ML12100A024
| ML12100A024 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, McGuire  |
| Issue date: | 03/28/2012 |
| From: | Repko R Duke Energy Carolinas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML12100A024 (9) | |
Text
REGIS T. REPKO LDuke Vice President EEnergy McGuire Nuclear Station Duke Energy MG01 VP / 12700 Hagers Ferry Rd.
Huntersville, NC 28078 980-875-4111 980-875-4809 fax regis.repko@duke-energy.com March 28, 2012 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
Subject:
Duke Energy Carolinas, LLC (Duke Energy)
McGuire Nuclear Station, Units 1 and 2 Docket Numbers 50-369 and 50-370 Summary Report of Evaluations Performed Pursuant to 10 CFR 50.59 Changes, Tests, and Experiments Pursuant to 10 CFR 50.59(d)(2), attached is a summary report of evaluations performed at McGuire Nuclear Station for the period from January 1, 2011 to December 31, 2011. These evaluations demonstrate that the associated changes do not meet the criteria for license amendments as defined by 10 CFR 50.59(c)(2).
This submittal document contains no regulatory commitments.
If there are any questions or if additional information is needed, please contact M. K. Leisure at (980) 875-5171.
Sincerely, Regis T. Repko Attachment www. duke-energy corn
U.S. Nuclear Regulatory Commission March 28, 2012 Page 2 xc:
V. M. McCree Regional Administrator, Region II U.S. Nuclear Regulatory Commission Marquis One Tower 245 Peachtree Center Ave. NE, Suite 1200 Atlanta, GA 30303-1257 J. H. Thompson (addressee only)
NRC Project Manager U.S. Nuclear Regulatory Commission Mail Stop 0-8 G9A 11555 Rockville Pike Rockville, MD 20852-2738 J. Zeiler NRC Senior Resident Inspector McGuire Nuclear Station
U.S. Nuclear Regulatory Commission March 28, 2012 Attachment Page 1 McGuire Nuclear Station (MNS)
Changes Evaluated Under 10 CFR 50.59 Upgrade W-7300 Process Control System with Ovation DCS Modification MD200243 (Unit 2)
(Action Request Nos. 00253414, 00304908, and 00305014)
The control (non-safety related) portion of the 7300 Process Instrumentation and Control System was replaced with a modern digital Distributed Control System (DCS) using Emerson/Westinghouse Ovation equipment. Ovation is commercial off-the-shelf (COTS) technology that provides a secure open-systems architecture. The Ovation platform is extendable, and incorporates interfaces with most widely adopted communications bus standards, allowing it to interface with "smart" technologies, e.g.
Highway-Accessible Remote Transducer (HART).
The new DCS is capable of performing the same control and monitoring functions as the existing system. In addition, the DCS provides advanced functions commonly found in a modern DCS, such as graphical displays, trending, logging, soft controls, more sophisticated control algorithms, and control algorithm detuning. The DCS provides current and historical plant data to the operator aid computer (OAC) and to the MNS site local area network (LAN) via a secure data server that meets the appropriate industry standards and regulatory guidance for cyber security.
The new DCS is more fault-tolerant than the legacy 7300 system being replaced. The design includes redundant controller pairs, communication buses, and power supplies.
In key control loops, additional field devices were added to increase the level of redundancy among the sensor inputs. An automatic signal selection of process variables (e.g., median select) validates input signals and alarms when a faulty input is detected. Diagnostic capabilities for both internal and external field devices added improved reliability. Instrument channels installed in the reactor protection cabinets that no longer perform a protection function were relocated to the DCS. The potential for undesirable control and protection system interaction has been reduced.
The new DCS automates several tasks currently performed manually in the control room. Manual channel selector switches were replaced with automatic algorithms, relieving the control room operator of manual operator actions when a control loop fails.
Paper chart recorders were replaced with computer-based trending and graphic user interfaces, reducing the clutter on the control board and the maintenance expenditure on chart recorders. The risk of operator errors causing an event has been minimized in the DCS design by following the appropriate human factors engineering standards and regulatory guidance in the development of work station screens and displays.
The new DCS has been evaluated for potential failure modes with new or different consequences. The failures considered include those unique to digital systems and microprocessors, e.g. electro-magnetic and radio frequency interference, electro-static discharges, software configuration management, cyber security threats, and other
U.S. Nuclear Regulatory Commission March 28, 2012 Attachment Page 2 software common cause failure mechanisms. The evaluations concluded that the risk of such failures is minimal and the potential consequences are acceptable.
Revised Accident Analysis for Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position (Action Request No. 00264163)
The purpose of this 50.59 Evaluation is to update the accident analysis for the Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position (hereafter referred to as the "Misloaded Fuel Assembly Accident" or "misload"). The current analysis is based on the original Westinghouse analysis provided in WCAP-9500, "Reference Core Report," (circa 1982). This analysis was considered generic and bounding at the time. In recent years this position has been challenged and a PIP (Duke Energy Corrective Action Program "Problem Investigation Process") was written to address this issue for Duke Energy, including the impact of both the non-instrumented locations and inoperable incore detectors on the original analysis and resulting conclusions. A recognized weakness in the original WCAP was the lack of specified analysis limitations/ assumptions (such as incore detector availability, flux map measurement acceptance criteria, significant fuel type changes involving enrichment/burnable poisons, etc.).
Since the 10 CFR 50.59 screen is retained, and since this evaluation is the result of an affirmative answer to screen question #3, only evaluation question #8 is addressed in this evaluation per Section 4.2.1.3 of NEI 96-07, "Guidelines for 10 CFR 50.59 Implementation," Revision 1. The revised method fully encompasses the original Westinghouse analysis but is much more thorough. The number of potential misload cases analyzed has been greatly increased. The methodology as currently described is very high level. From that perspective, the methodology associated with this change is equivalent. The proposed change uses different codes. The proposed change also uses MARPs (Maximum Allowable Radial Peaks) for the development of operating limits. The revised codes are CASMO 4/SIMULATE 3 MOX. These codes are Studsvik Scandpower products, not Westinghouse products. These codes are NRC-approved for Duke Energy to perform nuclear design analyses for the MNS. CASMO-4 and SIMULATE-3 MOX is the standard methodology used by Duke Energy to perform reload design nuclear calculations. The use of MARPs is consistent with the current Duke Energy Nuclear Design methodology and licensing basis and includes the appropriate NRC-approved departure from nucleate boiling (DNB) uncertainties (including, but not limited to: measurement uncertainty, state point uncertainties, and axial peak uncertainty). The misload analysis uses CASMO 4/SIMULATE-3 MOX and MARPs in the same manner (or a subset of) as used in the reload design process. The only difference is that the core design is slightly altered for the misloaded scenario. All appropriate restrictions on the use of CASMO 4/SIMULATE-3 MOX and MARPs as defined in the applicable Safety Evaluation Reports (SERs) are also met. Thus, CASMO 4/SIMULATE-3 MOX is an NRC-approved methodology which is used based on sound engineering practice, appropriate for the intended application, and within the limitations of the applicable SER. Since CASMO 4/SIMULATE 3 MOX is an NRC approved code for MNS which Duke Energy is qualified to use, the evaluation method is
U.S. Nuclear Regulatory Commission March 28, 2012 Attachment Page 3 satisfied. Since MARPs is an NRC approved methodology for determining operating limits for MNS, which Duke Energy is qualified to use, the evaluation method is again satisfied. Therefore, the proposed activity does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.
Replacement of Nuclear Instrumentation System Source Range and Intermediate Range Excore Detectors and Electronics with Thermo Scientific Fission Chamber Detectors and Electronics Engineering Change EC97371 (Unit 2)
(Action Request No. 00292297)
The Westinghouse excore Nuclear Instrumentation System (NIS) Source Range and Intermediate Range detectors (N31/N35 and N32/N36), including cabling and electronics were replaced with Thermo Scientific (formerly Gamma-Metrics) fission chamber detectors and electronics. The cabling and other preliminary field work was accomplished previously under two (2) nuclear station modifications: NSM-1 2569/P1, and NSM-22569/P1. The new detectors are similar to the existing, wide-range Gamma-Metrics detectors, channels N51/N52, installed previously in response to post-accident monitoring requirements imposed by NUREG-0737, "Clarification of TMI Action Plan Requirements".
The changes entailed revision of Technical Specification (TS) Table 3.3-1 and the associated TS Bases, because the existing NIS Intermediate Range trip setpoints are specified in units of ion chamber amperes (ICA), whereas the replacement Thermo Scientific system is calibrated in units of percent Rated Thermal Power (RTP). These changes were submitted in a License Amendment Request (LAR) dated July 1, 2009, and were approved by the NRC (license amendment dated August 2, 2010). Therefore, the scope of the change made under 10CFR50.59 excludes the specified setpoint changes.
Revise Selected Licensee Commitment (SLC) 16.9.7, SLC Bases, UFSAR, and System Design Bases Document Regarding Standby Shutdown Facility (SSF)
Requirements (Action Request No. 00301375)
This activity entails increasing the required Standby Shutdown Facility (SSF) minimum fuel oil storage capacity requirements, and clarifying the stated mission time and associated margin for supporting SSF equipment within the Updated Final Safety Analysis Report (UFSAR), system Design Bases Document, and the SLC Bases.
The SSF diesel fuel oil capacity increase was required due to completion of a recent diesel fuel oil consumption calculation. The calculation utilized numerous conservative inputs, which were not originally considered. The stated fuel oil supply duration (3.5 days vs. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) in the SLC and UFSAR is also revised to reflect the required SSF mission time, as opposed to the available margin beyond the required mission time.
U.S. Nuclear Regulatory Commission March 28, 2012 Attachment Page 4 The SSF is credited during Station Blackout, fire, and security events. The SSF was designed as a non-seismic, non-safety related standby system. The "design function" of the SSF fuel oil storage tank is to provide adequate fuel volume to support SSF diesel loads for the required system mission time (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> based on limiting Appendix R requirements). Similarly, the "design function" for the embedded condenser circulating water piping is to provide an adequate make-up source to support steam generator make-up for the required system mission time (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> based on limiting Appendix R requirements). Discussion of explicit margin/capability beyond the required SSF mission time was deleted from the UFSAR and SLC bases.
Emergency Core Cooling System (ECCS) Water Management and Containment Spray (NS) Higher Design Temperature Engineering Change EC101539 (Unit 1) and EC101538 (Unit 2)
(Action Request No. 00323982)
The ECCS Water Management initiative delays the Containment Spray (NS) actuation during an accident, which causes the Containment Spray system and Containment Sump (part of the Safety Injection (NI) system) to experience higher peak temperatures than were previously analyzed. The current NS temperature is documented as 190°F and the Containment Sump temperature is 250°F but the original analyses showed that the sump was actually analyzed to 190 0 F. The NS temperature changes to 200°F and the sump temperature changes to 230 0 F. These temperatures are documented in McGuire Calculation MCC-1552.08-00-0403, "Long-Term Containment Response -
Manual NS Initiation". These temperature changes are addressed in a McGuire Licensing Amendment Request (LAR) dated May 28, 2010. The changes will be made prior to the LAR approval and will be a conservative design to the Current licensing Basis.
The new temperatures were used as inputs into stress and hanger analyses which conform to analytical formulas and stress tables required by ASME Section III, 1971 Edition through Winter 1971. The analyses show that the affected portions of the NI and NS systems will remain of robust design and code compliant with these temperature changes.
Elimination of "Water Follow" When Using the Polar Crane for Reactor Vessel Head Lifts Engineering Change EC104511 (Unit 1) and EC104513 (Unit 2)
(Action Request No. 00337314)
Currently, the MNS UFSAR (Section 9.1.5.4, Control of Heavy Lifts Program) requires that the core remains covered with a water cushion, by increasing water level in the reactor cavity, when performing a reactor vessel head lift. This is commonly known as "water follow".
U.S. Nuclear Regulatory Commission March 28, 2012 Attachment Page 5 This procedure can contribute a significant amount of critical path time to an outage. In addition, the time required for the water level to follow the reactor vessel head during the lift leaves the head suspended on the polar crane longer than if the head were lifted as a normal load. The risk of an accident increases with suspension time.
The requirement to use "water follow" stems from analyses done in support of NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," which postulated a drop of the reactor vessel head and evaluated the resulting impacts against ASME Code allowable stresses.
NEI 08-05, "Industry Initiative on Control of Heavy Loads," provides alternative guidance for this evaluation. If the polar crane can be demonstrated to be single failure proof, then the requirement to evaluate a postulated drop may be eliminated on qualifying polar cranes.
To establish a single-failure-proof polar crane design, the crane controls were modified by EC104511 (Unit 1) and EC104513 (Unit 2). The physical modifications were screened out under 10CFR50.59, which determined that the physical changes to the crane are non-adverse. This includes the small amount of Aluminum added to the containment inventory by the new electrical switches on the polar crane. So, these scope elements are not considered in this evaluation.
In addition to the physical modifications, NEI 08-05 specifies analyses and enhanced inspection requirements, both of which are used to credit the determination of single-failure equivalency. Required revisions to crane operating procedures when lifting the reactor vessel head are included.
The single failure proof equivalency allows the reactor head lift to be performed without using water follow. It is considered a change in the evaluation methodology described in the UFSAR. Upon installing the minimum safety features and procedure changes defined in NEI 08-05, and revising the UFSAR, the reactor head may be lifted in accordance with the Initiative, and "water follow" is no longer required. The NRC formally endorsed NEI 08-05, complete with an SER (Safety Evaluation Report). This endorsement can be located on the NRC's website by using Accession Number ML082410532.
McGuire Unit 2 Cycle 21 (M2C21) Reload Safety Evaluation (Action Request No. 00346024)
This activity installs the core designed for McGuire Nuclear Station Unit 2 Cycle 21 (M2C21). The M2C21 Reload Design Safety Analysis Review (REDSAR), performed in accordance with Engineering Directives Manual EDM-501, "Engineering Change Program for Nuclear Fuel", and the M2C21 Reload Safety Evaluation confirm the UFSAR accident analyses remain bounding with respect to predicted M2C21 safety analysis physics parameters (SAPP), and fuel thermal and mechanical performance limits. The SAPP method is described in topical report DPC-NE-3001-PA,
U.S. Nuclear Regulatory Commission March 28, 2012 Attachment Page 6 "Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology."
The M2C21 core reload is similar to past cycle core designs, with a design generated using approved methods. The M2C21 Core Operating Limits Report (COLR) was prepared in accordance with Technical Specification 5.6.5. Additionally, applicable Technical Specifications and the UFSAR have been reviewed and no changes are required for the operation of M2C21. This 10CFR50.59 evaluation concluded that no prior NRC approval is necessary for M2C21 operation.
Update to UFSAR Table 15-12 (Action Request No. 00363701)
An update to UFSAR Table 15-12 has been prepared to report revised values of radiation doses for the fuel handling accidents (FHAs) and weir gate drop (WGD) at the McGuire Nuclear Station. The supporting Alternative Source Term (AST) analysis was revised only to correct the LOCADOSE library files used in this analysis. No other changes were made in revising this AST analysis.
The LOCADOSE library files identify all radioisotopes in the radioactive source terms used in the analysis of radiological consequences of a design basis event (DBEs) and contain information intrinsic to each isotope. In particular, the library files list the number and identity of the radioactive daughter products of each isotope, assuming there are any. These items of information are classified as an element of the AST method. The corrections to the LOCADOSE library files for the AST analysis of FHAs and WGD add daughter isotopes to the radioactive source terms for these DBEs and so produce increases in the resultant offsite and control room radiation doses. This "revision" then is "more conservative than the previous revision of the same methodology."
Accordingly, the corrections to the LOCADOSE library files for the revised AST analysis of the FHAs and WGD supporting the updates to UFSAR Table 15-12 is not a departure from a method of evaluation described in the UFSAR and used in establishing the design basis or in the safety analysis. The update to Table 15-12 may be inserted into the UFSAR without prior approval from the NRC.
McGuire Unit I Cycle 22 (MLC22) Reload Design (Action Request No. 00368070)
This activity installs the core designed for McGuire Nuclear Station Unit 1 Cycle 22 (M1C22). The M1C22 Reload Design Safety Analysis Review (REDSAR), performed in accordance with Engineering Directives Manual EDM-501, "Engineering Change Program for Nuclear Fuel", and the M1C22 Reload Safety Evaluation confirm the UFSAR accident analyses remain bounding with respect to predicted M1C22 safety analysis physics parameters (SAPP), and fuel thermal and mechanical performance limits. The SAPP method is described in topical report DPC-NE-3001-PA, "Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology."
U.S. Nuclear Regulatory Commission March 28, 2012 Attachment Page 7 The M1C22 core reload is similar to past cycle core designs, with a design generated using approved methods. The M1C22 Core Operating Limits Report (COLR) was prepared in accordance with Technical Specification 5.6.5. Additionally, applicable Technical Specifications and the UFSAR have been reviewed and no changes are required for the operation of M1C22. This 10CFR50.59 evaluation concluded that no prior NRC approval is necessary for M1C22 operation.
Implementation of Revision 1-A to WCAP-1 01 25-P-A Addendum 1-A into Methodology Report DPC-NE-2009-PA Revision 3a (Action Request No. 00370273)
The proposed activity being evaluated is a revision to the fuel rod analysis methodology (Section 4.0) in the Westinghouse Fuel Transition methodology report (DPC-NE-2009-PA). The proposed activity removes the transient cladding stress criterion and replaces it with a static cladding stress criterion based on the ASME Boiler and Pressure Vessel Code. Specifically, methodology report DPC-NE-2009-PA is being revised to implement Revision 1-A to WCAP-10125-P-A Addendum 1-A, Extended Burnup Evaluation of Westinghouse Fuel, Revision to Design Criteria, which has been previously approved by the NRC, into the licensing basis. DPC-NE-2009 methods are considered part of the Updated Final Safety Analysis Report (UFSAR) via reference, and as such, are considered "as described in the UFSAR."
In addition, removal of the transient cladding stress criterion from the design criteria of Westinghouse fuel means that it is no longer used to determine the Over Power Delta Temperature (OPAT) trip function setpoints in Technical Specifications (TS).
Specifically, removal of the transient cladding stress criterion means that 1% cladding strain becomes the limiting criterion used to establish the imbalance penalty function (f2AI) provided in the Core Operating Limits Report (COLR) for the OPAT trip function.
This is already consistent with what's written in the TS Bases, but is not consistent with the current wording in DPC-NE-2009.
An alternate cladding stress calculation methodology is needed because PAD Version 10.5.1, which was used for the generic cladding stress analysis, contained an error in the cladding stress equation that resulted in non-conservative Akw/ft limits for cladding stress. Since the OPAT f2AI breakpoints in the COLR were based on the cladding stress calculations per the method in DPC-NE-2009-PA, an operability evaluation was performed and concluded that the current core designs were operable, but also degraded non-conforming (OBDN) since there was a failure to conform to all aspects of the licensing basis in DPC-NE-2009-PA. Therefore, the proposed activity is necessary to clear the OBDN condition.
This 10 CFR 50.59 evaluation is performed in accordance with NEI 50.59 guidance that was endorsed by the NRC via Regulatory Guide 1.187. It was concluded that the proposed activity is not a departure from a method of evaluation described in the UFSAR, and thus does not require NRC review and approval.