ML101170120
| ML101170120 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 04/15/2010 |
| From: | Repko R Duke Energy Carolinas, Duke Energy Corp |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML101170120 (7) | |
Text
REGIS T. REPKO PDuke 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.corn April 15, 2010 U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTN: Document Control Desk Duke Energy Carolinas, LLC (Duke)
McGuire Nuclear Station, Units 1 and 2 Docket Numbers 50-369 and 50-370
SUBJECT:
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 September 1, 2008 to December 31, 2009. 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 K 1
U.S. Nuclear Regulatory Commission April 15, 2010 Page 2 xc:
L. A. Reyes 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. B. Brady NRC Senior Resident Inspector McGuire Nuclear Station
U.S. Nuclear Regulatory Commission April 15, 2010 Attachment Page 1 McGuire Nuclear Station (MNS)
Chanqes Evaluated Under 10 CFR 50.59 McGuire Unit 1 Cycle 20 Core Design (Action Request No. 00238551)
This activity installs the core designed for McGuire Nuclear Station Unit 1 Cycle 20 (M1C20). The M1C20,Reload Design Safety Analysis Review (REDSAR), performed in accordance with Engineering Directives Manual EDM-501, "Engineering Change Program for Nuclear Fuel", and the M1C20 Reload Safety Evaluation confirm that the MNS Updated Final Safety Analysis Report (UFSAR) accident analyses remain bounding with respect to predicted M1C20 safety analysis physics parameters (SAPP) and fuel thermal and mechanical performance limits. The SAPP method is described in Duke topical report DPC-NE-3001-PA, "Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology."
The M1C20 core reload is similar to past cycle core designs, with a design generated using NRC-approved methods. The M1C20 Core Operating Limits Report (COLR) was prepared in accordance with Technical Specification (TS) 5.6.5. Additionally, applicable Technical Specifications and the UFSAR have been reviewed and no changes are required for the operation of M1C20. This 10CFR50.59 evaluation concluded that no prior NRC approval is necessary for M1C20 operation.
Main Generator Protective Relaying Upgrade Modification MD200166 (Unit 2 only; Unit 1 implementation is pending)
(Action Request No. 00243703)
The scope of this modification is to improve the reliability, security and monitoring capabilities of the protective relaying for the main generators by replacing the existing protective relays with a set of multifunction microprocessor-based relays. In addition to replacing the protective functions for the existing relays, the new microprocessor-based relays will include protective functions that are not provided by the existing relaying scheme.
An evaluation of the proposed protective relaying modification has demonstrated that the change will have no adverse impact on any systems, structures, or components (SSCs) with accident mitigation functions, and that the American Nuclear Society (ANS)
Condition II transients that may be initiated by the protective relaying will remain bounded by the present UFSAR analysis.
New potential failure modes and increased susceptibility to existing failures have been evaluated. These include common mode software failure, failures resulting from changing the coincident trip logic, and electromagnetic interference-induced or radio frequency interference-induced failures. When considering the features of redundant
U.S. Nuclear Regulatory Commission April 15, 2010 Attachment Page 2 trains, independent channels, diversity; self-diagnostics; and improved human-machine interfaces, the proposed relaying scheme proves to be as reliable as and more secure than the existing system.
The proposed protective relaying scheme contains protective functions that are not provided with the existing scheme. These include inadvertent energization, reverse power breaker failure and out-of-step protection. Each of the new protective features has been evaluated to ensure the proposed changes will not lead to additional challenges or otherwise adversely impact SSCs that serve to mitigate transients initiated by protective relay action.
The evaluation demonstrates that the proposed main generator protective relaying modification will preserve the current licensing basis. The activity will not create more than a minimal increase in the frequency or consequences of accidents or malfunctions of SSCs important to safety. The proposed activity will not create the potential for a new type of unanalyzed event, has no impact on the fission product barriers, and does not affect evaluation methodology. Therefore under 10 CFR 50.59, it is permissible to implement this modification without prior approval from the NRC.
Update to UFSAR Table 15-12 Post FHA-WGD Control Room Dose (Action Request No. 00256925)
An update to Table 15-12 of the McGuire Nuclear Station Updated Final Safety Analysis Report (UFSAR) has been prepared. This update reports revised values for radiation doses to the control room operators for the fuel handling accidents (FHAs) and weir gate drop (WGD). These control room radiation doses were recalculated to account for increases in the atmospheric dispersion factors for transport of fission products to the control room outside air intakes (control room x/Qs).
The actual change made to the control room x/Qs was reported to the NRC (letter from Bruce H. Hamilton to the United States Nuclear Regulatory Commission dated December 17 2008) and so was not evaluated. Only the effects of the changes to the control room x/Qs (particularly the control room x/Qs for fission product releases from the equipment hatch and unit vent stack) were evaluated. No change to any system, structure, or component (SSC) was included with the change to the control room x/Qs.
The update to the UFSAR did not include any change to the description of the configuration or operation of any SSC. The increases in radiation doses to the control room operators for the FHAs and WGDs were found to be minimal. UFSAR Table 15-12 may be updated without prior approval from the NRC.
UFSAR Change for Section 9.2.2, Procedure Change for API1&2/AI55001020, and Technical Specification Bases 3.7.7 Update (Action Request No. 00266451)
The changes are to update Section 9.2.2 and related tables of the UFSAR, update the Bases for Technical Specification (TS) 3.7.7 and revise the "Loss of RN" emergency
U.S. Nuclear Regulatory Commission April 15, 2010 Attachment Page 3 procedures AP/1 (2)/A/5500/020. NRC Inspection Report 05000369/2008002 and 05000370/2008002 dated April 24, 2008 identified a non-cited violation (NCV) of Technical Specification 5.4.1 .a for failure to adequately establish and maintain procedures required by Regulatory Guide 1.33, Appendix A, Section 5, "Procedures for Abnormal Conditions". Specifically, loss of Nuclear Service Water (RN) procedures were not established and maintained with an adequate safety analysis for the sharing of RN between units. In response to the NCV, the abnormal procedures for loss of RN are revised. The UFSAR and the TS Bases are revised to incorporate the McGuire response to Generic Letter (GL) 91-13; and enhance the discussion of the shared portion of the RN system and the use of the crossover feature of the RN system to address the beyond design basis event of loss of all RN on one unit.
Duke Methodology Report DPC-NE-3002-A, Revision 4a (Action Request No. 00273738)
The purpose of this 50.59 Evaluation is to update Duke Methodology Report DPC-NE-3002-A, "UFSAR Chapter 15 System Transient Analysis Methodology."
Section 5.4 of DPC-NE-3002-A is the single rod withdrawal accident (UFSAR Section 15.4.3.d). The pressurizer level initial condition in Section 5.4.2 of DPC-NE-3002-A is in error as verified in the analyses of record (AOR). The conservative assumption is low initial pressurizer level, and the text is changed to reflect this. Additionally, the pressurizer pressure control boundary condition in Section 5.4.4 of DPC-NE-3002-A states that pressurizer spray enabled and pressurizer power operated relied valves (PORVs) disabled is conservative. The AOR all perform sensitivities on spray and PORVs and come to different conclusions as to which combination is conservative. It is noted that the UFSAR states that a sensitivity study is performed, which is consistent with the AOR. To be consistent with the UFSAR, DPC-NE-3002-A is changed to make it clear that a sensitivity study is performed to determine the most conservative combination. Since this evaluation is the result of an affirmative answer to screen question # 3 (Evaluation Methodology), only evaluation question #8 is addressed in this evaluation per Section 4.2.1.3 of NEI 96-07, Revision 1.
The purpose of this 50.59 Evaluation is to update the methodology report DPC-NE-3002-A. This involved revising or replacing an evaluation methodology described in the UFSAR that is used in establishing the design basis or used in the safety analysis. However, per NEI 96-07 Section 4.3.8, changes are not considered departures from a method of evaluation described in the UFSAR for a methodology revision that is documented as providing results that are essentially the same as, or more conservative than as defined in NEI-96-07, either the previous revision of the same methodology or another methodology previously accepted by NRC through issuance of an SER. DPC-NE-3002-A provides the NRC-approved methodology for the single rod withdrawal accident (as well as other accidents). By changing the assumed initial pressurizer level from a high value to a low value, a conservative system response is obtained for the UFSAR 15.4.3.d analysis. Also, stipulating that a sensitivity study is performed on the pressurizer spray and PORVs assumption ensures that a conservative system response is obtained for the UFSAR 15.4.3.d analysis. Thus, both changes to the methodology report were specifically made to ensure more conservative results.
U.S. Nuclear Regulatory Commission April 15, 2010 Attachment Page 4 Consequently, this activity does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design basis or iri the safety analysis.
McGuire Unit 2 Cycle 20 Core Design (Action Request No. 00282537)
This activity installs the core designed for McGuire Nuclear Station Unit 2 Cycle 20 (M2C20). The M2C20 Reload Design Safety Analysis Review (REDSAR), performed in accordance with Engineering Directives Manual EDM-501, "Engineering Change Program for Nuclear Fuel", and the M2C20 Reload Safety Evaluation confirm the UFSAR accident analyses remain bounding with respect to predicted M2C20 safety analysis physics parameters (SAPP) and fuel thermal and mechanical performance limits. The SAPP method is described in Duke topical report DPC-NE-3001-PA, "Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology."
The M2C20 core reload is similar to past cycle core designs, with a design generated using approved methods. The M2C20 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 M2C20. This 10 CFR 50.59 evaluation concluded that no prior NRC approval is necessary for M2C20 operation.
Evaluation of Alternative Shutdown Boron Concentration Methodology for Revision 2a to Duke Methodology Report DPC-NF-2010 (Action Request No. 00289680)
The purpose of this 10 CFR 50.59 evaluation is to determine whether a license amendment request is required to update the methodology used to calculate shutdown boron concentrations in the methodology report DPC-NF-2010-A, "Nuclear Physics Methodology for Reload Design." An alternate approach for calculating shutdown boron concentrations is added to be consistent with current practice. In this approach, the all rods in (ARI) boron concentration corresponding to the appropriate shutdown margin (1.3% Ap or 1.0% Ap) is initially calculated, and then adjusted by an equivalent boron concentration to account for a stuck rod and 10% of the ARI less the highest worth stuck rod worth. The alternate approach differs from the described approach in that an ARI critical boron concentration is initially calculated versus calculating an ARI highest worth stuck rod out critical boron concentration. In the alternate approach, both the stuck rod worth and 10% of the ARI less highest stuck rod out worth are converted to a boron concentration using an appropriate boron worth, and added to the ARI critical boron concentration. In the original approach, both the magnitude of the shutdown margin (1.3% Ap or 1.0% Ap) and 10% of the ARI less stuck rod out worth are converted to a boron concentration, and then added to the ARI stuck rod out critical boron concentration. The condition at which the differential boron worth is calculated between the two approaches is also different. Both methods preserve the essential elements of the method. These elements consist of increasing the ARI boron concentration by
U.S. Nuclear Regulatory Commission April 15, 2010 Attachment Page 5
- 1) the stuck rod worth, 2) 10% rod worth allowance calculated relative to the ARI(N-1) worth, and 3) a factor of safety.
This alternative approach for calculating shutdown boron concentrations is considered a change in methodology used to maintain TS 3.1.1 shutdown margin, and therefore requires an evaluation. NEI 96-07, Section 4.3.8, states that changes to a methodology are not considered a departure from a method of evaluation provided the change to elements of the analysis method yield results that are essentially the same as, or more conservative than, either the previous revision for the same methodology or another methodology previously accepted by the NRC through issuance of a Safety Evaluation Report. Demonstration calculations were performed as a function of burnup and reactor coolant temperature to demonstrate equivalency between the alternative methodology relative to the methodology described in DPC-NF-201 0-A. The results of this calculation showed that the two methods were essentially the same. Therefore, this activity does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design basis or in the safety analysis.