This leads to the possibility that the LLRT totals calculated understate the actual leakage potential of the penetrations. Given the required margin included with the performance criterion and the considerable extra margin most plants consistently show with their testing, any understatement of the LLRT total using a 5-year test frequency is thought to be conservatively accounted for. Extending the LLRT intervals beyond 5 years to a 75-month interval should be similarly conservative provided an estimate is made of the potential understatement and its acceptability determined as part of the trending specified in NEI 94-01, Revision 3, Section 12.1.
When routinely scheduling any LLRT valve interval beyond 60-months and up to 75-months, the primary containment leakage rate testing program trending or
monitoring must include an estimate of the amount of understatement in the Type B & C total and must be included in a licensee's post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.
Condition 2 identifies two issues that are required to be addressed:
(1)
Extending the Type C LLRT intervals beyond 5 years to a 75-month interval should be similarly conservative, provided an estimate is made of the potential understatement and its acceptability determined as part of the trending specified in NEI 94-01, Revision 3, Section 12.1 and (2)
When routinely scheduling any LLRT valve interval beyond 60-months and up to 75-months, the primary containment leakage rate testing program trending or monitoring must include an estimate of the amount of understatement in the Type B and Type C total and must be included in a licensees post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.
The licensees response to Condition 2, Issue 1 states:
The change in going from a 60-month extended test interval for Type C tested components to a 75-month interval, as authorized under NEI 94-01, Revision 3-A, represents an increase of 25% in the LLRT periodicity. As such, Callaway will conservatively apply a potential leakage understatement adjustment factor of 1.25 to the actual As-Left leak rate, which will increase the As-Left leakage total for each Type C component currently on greater than a 60-month test interval up to the 75-month extended test interval. This will result in a combined conservative Type C total for all 75-month LLRTs being carried forward and will be included whenever the total leakage summation is required to be updated (either while on-line or following an outage).
When the potential leakage understatement adjusted leak rate total for those Type C components being tested on greater than a 60-month test interval up to the 75-month extended test interval is summed with the non-adjusted total of those Type C components being tested at less than or equal to a 60-month test interval, and the total of the Type B tested components, results in the MNPLR being greater than the Callaway administrative leakage summation limit of 0.50 La, but less than the regulatory limit of 0.6 La, then an analysis and corrective action plan shall be prepared to restore the leakage summation value to less than the Callaway leakage limit. The corrective action plan should focus on those components which have contributed the most to the increase in the leakage summation value and what manner of timely corrective action, as deemed appropriate, best focuses on the prevention of future component leakage performance issues.
The licensees response to Condition 2, Issue 2 states, in part:
A post-outage report shall be prepared presenting results of the previous cycles Type B and Type C tests, and Type A, Type B and Type C tests, if performed during that outage. The report shall show that the applicable performance criteria are met and serve as a record that continuing performance is acceptable.
The report shall also include the combined Type B and Type C leakage summation, and the margin between the Type B and Type C leakage rate summation and its regulatory limit.
At Callaway, in the event an adverse trend in the aforementioned potential leakage understatement adjusted Types B and C summation is identified, then an analysis and determination of a corrective action plan shall be prepared to restore the trend and associated margin to an acceptable level.
At Callaway, an adverse trend is defined as three (3) consecutive increases in the final pre-mode change Types B and C MNPLR leakage summation values, as adjusted to include the estimate of applicable Type C leakage understatement, as expressed in terms of La.
The NRC staff has reviewed the requirements of NEI 94-01, Revision 3-A, against the licensees responses for Issues (1) and (2) of Condition 2. The licensees responses indicate that following approval of the requested amendment, the licensees actions will be consistent with the guidance of NEI 94-01, Revision 3-A. The NRC staff notes that revised guidance contained in NEI 94-01, Revision 3-A, section 11.3.2, Programmatic Controls, and section 12.1, Report Requirements, reflects the NRC staffs SE input pertaining to Issues (1) and (2). The NRC staff concludes that the licensee has acknowledged all the requirements of Condition 2, and that the licensee has established its intent for Callaway to comply with these requirements; therefore, the licensee has adequately addressed Condition 2.
3.5.3 NEI 94-01, Revision 3-A, Limitations and Conditions Conclusion Based on the above evaluation of each condition, the NRC staff determined that the licensee has adequately addressed the conditions in section 4.0 of the NRC SE of NEI 94-01, Revision 3. Therefore, the NRC staff finds it acceptable for the licensee to adopt NEI 94-01, Revision 3-A, as the implementation document listed in Callaway TS 5.5.16.
3.6 PRA of the Proposed Extension of the ILRT Test Intervals 3.6.1 Plant-Specific Risk Evaluation In attachment 1 to the LAR, the licensee provided a plant-specific risk assessment for permanently extending the currently allowed containment Type A ILRT interval from 10 years to 15 years for Callaway.
The licensee stated that the plant-specific risk assessment follows the guidance in NEI 94-01, Revisions 3-A and 2-A; the NEI Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals, dated October 2001 (Reference 26); and RG 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2, dated March 2009 (Reference 27), as applied to ILRT interval extensions, as well as
the guidance outlined in RG 1.174, Revision 3. Additionally, the licensee applied the methodology from the Calvert Cliffs Nuclear Power Plant to estimate the likelihood and risk implications of corrosion-induced leakage of steel liners going undetected during the extended test interval (Reference 28), and the methodology used in EPRI 1018243 (Revision 2-A of EPRI 1009325).
The licensee addressed each of the four conditions for the use of EPRI 1009325, Revision 2-A, which are listed in section 4.2 of the NRC SE dated June 25, 2008. A summary of how each condition is met is provided below.
3.6.2 PRA Technical Adequacy - Condition 1 The first condition stipulates that the licensee submit documentation indicating that the technical adequacy of its PRA is consistent with the requirements of RG 1.200 relevant to the ILRT extension application. This RG describes one acceptable approach for determining whether the technical adequacy of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision-making for light-water reactors.
In section 3.2.4.1, Quality of the PRA, of the SE of EPRI 1009325, Revision 2, the NRC staff stated that Capability Category I of the ASME PRA Standard ASME/ANS-RA-Sa-2009, Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications (Reference 29), shall be applied as the standard for assessing PRA quality for ILRT extension applications since approximate values of core damage frequency (CDF) and large early release frequency (LERF) and their distribution among release categories are sufficient to support the evaluation of changes to ILRT frequencies. The NRC SE also states that the assessment of external events can be taken from existing, previously submitted and approved analyses or other alternate methods of assessing an order of magnitude estimate for contribution of the external event to the impact of the changed interval. The licensee stated that the ILRT interval extension risk assessment is allowed to use the existing internal flooding and fire PRA models and other existing seismic and external hazard evaluations.
The licensee addressed the Callaway PRA technical adequacy in attachment 1 of the LAR, Evaluation of Risk Significance of Permanent ILRT Extension. As discussed in section A of attachment 1 of the LAR, the risk assessment performed to support the Callaway ILRT application utilized the current models for internal events, internal flooding, high winds, fire PRA, and seismic PRA models which have been peer reviewed with no open findings and observations. The licensee stated that all models meet Capability Category II or higher of the NRC-endorsed ASME/ANS PRA Standard.
of the LAR provides a more detailed discussion of the external hazard evaluations and the PRA acceptability for the ILRT interval extension risk impact assessment. The licensee provided information in attachment 1 that demonstrates that the PRA is of sufficient quality and level of detail to support this application and has been subjected to a peer review process assessed against the appropriate standards.
3.6.2.1 Internal Events and Internal Flooding The licensee stated that the results for the internal events and flooding hazard are from the plant-specific PRA models which meet the ASME PRA Standard at Capability Category II or
higher. The licensee stated that its risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant.
The Callaway PRA utilizes the methodology provided in Pressurized Water Reactor Owners Group (PWROG)-18027-NP, Loss of Room Cooling in PRA Modeling, for assessing the loss of room cooling in PRA modeling (Reference 30). The licensee recognized that this process was not an endorsed process until RG 1.200, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 3 (Reference 31), was issued in December 2020. An implementation peer review and associated facts and observation (F&O) closure review have been completed using NRC-approved processes, with no open findings identified against implementation of the method.
3.6.2.2 Fire Hazards The licensee stated that the Callaway fire PRA model meets the expectations for PRA scope and technical adequacy as presented in PRA Standard ASME/ANS RA-Sa-2009 and RG 1.200 to fully support this ILRT extension application. The licensee stated that its risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant.
The internal fire PRA model was developed consistent with EPRI 1011989 / NUREG/CR-6850, Fire PRA Methodology for Nuclear Power Facilities, Volumes 1, 2, and Supplement 1 (References 32, 33, and 34, respectively), and only utilizes methods previously accepted by the NRC. The licensee was approved to implement National Fire Protection Association (NFPA)
Standard 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (NFPA 805), 2001 Edition, in January 2014 (Reference 35) and since that time, new or revised guidance has been specifically addressed through the Callaway PRA maintenance and update process. As part of the transition to NFPA 805, there were several committed modifications and implementation items which have been completed and, therefore, there are no NFPA 805 open items impacting this ILRT extension application.
3.6.2.3 Seismic Hazards The licensee stated that that the Callaway seismic PRA model meets the expectations for PRA scope and technical adequacy as presented in ASME/ANS RA-S CASE 1, Case for ASME/ANS RA-Sb-2013, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications (Reference 36), and RG 1.200, Revision 2 to fully support this ILRT extension application. The licensee stated that its risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant and that there are no open peer review findings for the seismic PRA model.
3.6.2.4 High Winds Hazards The licensee stated that the Callaway high winds PRA model meets the expectations for PRA scope and technical adequacy as presented in PRA Standard ASME/ANS RA-Sa-2009 and RG 1.200, Revision 2 to fully support this ILRT extension application. The licensee stated that its risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant.
The licensee stated that the PRA models have been assessed against ASME/ANS RA-Sa-2009 and RG 1.200 and that there have been several industry peer-reviews between 2009 and 2020, with a mix of full-scope and focused-scope reviews. The licensee further stated that there are no open finding F&Os against any of the models discussed in this ILRT extension application, and all finding F&Os have been independently assessed and closed.
3.6.2.5 PRA Technical Adequacy Conclusion Based on the above, the NRC staff determined that the licensee has assessed its PRA models against the ASME/ANS RA-Sa-2009 PRA standard and RG 1.200 as appropriate and has addressed the relevant findings from the peer reviews and that they have no impact on the LAR.
Therefore, the NRC staff concludes that the PRA models used by the licensee are of sufficient quality to support the changes to ILRT frequencies. Accordingly, the first condition is met.
3.6.3 Estimated Risk Increase - Condition 2 The second condition stipulates that the licensee submit documentation indicating that the estimated risk increase associated with permanently extending the ILRT interval to 15 years is small, consistent with the guidance in RG 1.174, the clarification provided in section 4.2 of the NRC SE of NEI 94-01, Revision 2, and EPRI 1009325, Revision 2-A. Specifically, a small increase in population dose should be defined as an increase in population dose of less than or equal to either 1.0 person-rem per year or 1 percent of the total population dose, whichever is less restrictive. In addition, a small increase in conditional containment failure probability (CCFP) should be defined as a value marginally greater than that accepted in previous one-time 15-year ILRT extension requests, which would require that the increase in CCFP be less than or equal to 1.5 percentage points. Lastly, for plants that rely on containment over-pressure for NPSH for ECCS injection, both CDF and LERF will be considered in the ILRT evaluation and compared with the risk acceptance guidelines in RG 1.174. RG 1.174 defines very small changes in risk as resulting in increases of CDF and LERF of less than 1.0E-6/year and 1.0E-07/year, respectively. Thus, the associated risk metrics include LERF, population dose, CCFP, and delta CDF and LERF.
The licensee reported the results of the plant-specific risk assessment in LAR section 5.2, Analysis. The reported risk impacts are based on a change in the Type A containment ILRT frequency from three tests in 10 years (the test frequency under 10 CFR Part 50, Appendix J, Option A, Prescriptive Requirements) to one test in 15 years and account for the risk from undetected containment leaks due to steel liner corrosion. The following conclusions can be drawn from the licensees analysis associated with extending the Type A ILRT frequency:
- 1. RG 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 defines very small changes in risk as resulting in increases of CDF less than 1E-06 per reactor year and increases in LERF less than 1E-07 per reactor year.
There is no quantifiable change in CDF as a result of the proposed ILRT Type A test interval extension. Therefore, the RG 1.174 acceptance guideline for a very small change in CDF is considered met as the impact on CDF for the Type A test interval extension is negligible. Thus, the relevant acceptance criterion is LERF.
The increase in LERF resulting from a change in the Type A ILRT test interval from 3-in-10 years to 1-in-15 years is estimated as 6.46E-8/year. This value increases
negligibly if the risk impact of corrosion-induced leakage of the steel liners occurring and going undetected during the extended test interval is included.
The impact due to an increase in the Type A ILRT interval to 1-in-15 years is very small when considering only internal events. RG 1.174 states that when the calculated increase in LERF is in the small range of 1.0E-07 per reactor year to 1.0E-06 per reactor year, applications will be considered only if it can be reasonably shown that the total LERF is less than 1.0E-05 per reactor year. When including the impact from internal flood and external events, the change to LERF is in the small range. The resulting total LERF is approximately 4.08E-06/year. The total LERF is below the RG 1.174 acceptance criterion for total LERF of 1.0E-05/year and, therefore, this change satisfies both the incremental and absolute criteria with regards to the RG 1.174 LERF metric.
- 2. The calculated increase in the total 50-mile radius population dose risk for the proposed ILRT Type A interval change from 3-in-10 years to 1-in-15 years is measured as an increase to the total integrated dose risk for all accident sequences influenced by Type A testing. The total 50-mile population dose increase (relative to the base case, with corrosion) is 0.863 percent of the total dose rate. NEI 94-01, Revision 2-A, states that a small population dose is defined as an increase of less than or equal to 1.0 person-rem per year, or less than or equal to one percent of the total population dose, whichever is less restrictive. Thus, the estimated 50-mile population dose increase is small using the guidelines of NEI 94-01, Revision 2-A.
- 3. The increase in the CCFP from the 3-in-10 years to once in 15 years Type A test interval including corrosion effects is 0.913 percent. NEI 94-01, Revision 2-A, states that increases in CCFP of less than or equal to 1.5 percentage points are small. Therefore, this increase is judged to be small.
Based on the risk assessment results, the NRC staff concludes that the increase in LERF is small and within the acceptance guidelines of RG 1.174, and that the increase in the total population dose and the magnitude of the change in the CCFP for the proposed change are also small. The defense-in-depth philosophy is maintained as the independence of barriers will not be degraded because of the requested change, and the use of the quantitative risk metrics collectively ensures that the balance between prevention of core damage, prevention of containment failure, and consequence mitigation is preserved. Accordingly, the second condition is met.
3.6.4 Leak Rate for the Large Pre-Existing Containment Leak Rate Case - Condition 3 The third condition stipulates that the average leak rate for the pre-existing containment large leak rate accident case (i.e., accident case 3b) used by the licensee should be 100 La instead of 35 La. As noted by the licensee in section 4.0, Assumptions and Limitations, of attachment 1 to the LAR, the methodology in EPRI 1009325, Revision 2-A, incorporated the use of 100 La as the average leak rate for the pre-existing containment large leakage rate accident case (accident case 3b), and this value has been used in the plant-specific risk assessment.
Accordingly, the third condition is met.
3.6.5 Containment Overpressure is Relied Upon for ECCS Performance - Condition 4 The fourth condition stipulates that in instances where containment over-pressure is relied upon for ECCS performance, an LAR is required to be submitted. In section 5.2.9, Containment
Overpressure, of attachment 1 of the LAR, the licensee stated that Callaway does not rely on containment overpressure for ECCS performance. Accordingly, the fourth condition is not applicable.
3.7 Technical Conclusion Based on the preceding regulatory and technical evaluations, the NRC staff finds that the licensee has adequately implemented its existing primary containment leakage rate testing program consisting of ILRTs and LLRTs. The results of the recent ILRTs and of the LLRTs combined totals demonstrate acceptable performance and support a conclusion that the structural and leak-tight integrity of the primary containment is adequately managed and will continue to be periodically monitored and managed effectively with the proposed changes. The NRC staff finds that the licensee has addressed the NRC limitations and conditions to demonstrate the acceptability of adopting NEI 94-01, Revision 3-A, and the limitations and conditions identified in the NRC staff SE incorporated in NEI 94-01, Revision 2-A. The NRC staff also finds that the PRA used by the licensee is of sufficient technical adequacy to support the evaluation of changes to ILRT frequency. Therefore, the NRC staff concludes that the proposed changes to Callaway TS 5.5.16 regarding the containment leakage rate testing program are acceptable.
3.8 Evaluation of TS Changes The licensee proposed to change Callaway TS 5.5.16 as described in section 2.1 of this SE.
Based on the preceding regulatory and technical evaluations, the NRC staff concludes that the proposed change to Callaway TS 5.5.16.a to replace the reference to RG 1.163 with a reference to the guidelines contained in NEI 94-01, Revision 3-A, dated July 2012, and the conditions specified in NEI 94-01, Revision 2-A, dated October 2008, is acceptable because the proposed change continues to ensure operation of the facility in a safe manner. Therefore, the NRC staff finds that the Callaway TS 5.5.16.a proposed change will continue to meet 10 CFR 50.36(c)(5).
In addition, the current Callaway TS 5.5.16.a contains exceptions. The LAR proposes to delete exceptions 3 and 4. Specifically, in exception 3, Callaway was excepted from the post-modification ILRT associated with the SG replacement that occurred in 2005, which exception was approved by the NRC in Amendment No. 168 (Reference 37). Callaway exception 4 pertains to the performance of a Type A test no later than October 25, 2014, which exception was approved by the NRC in Amendment No. 195 (Reference 38). These exceptions are associated with activities that have already taken place; therefore, the NRC staff finds that the deletion of Callaway TS 5.5.16.a exceptions 3 and 4 is an administrative action that does not substantively alter TS requirements and is acceptable.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Missouri State official was notified of the proposed issuance of the amendment on August 15, 2023. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 or change surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that
may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration published in Federal Register on January 24, 2023 (88 FR 4218), and there has been no public comment on such finding.
Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
7.0 REFERENCES
- 1.
Bianco, F. J., Union Electric Company, dba Ameren Missouri, letter to NRC, License Amendment Request to Revise Callaway Plant, Unit 1 Technical Specification 5.5.16, Containment Leakage Rate Testing Program, for Permanent Extension of Type A and Type C Leak Rate Test Frequencies (LDCN 2020-0004), dated November 3, 2022 (Agencywide Documents Access and Management System Package Accession No. ML22307A310).
- 2.
Witt, T. A., Union Electric Company, dba Ameren Missouri, letter to NRC, Response to Request for Additional Information Regarding License Amendment Request to Revise Callaway Plant, Unit 1 Technical Specification 5.5.16, Containment Leakage Rate Testing Program, for Permanent Extension of Type A and Type C Leak Rate Test Frequencies (LDCN 2020-0004), dated June 5, 2023 (Package ML23156A635).
- 3.
Witt, T. A., Union Electric Company, dba Ameren Missouri, letter to NRC, Minor Correction to License Amendment Request to Revise Callaway Plant, Unit 1 Technical Specification 5.5.16, Containment Leakage Rate Testing Program, for Permanent Extension of Type A and Type C Leak Rate Test Frequencies (LDCN 2020-0004),
dated July 25, 2023 (Package ML23206A222).
- 4.
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Principal Contributors: D. Nold B. Lee A. Istar S. Lai R. Rodriguez J. Dozier J. Robinson C. Ashley A. Sallman G. Bedi Date: September 25, 2023
- by email OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA*
NRR/DEX/ESEB/BC*
NRR/DEX/EMIB/BC*
NAME MChawla PBlechman ITseng SBailey DATE 8/28/2023 8/28/2023 w/comments 8/29/2023 9/1/2023 OFFICE NRR/DRA/APLB/BC*
NRR/DSS/SCPB/BC*
NRR/DSS/SNSB/BC NRR/DSS/STSB/BC (A)*
NAME JWhitman HWagage for BWittick PSahd MJardaneh DATE 8/30/2023 9/1/2023 8/30/2023 8/29/2023 OFFICE OGC*
NRR/DORL/LPL4/BC*
NRR/DORL/LPL4/PM*
NAME JWachutka JDixon-Herrity MChawla DATE 9/21/2023 9/25/2023 09/25/2023