ML091540570
| ML091540570 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 06/24/2009 |
| From: | Wang A Plant Licensing Branch IV |
| To: | Entergy Operations |
| Wang, A B, NRR/DORL/LPLIV, 415-1445 | |
| References | |
| TAC MD7966 | |
| Download: ML091540570 (27) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June 24, 2009 Vice President, Operations Entergy Operations, Inc.
River Bend Station 5485 US Highway 61 N S1. Francisville, LA 70775
SUBJECT:
RIVER BEND STATION, UNIT 1 - ISSUANCE OF AMENDMENT RE: MAIN TURBINE BYPASS SYSTEM (TAC NO. MD7966)
Dear Sir or Madam:
The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 163 to Facility Operating License No. NPF-47 for the River Bend Station, Unit 1. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated January 25, 2008, as supplemented by letters dated April 14 and 29, 2009.
The amendment revises TS 3.7.5, "Main Turbine Bypass System." The change provides an alternative to the existing Limiting Condition for Operation for the Main Turbine Bypass System (MTBS). The revised TS will require that the MTBS be operable or that the Average Planar Linear Heat Generation Rate, the Minimum Critical Power Ratio, and the Linear Heat Generation Rate limits for the inoperable MTBS be placed in effect as specified in the Core Operating Limits Report.
A copy of our related Safety Evaluation is enclosed. The Notice of Issuance will be included ill the Commission's next biweekly Federal Register notice.
Sincerely,
~W~
Alan B. Wang, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-458
Enclosures:
- 1. Amendment No. 163 to I\IPF-47
- 2. Safety Evaluation cc w/encls: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 ENTERGY GULF STATES LOUISIANA. LLC AND ENTERGY OPERATIONS, INC.
DOCKET NO. 50-458 RIVER BEND STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 163 License No. NPF-47
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Entergy Operations, Inc. (the licensee), dated January 25, 2008, as supplemented by letters dated April 14 and 29, 2009, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, as amended, 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 Commission's regulations; D. The issuance of this license 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 Commission's regulations and all applicable requirements have been satisfied.
Enclosure 1
-2
- 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 Facility Operating License No. NPF-47 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. 163 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. EOI shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3. The license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION A.t-i-r }t~
Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Facility Operating License No. NPF-47 and Technical Specifications Date of Issuance: June 24, 2009
ATTACHMENT TO LICENSE AMENDMENT NO. 163 FACILITY OPERATING LICENSE NO. NPF-47 DOCKET NO. 50-458 Replace the following pages of the Facility Operating License No. NPF-47 and 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.
Facility Operating License Remove
-3 Technical Specifications Remove 3.7-14 3.7-14 3.7-14a
-3 (3) EOI, pursuant to the Act and 10 CFR Part 70, to receive, possess and to use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4) EOI, 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; (5) EOI, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6) EOI, 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 license shall be deemed to contain and is subject to the conditions specified in the Commission's 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 EOI is authorized to operate the facility at reactor core power levels not in excess of 3091 megawatts thermal (100% rated power) in accordance with the conditions specified herein. The items identified in Attachment 1 to this license shall be completed as specified. Attachment 1 is hereby incorporated into this license.
(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 163 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. EOI shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
Amendment No. 163
Main Turbine Bypass System 3.7.5 3.7 PLANT SYSTEMS 3.7.5 Main Turbine Bypass System LCO 3.7.5 A, The Main Turbine Bypass System shall be OPERABLE.
B. The following limits for inoperable Main Turbine Bypass System, as specified in the COLR are made applicable.
APPLICABI L1TY: THERMAL POWER:?: 23.8 RTP.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A, Main Turbine Bypass A,1 Restore Main Turbine 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> System inoperable. Bypass System to OPERABLE status.
OR A,2 Apply the APLHGR, LHGR 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and MCPR limits for inoperable Main Turbine Bypass System as specified in the COLR.
B. Required Action and B.1 Reduce THERMAL POWER 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion to < 23.8% RTP.
Time not met.
RIVER BEND 3.7-14 Amendment No.84-444, 163
Main Turbine Bypass System 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 Verify one complete cycle of each main turbine 31 days bypass valve.
SR 3.7.5.2 Perform a system functional test. 18 months SR 3.7.5.3 Verify the TURBINE BYPASS SYSTEM RESPONSE 18 months TIME is within limits.
RIVER BEND 3.7-14a Amendment No. ~, 163
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 163 TO FACILITY OPERATING LICENSE NO. NPF-47 ENTERGY OPERATIONS, INC.
RIVER BEND STATION, UNIT 1 DOCKET NO. 50-458
1.0 INTRODUCTION
By application dated January 25, 2008 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML080440293) (Reference 17). as supplemented by letters dated April 14 and 29, 2009 (ADAMS Accession Nos. ML091110056 and ML091260596, respectively) (References 15 and 18, respectively), Entergy Operations, Inc. (the licensee),
requested changes to the Technicai Specifications (TSs) for the River Bend Station, Unit 1 (RBS). The supplementalle't~rs dated April 14 and 29, 2009, 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) staffs original proposed no significant hazards consideration determination as published in the Federal RegIster on March 11, 2008 (73 FR 13023).
The proposed changes would revise TS 3.7.5, "Main Turbine Bypass System." Specifically, the revised TS 3.7.5 will require that either the Main Turbine Bypass System (MTBS) be OPERABLE or that Average Planar Linear Heat Generation Rate (APLHGR), Minimum Critical Power Ratio (MCPR), and Linear Heat Generation Rate (LHGR) limits for the inoperable MTBS be placed in effect as specified in the Core Operating Limits Report (COLR). The change provides an alternative to the existing Limiting Condition for Operation (LCO) for the MTBS.
2.0 REGULATORY EVALUATION
In Section 50.36, "Technical specifications," of Title 10 of the Code of Federal Regulations (10 CFR). the Commission established its regulatory requirements related to the content of TSs.
Pursuant to 10 CFR 50.36, TSs are required to include items in the following five specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) LCOs; (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls. The rule does not specify the particular requirements to be included in a plant's TSs.
Paragraph 50.36(c)(2)(ii)(C) of 10 CFR specifies that a TS LCO must be established for a structure, system, or component that is part of the primary success path and which functions or Enclosure 2
- 2 actuates to mitigate a design basis event'or transient that either assumes the failure or presents a challenge to the integrity of a fission product barrier. Paragraph 10 CFR 50.36(c)(3) specifies that SRs are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components would be maintained within safety limits, and that the LCOs will be met. As required by these two sections of 10 CFR 50.36, an LCO and SRs for the operability of the MTBS are specified in TS 3.7.5 to assure the quality of the MTBS functions.
The NRC provides guidance for the implementation of the requirements of 10 CFR 50.36 to General Electric (GE) BWR-6 plants in NUREG-1434, Revision 3, "Standard Technical Specifications General Electric Plants, BWRl6," dated June 2004 (Reference 19). Since the RBS is a GE BWRl6 plant, the NRC staff Will use the NUREG-1434, Revision 3, guidance for the review of the proposed TS changes.
In Appendix A to 10 CFR Part 50, General Design Criterion (GDC) 10, "Reactor design,"
requires that the reactor core and associated coolant control, and protection systems be designed with appropriate margins to assure that specified acceptable fuel design limits
[SAFDLs] are not exceeded during any conditions of normal operation, including the effects of anticipated operational occurrences (AOOs). In the application of boiling-water reactors (BWRs), the Safety Limit Critical Power Ratio (SLCPR) is established to assure compliance with SAFDLs. Above the SLCPRs, the fuel rods would not experience a boiling transition during normal operation conditions and AOOs. The LHGR and APLHGR safety limits are the other SAFDLs specified to avoid the fuel design limits from being exceeded. In support of safe plant operations, the MTBS is used to provide consequence mitigation for applicable transients assumed in safety analysis to assure that the MCPR, LHGR, and APLHGR safety limits are not exceeded during normal operation conditions and AOOs. When the MTBS is inoperable, compensation to the operating limits of the MCPR, LHGR, and APLHGR is made for maintaining the margin to the operating limits assumed in the safety analysis.
NRC Generic Letter (GL) 88-16, "Removal of Cycle-Specific Parameter Limits From Technical Specifications," dated October 3, 1988 (Reference 20), allows licensees to include the cycle specific parameters (Le., MCPR, LHGR, and APLHGR etc.) in the COLR, provided the changes in the parameters are determined using an NRC-approved methodology and consistent with all applicable limits of the analysis of record.
3.0 TECHNICAL EVALUATION
The proposed changes would allow revision of reactor operational limits, as specified in the RBS COLR, to compensate for the inoperability of the MTBS. The MTBS is designed to control steam pressure when reactor steam generation exceeds turbine capacities during plant startup, sudden load reduction, and cooldown. It allows excess steam flow from the reactor to condenser without going through the turbine. The bypass capacity of the steam is about 9.5 percent of the nuclear steam supply system-rated steam flow. Sudden load reductions within the capacity of the steam bypass can be accommodated without a reactor trip. The bypass valve assembly consists of a single multi-valve manifold. There are two bypass valves connected to the main steam lines between the main steam isolation valves and the turbine stop valves. The bypass valves are controlled by the pressure regulation function of the turbine pressure regulator and control system. The bypass valves are normally closed, and the pressure regulator controls the turbine control valves, directing all steam to the turbine. If the speed governor or the load limiter restricts steam flow to the turbine, the pressure regulator
-3 controls the system pressure by opening the bypass valves that vent steam flow to the two main condenser shells.
The MTBS is also required to be operable to limit the reactor pressure and power increases during applicable transients assumed in the accident analysis so that SAFDLs are not exceeded during ADOs. Specifically, the mitigation function of the MTBS is to assure that the safety limits of the MCPR, the LHGR, and the APLHGR are not exceeded.
The MCPR is a ratio of the fuel assembly power that results in the onset of boiling transition to the actual fuel assembly power. Operating limits on the MCPR are specified to assure that no fuel damage occurs during ADOs.
The LHGR is a measure of the heat generation rate of a fuel rod in a fuel assembly at any axial location. Operating limits on the LHGR are specified to assure that the fuel design limits are not exceeded anywhere in the core during ADOs.
The APLHGR is a measure of the average LHGR of all the fuel rods in a fuel assembly at any axial location. Operating limits on the APLHGR are specified to assure that fuel design limits are not exceeded anywhere in the core during ADOs and that the peak cladding temperature during the postulated loss-of-coolant accident (LOCA) does not exceed the limit specified in 10 CFR, Section 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors."
With the MTBS inoperable, the licensee would assess the need to revise the operating limits of the MCPR, LHGR, and APLHGR in the COLR to assure that these values remain within the accident analysis. The proposed TS changes would avoid a reactor trip with the MBTS being inoperable, if the operating MCPR, LHGR, and APLHGR limits are within the ranges specified in the COLR. The changes would increase plant flexibility.
3.1 Evaluation of AREVA NP Methods In its supplemental letter dated April 14,2009, the licensee identified the transient analyses affected by the proposed amendment. The safety analyses for the RBS with the MTBS-out-of service (MTBS-OOS) were performed in accordance with NRC-approved transient methods.
The methods referenced in the safety analysis are provided in Table 1 below.
Table 1: AREVA NP Approved Analysis Methods Analysis Methodology NRC Approval References Nuclear Design CASM04/MICROBURN-B2 10 Transient Reactor COTRANSA2 8, 11, and 14 Power/Pressure Transient CPR XCOBRA-T 6 and 14 rCritical Power Ratiol Transient LHGR XCOBRA-T 6 and 14 RODEX2A, RAMPEX, and LHGR Limits 1, 2, 3, 4, and 5 COLAPX
- 4 The licensee also provided clarifying information to ensure that the approved methods were executed appropriately in accordance with the NRC staff approval of the methods. The NRC staff review of this information is discussed in the following sections.
3.1.1 Nuclear Design The nuclear design calculations are performed using the CASM04/MICOBURN-B2 code system. The NRC staff approval of CASM04/MICOBURN-B2 for this purpose is documented in Reference 10. Nuclear design calculations are performed off-line to generate input for downstream analyses and are performed on-line as part of the core monitoring methodology.
The nuclear design calculations determine the reactor power distribution and kinetics parameters for the transient calculations. The on-line calculated power distribution is used to monitor key variables against relevant safety limits.
3.1.1.1 Off-line Calculations Off-line calculations are performed by the licensee using the nuclear design methods to establish the initial conditions for transient analyses. The NRC staff reviewed these calculations to ensure that: (1) sufficient modeling rigor is included in the methods to meet the conditions and limitations specified in the NRC staff's approving safety evaluation (Reference 10), and (2) that the initial conditions specified for the transient safety analyses are conservative relative to the allowable operating flexibility.
The NRC staff requested that the licensee demonstrate that the transversing incore probe (TIP) uncertainties were sufficiently small relative to the conditions imposed on the nuclear design methods in Reference 10. The response provides several statistical tests of five TIP measurements that were performed during RBS Cycle 15. The results of the test confirm that the TIP uncertainties are within the bounds of the NRC staff's condition on CASM04/MICROBURN-B2 (Entergy letter dated April 14,2009). Therefore, the NRC staff concludes that the licensee has adequately demonstrated compliance with the NRC staff's limitations and conditions on the approval of this methodology.
The NRC staff requested that the licensee describe how upstream calculations performed with CASM04/MICROBURN-B2 are conservative given the operating flexibility at RBS. The limiting conditions for transients tend to occur at the end-of-cycle (EOC). During cycle operation, the plant operates within an allowable flow control window to control reactivity. This gives the plant operating flexibility during the cycle to elect how to control excess reactivity. The EOC transient analyses, therefore, must be initiated from a condition that bounds the allowable cycle operation. The licensee stated in its supplemental letter dated April 14, 2009, that the safety analyses are performed by simulating cycle operation with an aggressive burn of the lower portion of the core. This is achieved by simulating the cycle exposure at the minimum allowable flow condition for the entire cycle. The burn-out of the bottom of the core ensures that the reactor power at the EOC is shifted axially upwards (Entergy letter dated April 14, 2009). The axially upward shifted power shape is conservative for the transient calculations. On the basis that the analytical power shape is conservative, and the degree of the upward peaking is consistent with the maximal upward peaking achievable within the allowable operating domain, the NRC staff concludes that this analysis approach is acceptable.
-5 3.1.1.2 On-line Calculations (Monitoring)
In its letter dated April 14, 2009, in response to an NRC staff request for additional information regarding the core monitoring methodology, the licensee stated that the core monitoring system is used to monitor the thermal margin relative to the maximum linear heat generation rate (MLHGR) limit.
The NRC staff requested that the licensee clarify the use of the core monitoring software in terms of modeling parameters affecting the capability of the system to determine the local power distribution and that the licensee verify that the effects of channel bow, local power range monitor (LPRM) rod power biases, and part-length fuel rod (PLFR) fission gas plena, respectively, were incorporated in the core monitoring system.
In Entergy letter dated April 14, 2009, the licensee stated that the effects of channel bow were negligible relative to the margin to the 1 percent plastic strain and fuel centerline melt criteria.
The licensee's response (Entergy letter dated April 14,2009) refers to the AREVA NP critical power methodology (Reference 9). Reference 9 is specific to AREVA 8X8 and 9X9 fuel designs. The licensee's response confirms that similar margins are available for the 1OX1 0 design of the ATRIUM-1 0 thus justifying the continued applicability of the basis provided in Reference 9. The NRC staff concludes that the basis is equally applicable and, therefore, the licensee's response is acceptable.
The NRC staff requested additional information regarding the treatment of LPRM rod power biases. The presence of the LPRM instrument in the bypass has an impact on the neutron moderation near the bundle corner. The nuclear design methods have an option to account for the presence of the LPRM on rod power. The licensee verified that the LPRM rod power biases are explicitly modeled in the core monitoring for RBS (Entergy letter dated April 14, 2009) and, therefore, the NRC staff concludes that the response is acceptable.
The licensee states that the core monitoring does not consider the fission gas plena above the PLFRs. The response refers to sensitivity studies that show the explicit modeling of the PLFR fission gas plena has a negligible impact on the predicted LHGR (Entergy letter dated April 14, 2009). The sensitivity was calculated using an equilibrium core model with and without PLFR plena. The results confirm that the impact on the LHGR margin is negligible and, therefore, the NRC staff concludes that the response is acceptable.
3.1.2 Transient Reactor Power and Pressure Response To evaluate the plant transient response, the coupled nuclear/thermal hydraulic COTRANSA2 code is used. COTRANSA2 was reviewed and approved by the NRC staff to replace the COTRANSA methodology in Reference 8. The NRC staff clarified the scope of its approval of COTRANSA2 and XCOBRA-T in Reference 14. The COTRANSA2 methodology specifically evaluates the transient reactor power and pressure in response during an AOO. These parameters are fed into the downstream XCOBRA-T code to determine the transient critical power ratio (CPR) and LHGR. The NRC staff requested additional information regarding the execution of COTRANSA2, as discussed below.
-6 The NRC staff requested that the licensee provide additional information regarding the inputs used in the safety analysis performed using COTRANSA2. In particular, the NRC staff requested that the licensee confirm that the code inputs include appropriate conservatism consistent with the plant TSs and Reference 11. The response provided by the licensee confirms that the appropriate inputs were used in the safety analysis (Entergy letter dated April 14, 2009), and, therefore, the usage is acceptable.
The NRC staff requested additional information regarding particular models in COTRANSA2. In particular, the NRC staff requested that the licensee provide descriptive details of the fuel rod parameters and the void quality correlation. COTRANSA2 includes input options for the fuel rod parameters as well as the void quality correlation. The licensee clarified in its supplemental letter dated April 14, 2009, that the fuel rod parameters are based on approved RODEX2 models. The licensee also provided justification of the conservatism of the methodology relative to known biases in the void quality correlation (Entergy letter dated April 14, 2009).
The NRC staff reviewed the portion of the response dedicated to the thermal-mechanical model.
The response provides the results of sensitivity studies that demonstrate that biasing the fuel rod heat resistance parameters in either direction does not produce a net impact on the transient critical power ratio. The primary reason is due to competing effects in terms of the local heat flux and the plant void reactivity response. Therefore, the NRC staff concludes that it is appropriate to treat these parameters in COTRANSA2 as the average values.
The NRC staff reviewed the licensee's void quality correlation. In the Entergy letter dated April 14, 2009, the licensee provided information justifying that the conservatism of the 110 percent integral thermal power multiplier is sufficient to bound any uncertainty or non conservatism introduced by the void quality correlation bias. On the basis of the sensitivity analysis results quoted in the licensee's response, the NRC staff concurs that the conservatism of the integral thermal power multiplier is sufficient to bound any bias in the predicted thermal margin using the void quality correlation and, therefore, the NRC staff concludes that its usage is acceptable.
In response to the NRC staffs question if the transient analyses were performed at 102 percent of the licensed thermal power, the licensee stated that the transient analyses, except for the American Society of Mechanical Engineers (ASME) overpressure analysis, are performed from initial conditions of 100 percent licensed thermal power and that model conservatisms or statistical analyses account for the impact of the power uncertainty (Entergy letter dated April 14, 2009). The NRC staff concludes that this approach is acceptable because the power uncertainty is still captured in the safety analysis.
The NRC staff reviewed the clarifications provided in the licensee's supplemental letter dated April 14, 2009, and found that the analysis conditions had been adequately justified and, therefore, the NRC staff finds that the usage of COTRANSA2 at RBS is acceptable.
3.1.2.1 Transient CPR and LHGR Output from COTRANSA2 is used with the NRC-approved XCOBRA-T core performance code to determine the transient CPR and the transient LHGR. The NRC staff requested additional information regarding the transient analysis method, as discussed below.
-7 In particular, the NRC staff requested that the licensee specify the code used to perform the transient LHGR analysis. The licensee stated it used XCOBRA-T (Entergy letter dated April 14, 2009). The NRC staff approval of XCOBRA-T is documented in Reference 6.
The NRC staff also reviewed the appropriateness of assumptions in the XCOBRA-T transient LHGR methodology and the inputs to the code, as discussed below.
3.1.2.1.1 Methodology Updates The NRC staff requested that the licensee describe any differences in the version of XCOBRA-T used to perform the safety analysis and the code described in the approved XCOBRA-T licensing topical report (LTR). The scope of the NRC staff request included any updates to the code to support modeling of modern fuel bundle designs such as ATRIUM-10. The licensee's response details two revisions to the code to support modern fuel bundle designs (Entergy letter dated April 14, 2009).
The first modification is a generalization of the rod power model. The generalization allows for axial variation in key thermal hydraulic parameters such as the heated and wetted perimeter and the number of heated rods. The response provides an updated version of Equation 2.130 from Reference 6. The first noted difference is the correction of a typographical error in the original LTR. The second difference is the specification that the parameters are nodal-specific values that vary axially within the bundle to account for PLFRs.
The second modification is an update to the local spacer loss coefficients. A two-phase multiplier was developed based on comparisons of predicted and measured pressure losses at Karlstein Thermal Hydraulic Test Facility (KATHY) and the Portable Hydraulic Test Facility (PHTF). The licensee provided the results of the tests in its response (Entergy letter dated April 14, 2009). The NRC staff reviewed the modified loss coefficient multiplier and the test results.
The qualification data are sufficient to justify that the modified loss coefficient provides acceptable predictions of the local losses consistent with the expected accuracy of pressure drop models. The NRC staff concludes that the inclusion of the multiplier eliminates biases in the predicted pressure losses, and, therefore, is appropriatefor inclusion in analyses of the ATRIUM-10 fuel design.
The NRC staff, therefore, concludes that the response provides an adequate description of the methodology updates and that these revisions are appropriate and acceptable for the analyses for the use of ATRIUM-10.
3.1.2.1.2 Heat Deposition Assumptions The NRC staff requested that the licensee clarify the usage of XCOBRA-T modeling capabilities in terms of direct heat deposition. The approved LTR describes methods in XCOBRA-T for detailed modeling of the direct energy deposition in various core components (Reference 6).
The licensee stated in its supplemental letter dated April 14, 2009, that several simplifying assumptions were made regarding the analysis of direct heat deposition in the XCOBRA-T analysis. First, a non-conservative simplification of the XCOBRA-T treatment of the fuel pin was made. Second, the power fractions (heat deposited to fuel, coolant, and bypass) were.
-8 generically determined and are not specific to the RBS core loading (Entergy letter dated April 14,2009).
The NRC staff reviewed the assumption regarding the cladding heat deposition. For analyses of critical power or transient LHGR, assigning a larger fraction of direct heat deposition to the cladding is conservative as this maximizes the transient prediction of the cladding surface heat flux. The response to request for additional information (RAI) 6.15 provides sensitivity analyses that demonstrate that assigning a conservatively high fraction of the heat directly to the cladding has only a negligible impact on the calculation of thermal margin (Entergy letter dated April 14, 2009). The NRC staff concludes that, while the first assumption is non-conservative and XCOBRA-T has the capability to explicitly model direct cladding heat deposition, the analytical approach for RBS will yield essentially the same prediction of thermal margin and, therefore, the NRC staff concludes that the approach is reasonably accurate and acceptable.
The NRC staff reviewed the licensee's assumption regarding the generic power fractions. A similar sensitivity study was provided by the licensee that demonstrates the impact on the transient response when the generic or case-specific power fractions are used (Entergy letter dated April 14, 2009). The results of the sensitivity study confirm that the transient response is essentially the same when the case-specific or generic power fractions are used and, therefore, the NRC staff concludes that the approach is reasonably accurate and acceptable.
3.1.2.1.3 Radial Power Shape Assumptions XCOBRA-T is a one-dimensional thermal hydraulic code. Therefore, it lacks the capacity to update the radial power distribution during transient calculations. As a result, the NRC staff requested the licensee to evaluate whether the assumption of a constant radial power shape adversely affects the calculation of the transient behavior in terms of CPR or LHGR.
In its supplemental letter dated April 14, 2009, the licensee included various analyses to address transient radial power shape changes. The response included a steady-state perturbation analysis to determine the magnitude of the anticipated variation in local peaking factor over the range of void conditions anticipated in the limiting fuel nodes during limiting transients. The response states that the local radial power is expected to flatten by approximately 0.3 percent for an increase in void fraction on the order of 5 percent. This prediction is based on steady-state calculations performed using the three-dimensional core simulator (MICROBURN-B2) (Entergy letter dated April 14,2009). The core power flattening is expected for a void fraction increase due to enhanced neutron migration with reduced nodal slowing down power.
The licensee performed transient calculations to determine the magnitude of the transient void fraction change in the limiting axial nodes. The licensee stated that the void fraction transient changes are largely offset during the transient (Entergy letter dated April 14, 2009). The change in void fraction was found to be of a consistent magnitude and, generally, void fraction decreases that increase power were found to be limited by the reactor inherent feedback mechanisms (increases in power result in downstream increased void production) and, therefore, the NRC staff concurs with this conclusion.
- 9 The NRC staff notes that the radial power shape sensitivity was evaluated using steady-state nuclear methods. However, as the licensee correctly stated in its supplementalletter'dated April 14, 2009, the actual rod heat flux during the transient is a function of the heat hold-up in the fuel due to fuel rod thermal resistance. Therefore, the rod power peaking sensitivity is an over-estimate of the sensitivity of the transient radial heat flux distribution sensitivity. The licensee provided the timing of the point of MCPR and peak heat flux. These points occur very early in the transient (1.25 seconds and 0.8 seconds, respectively), while the fuel thermal time constant is typically on the order of 5 seconds (Entergy letter dated April 14,2009). On this basis, the NRC staff concludes that the transient radial power shape effect is negligible (0.3 percent based on steady-state calculations) and that during the transient calculation, such a power shift will not appreciably manifest as actual rod heat flux until after the limiting conditions are encountered in the analysis.
Based on the above, the NRC staff concludes that the analysis methodology is reasonably accurate and that the approach to analyzing the transient CPR and LHGR is acceptable in terms of the treatment of the local radial power shape.
3.1.2.1.4 Hot Rod Thermal Properties Assumptions Similar to COTRANSA2, XCOBRA-T allows user input for the hot rod fuel thermal properties (e.g., gap conductance). The NRC staff requested additional information regarding the licensee's input assumptions for COTRANSA2 regarding the selection of the input for these parameters and regarding the code input for XCOBRA-T.
The licensee provided a description of the methodology used to set the fuel rod thermal resistance to a conservatively low value. This is generally achieved by forcing the closure of the gas gap. When the exposure is low, the gap is not closed, but AREVA NP applies conservatism in these cases as described in the response to RAI 6.17 (Entergy letter dated April 14, 2009).
In its supplemental letter dated April 14,2009, the licensee discussed competing effects in the selection of the gas-gap properties. The NRC staff concludes that setting a low gas-gap conductance will enhance the predicted neutron power response by suppressing negative reactivity feedback from void formation during the pressurization. Likewise, the NRC staff concludes that setting a high gas-gap conductance for the hot rod for the thermal hydraulic analysis will enhance heat transfer through the rod and conservatively predict higher transient cladding heat flux during the transient calculation.
The licensee demonstrated that by applying a conservative bias to either the COTRANSA2 or XCOBRA-T gas gap results in the same sensitivity in the total transient analysis result. The NRC staff concludes that it is appropriate to calculate the system response on an average basis. The licensee stated that the gas-gap conductance is conservatively biased for the hot rod analysis. Therefore, based on the above, the NRC staff concludes that this integrated approach is appropriately conservative and is acceptable.
3.1.2.1.5 Single Fluid Approximation and Heat Flux Ratio Approach The NRC staff notes that an inherent limitation on the XCOBRA-T methodology is the single fluid approximation. XCOBRA-T constrains the fluid conditions such that the two phases on the
- 10 fluid remain in thermal equilibrium. This may impact the prediction of the transient heat flux during the calculation. For example, during an event such as a load reject - no bypass, the reactor pressure increases. The reactor pressure increase, as modeled in XCOBRA-T, results in the instantaneous temperature rise in the liquid phase to the saturation temperature for that pressure once the fluid is in bulk boiling. In actuality, the temperature will increase but the film fluid conditions may not reach saturation temperature at the higher pressure. This may result in erroneous prediction of the cladding heat flux because the temperature of the fluid spikes instantaneously with the traversal of the back pressure wave under these conditions. The NRC staff requested the licensee justify the applicability of this approximation, specifically noting that the prediction of the transient LHGR is derived from the predicted cladding heat flux.
In its supplemental letter dated April 14, 2009, the licensee justified the heat flux ratio (H FR) approach. The licensee stated that the transient LHGR is approximated based on the ratio of the peak predicted axial nodal heat flux to the steady-state heat flux. The transient LHGR is estimated as the product of the operating limit LHGR and the ratio of the peak to steady-state heat flux. The licensee stated that this is the approximate result that would be generated if the XCOBRA-T initial conditions were adjusted to result in a rod operating at the operating limit LHGR (Entergy letter dated April 14, 2009).
The licensee also discussed the adequacy of the methodology to predict the transient heat flux.
The licensee stated that for conditions below the point of critical heat flux (dry-out), the process of nucleate boiling is such an efficient heat removal process that the fuel rod thermal resistance is essentially driven entirely by the resistance of the cladding, gap, and pellet (Ente.rgy letter dated April 14, 2009).
The NRC staff concurs with the licensee that the contribution of the fluid heat transfer coefficient to the fuel thermal resistance is negligible compared to the other components for conditions below critical heat flux. As the heat flux prediction is insensitive to the evaluation of the specific fluid conditions below the point of dry-out, the NRC staff concludes that the single fluid approximation does not adversely impact the prediction of. the transient heat flux.
The NRC staff further considered the approximation of the peak transient LHGR in the first portion of the response. The licensee conjectured that the use of the peak HFR results in an equivalent estimation of the peak transient LHGR as would be predicted by setting the XCOBRA-T radial peaking factor at a point where the hot rod is initially at the operating limit LHGR. In the COTRANSA2/XCOBRA-T methodology, the rod powers are generated based on the one-dimensional plant response predicted by COTRANSA2. The total reactor power is apportioned to the bundles and rods in XCOBRA-T according to the radial peaking factors predicted by MICROBURN-B2 (References 8, 6, and 15). If XCOBRA-T were run in a manner where the bundle radial peaking factor was adjusted to set the hot rod to the operating limit LHGR, the impact on the analysis would be to perturb the initial thermal hydraulic conditions in the hot bundle. However, the relative fission power is still an analysis input from the upstream COTRANSA2 calculatioll. Since the heat flux calculation is relatively insensitive to the fluid conditions when margin to dry-out is demonstrated, the relative peak heat flux is insensitive to the initial radial peaking factor for the hot bundle and, therefore, the NRC staff concludes that the approach to utilize the peak HFR is reasonably accurate and acceptable.
- 11 3.1.2.1.6 Bounds Checking The XCOBRA-T code includes bounds checking techniques. The purpose of the bounds checking is to ensure that the code does not allow the determination of the CPR outside the bounds of the qualification of the critical heat flux correlation. The critical heat flux correlation and its appropriate bounds are described in Reference 16. The NRC staff requested additional information regarding the execution of XCOBRA-T for the purpose of calculating the transient LHGR. In particular, the NRC staff is aware that conservative model substitutions may occur in the code if certain bounds are exceeded. As these bounds may be conservative for the evaluation of the transient CPR, the NRC staff was concerned that these model substitutions may be non-conservative for transient LHGR evaluation. The NRC staff requested additional information in regarding the model substitution and bounds checking approach.
In its supplemental letter dated April 14, 2009, the licensee provided the actions taken when the bounds of the critical heat flux correlation are exceeded. These actions are provided in Table 6.19-1 of the licensee's letter. The bounds that are evaluated are the mass flow rate, nodal enthalpy, pressure, and inlet subcooling. The response additionally states that negative flows force the code to stop (Entergy letter dated April 14, 2009). These actions are consistent with the NRC staff's conditions and limitations on the approval of XCOBRA-T and SPCB, as documented in References 6 and 16, respectively.
The fluid properties contribute minimally to the prediction of the actual heat flux. The heat flux is a function of the total thermal resistance between the fuel pellet and the coolant. Under conditions when the fuel rod is not in transition boiling, the heat transfer coefficient between the cladding and the coolant is driven by thermal conduction and convection for the subcooled regime and nucleate boiling for the limiting portions of the bundle. As a result, the contribution to the total thermal resistance from heat transfer to the coolant has a negligible impact on the prediction of the cladding surface heat flux. Therefore, the NRC staff concludes that any treatment of the fluid conditions driven by the bounds checking approach will have a negligible impact on the XCOBRA-T calculation of the transient heat flux.
The licensee stated that the transient LHGR is derived from the predicted transient heat flux (Entergy letter dated April 14, 2009). While it is the expectation that conformance with the boiling transition criterion ensures that the evaluation conditions are within the critical heat flux correlation bounds, even under conditions when these bounds are exceeded, the NRC staff agrees that the calculation of the LHGR will be impacted in a negligible manner and, therefore, the NRC staff concludes that the methodology is reasonably accurate and the approach is acceptable.
3.1.3 LHGR Limits LHGR limits are established to ensure that the one percent plastic strain and fuel centerline melt limits are met during normal operation including the effects of AOOs. The NRC staff requested that the licensee specify the codes that are used to determine these limits. The licensee stated that the approved RODEX2A and RAMPEX methods were used (Entergy letter dated April 14, 2009). The NRC staff requested information regarding the use of these codes to evaluate the fuel thermal mechanical criteria, specifically, that the licensee provide the input assumptions regarding the power history. The licensee stated that the approved RODEX2A, RAMPEX, and
- 12 COLAPX codes were used consistent with the previously approved method to convert the LHGR operating limit into power history input (Entergy letter dated April 14, 2009). This suite of methods was approved by the NRC staff in References 1, 2, 3, and 4. The NRC staff requested that the licensee to justify the power shapes used in the analysis. The licensee stated that the power shapes used are conservative and that they are consistent with the shapes approved in Reference 5 on the same basis.
The NRC staff reviewed these responses and found that the methods used to determin~ the fuel thermal mechanical criteria have been approved by the NRC. The NRC staff further confirmed that inputs to the methods including the power history and power shape are conservative based on findings of previous reviews as documented in References 2, 4, and 5. The NRC staff concludes that the RBS analyses were performed using approved methods and that these methods were executed using acceptable and conservative analysis inputs and, therefore, the NRC staff concludes that the licensee's analysis approach is acceptable.
3.2 TS Changes The current RBS LCO 3.7.5 states that "The Main Turbine Bypass System shall be OPERABLE." This requirement is applicable when the power level is greater than or equal to 23.8 percent of the rated thermal power (RTP).
The proposed changes would rename the current LCO as LCO A and add a new LCO B. The new LCO B contains additional requirements that would allow the alternative option when the MTBS is inoperable as specified in the COLR, and will read as follows:
OR B. The follOWing limits for inoperable Main Turbine Bypass System, as specified in the COLR are made applicable:
Current REQUIRED ACTION A of the LCO requires restoration of the MTBS to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> whenever the MTBS is inoperable. The proposed TS would rename the current REQUIRED ACTION A as REQUIRED ACTION A.1 and add a new REQUIRED ACTION A.2. The new REQUIRED ACTION A.1 would require the MTBS be returned to OPERABLE or implement REQUIRED ACTION A.2 to apply the limits as specified in the COLR, and will read as follows:
A.1 Restore the Main Turbine Bypass System to OPERABLE status OR
- 13 A.2 Apply the APLHGR, LHGR and MCPR limits for inoperable Main Turbine Bypass System as specified in the COLR.
The COMPLETION TIME for ACTION A.1 or ACTION A.2 is within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of whenever the MTBS is declared inoperable.
The current ACTION B remains unchanged. It would require a reduction in thermal power to less than 23.8 percent of the RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> whenever the MTBS is inoperable.
3.3 Transients and Accidents Analysis In support of the proposed TS changes, the licensee performed an analysis to identify the effects of the operation with an inoperable MTBS in the proposed TS on the operating limits of the MCPR, LHGR, and APLHGR. During the process of the analysis, the licensee screened all the events discussed in Chapter 15 of the Updated Safety Analysis Report (USAR) and classified (Entergy letter dated April 14,2009, Table 5.2) the events into the following three categories:
- 1. events that are not required to be analyzed during fuel reloads because they are identified in the analysis of record (AOR) as non-limiting events, events that are bounded by other events in the same event category, or events that are not credible to RBS,
- 2. events that are required to be addressed for each reload because they are identified in the AOR as the limiting events in the applicable event categories, and
- 3. new events or conditions that are identified from the events in the above Category 2 for inoperable MTBS to be addressed for each cycle.
The licensee only identified two events for the above screening and both are in Category 3:
(1) feedwater controller failure (FWCF) and (2) recirculation flow controller failure - slow opening (the slow core flow runout). For both of the events, the licensee performs analyses assuming an inoperable MTBS to establish the MCPR and linear heat generation rate factor (LHGRFAC) limits to avoid fuel damage during transients.
The FWCF event is analyzed using (Entergy letter dated April 14, 2009, RAI 1)the same methods and computer codes as the base FWCF event in the AOR with the exception that the MTBS being inoperable is assumed. The computer codes used are the NRC-approved COTRANSA2 (Reference 8) and XCOBRA-T (Reference 6). The results of the analysis indicate (Entergy letter dated April 14, 2009, RAJ 2) that for Cycle 15, the effects of the MTBS out-of service (MTBS-OOS) on calculated .1MCPR and HFR for the FWCF event are small and in the range of 0.00 and 0.02 depending upon the core exposure and initial core thermal power. At higher exposures, the effects on the power-dependent MCPR (MCPR p) are also small and are in the range of 0.01 to 0.02 for initial core power levels above 70 percent. Below 70 percent, other events, which are not affected by the MTBS-OOS, set the operating MCPR p limits.
Operation with the MTBS-OOS results in a greater transient HFR for the FWCF event.
However, when operations are combined with other equipment out-of-service (EOOS) options,
- 14 the required power-dependent LHGRFAC (LHGRFAC p) is either limited by the steady-state HFR limit or the required LHGRFACpfrom another EOOS condition.
The slow core flow runout event is analyzed for determination of flow-dependent thermal limits.
The event could result in vessel steam flow exceeding the capacity of the turbine control valves, and thus the turbine bypass valves would open to control reactor pressure. With the MTBS-OOS, the reactor pressure would increase relative to the nominal slow flow runout analysis. The slow core flow runout event with the MTBS-OOS is analyzed using the same methods and computer codes as the base slow core flow runout event in the AOR with the exception that the inoperable MTBS is assumed. The computer codes used for the analysis are the NRC-approved COTRANSA2 (Reference 8) for determination of the flow-dependent MCPR (MCPRF) limits and COSMO-4/MICROBURN-B2 (Reference 10) for determination of the flow dependent heat generation rate factor (LHGRFACF). The slow core flow runout event was analyzed for Cycle 15 with and without the MTBS-OOS. The results of the analysis show that a maximum increase in the MCPRF is 0.10. The Cycle 15 LHGRFACF limits were established to support base case operation and operation in the EOOS options for all cycle exposures.
APLHGR is determined by averaging the LHGR over each fuel rod in a plane. The limit for APLHGR is expressed as the maximum APLHGR (MAPLHGR) for any plane in the fuel assembly. The MAPLHGR is determined to meet the 10 CFR 50.46 criteria that require the peak cladding temperature for design-basis LOCAs not exceed 2200 degrees Fahrenheit (OF).
An inoperable MTBS does not result in an increase in severity of results associated with the LOCA analyses (Entergy letter dated April 14,2009, RAI 4); therefore, the MAPLHGR limits remain unchanged for an inoperable MTBS.
The licensee will perform the cycle-specific analysis (Entergy letter dated April 14, 2009, RAI 3) to address the events discussed above as an EOOS option. Appendix A of the current RBS COLR contains the EOOS thermal limits curves. The licensee will modify it to include the MCPRp* MCPRF. LHGRFAC p, and LHGRFACFcurves for the inoperable MTBS cases.
The NRC staff finds that that the licensee's analysis uses NRC-approved methodologies and computer codes, and the results of the analysis show that the safety thermal limits in the AOR are not exceeded for the events that are affected by the MTBS-OOS, thus meeting the GDC 10 requirements related to the thermal limits for maintaining integrity of the fuel rods during AOOs.
Therefore, the NRC staff concludes that the analysis is acceptable.
3.4 Compliance with Guidance in GL 88-16 and NUREG-1434 NRC GL 88-16 allows licensees to move cycle-specific parameters from the plant-specific TSs to a licensee-controlled document entitled "COLR." The GL states that for plants implementing a COLR process, the analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC and shall be included in its TSs. The .
proposed TS 3.7.5 allows relocation of the MCPR, APLHGR, and LHGR operating limits to the COLR. The NRC staff agrees with the licensee that the operating limits referenced in TS 3.7.5 to be included in the COLR are cycle-specific parameters. Also, the licensee will use (Reference 17) the NRC-approved methods referenced in TS 5.6.5, "Core Operating Limits Report," in analyses to determine cycle-specific parameters including the MCPR, APLHGR, and LHGR operating limits, for each cycle.
- 15 In addition, the licensee's commitments (discussed above in Section 3.4) assure that it will adjust the MCPR, APLHGR, and LHGR operating limits to support operation with the MTBS being inoperable, and incorporate the adjustments into the COLR.
Therefore, the NRC staff concludes that the proposed TS changes are in compliance with the GL 88-16 guidance and is, therefore, acceptable.
NUREG-1434, Revision 3, provides a TS corresponding to the RBS TS 3.7.5 in Standard TS 3.7.6, "Main Turbine Bypass System." The standard TS 3.7.6 states that "[t]he Main Turbine Bypass System shall be OPERABLE." For inoperable MTBS, the following limits are made applicable:
[a. LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," limits for an inoperable Main Turbine Bypass System, as specified in the [COLR] and]
[b. LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," limits for an inoperable Main Turbine Bypass System, as specified in the [COLR].]
The requirements are applicable when the power level is greater than or equal to 25 percent of the RTP.
REQUIRED ACTION A.1 would require the MTBS be returned to OPERABLE or to apply the limits as specified in the COLR within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> whenever the MTBS is inoperable. REQUIRED ACTION A.1 is stated as follows:
A.1 [Satisfy the requirements of the LCO or Restore the Main Turbine Bypass System to OPERABLE status.]
The requirements in the brackets are plant-specific related requirements. The COMPLETION TIME is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
REQUIRED ACTION B.1 would require a reduction in thermal power to less than 25 percent of the RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> whenever the MTBS is inoperable. REQUIRED ACTION B.1 is stated as follows:
B.1 [Satisfy the requirements of the LCO or Reduce THERMAL POWER to <
23.8% RTP.]
The requirements in the brackets are plant-specific related requirements. The COMPLETION TIME is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
The NRC staff comparison of the RBS TS 3.7.5 (discussed above in Section 3.2) with the equivalent standard TS 3.7.6 and concluded that:
(1) Both TSs allow operation with an inoperable MTBS;
- 16 (2) For both TSs for operation with the MTBS being inoperable, the operating thermal limits are required to be included in COLRs. RBS LCO 3.7.5 contains operating limits for three parameters (APLHGR, MCPR, and LHGR) while the equivalent standard LCO 3.7.6 contains limits for two parameters (APLHGR and MCPR);
(3) The applicable power levels are similar (23.8 percent versus 25 percent RTP that is supported by the plant-specific transient and accident analysis); and (4) REQUIRED ACTION statements are similar in the required actions and completion times.
Therefore, the NRC staff concludes that the RBS TS 3.7.5 meets the intent of standard TS 3.7.6 in NUREG-1434 and is, therefore, acceptable.
3.5 Conclusions The NRC staff confirmed that RBS is referencing the approved AREVA NP methods for performing the relevant safety analyses at RBS. The NRC staffs review confirmed that the methods were employed consistent with the NRC staff's conditions, limitations, and restrictions as specified in the approving safety evaluations for these methods. Additionally, the NRC staff reviewed clarifying information provided by the licensee to ensure that the code inputs and usage options are consistent with the required degree of accuracy and are appropriately conservative to account for effects such as operational flexibility, plant design, and fuel .
performance. Therefore, the NRC staff concludes that: (1) the licensee's analysis uses NRC approved methodologies and computer codes, consistent with the NRC staff's conditions, limitations, and restrictions, referenced in TS 5.6.5, to determine the operating thermal limits specified in TS 3.7.5; (2) the licensee's commitments listed in Section 4.0 of this safety evaluation assures that the safety thermal limits in the AOR will not be exceeded for the events that are affected by the MTBS-OOS; (3) the additional restrictions imposed in the TS 3.7.5 by the revised set of operating limits will adequately offset the impact of losing the MTBS function; and (4) the proposed TS changes are in compliance with the GL 88-16 guidance and meet the intent of the standard TS 3.7.6 in NUREG-1434, Revision 3. The NRC staff concludes that the proposed TS 3.7.5 provide adequate requirements to protect fuel rods from failure in meeting the GDC 10 requirement regarding SAFDLs, and satisfy the 10 CFR 50.36 requirements regarding LCOs and ACTIONS in assuring safe operation of the RBS and, therefore, the proposed TS 3.7.5 is, therefore, acceptable.
4.0 REGULATORY COMMITMENTS The cycle-specific fuel operating limits based on the design function of the MTBS are currently documented in the RBS COLR, a licensee-controlled document. Operating limits for the entire cycle are based on an operable MTBS. The proposed changes will allow the COLR to contain operating limits that are applicable when the MTBS function is not operable. In support of the proposed TS 3.7.5 application, the licensee provided, in Attachment 4 to Reference 17, the following regulatory commitments:
- 17
- 1. The Bases will be revised to confirm that the alternative MCPR, APLHGR, and LHGR operating limits are sufficient to mitigate pressurization transient effects and that the alternative limits restore the margin to the APLHGR, MCPR, and LHGR assumed in the safety analysis.
- 2. The adjustments will be incorporated into the COLR and into the core monitoring software in the core monitoring computer.
In the Entergy letter dated January 14, 2008, the licensee committed to complete the above commitments within 60 days of NRC staff approval of the proposed TS 3.7.5 in Reference 17.
The licensee's commitments listed above assure that RBS will appropriately revise the Bases to TS 3.7.5 to reflect the additional restrictions imposed by the revised set of operating limits to adequately offset the impact of losing the MTBS function.
5.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Louisiana State official was notified of the proposed issuance of the amendment. The State official had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. 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, and there has been no public comment on such finding published in the Federal Register on March 11, 2008 (73 FR 13023). 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.
7.0 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) such activities will be conducted in compliance with the Commission's 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.
8.0 REFERENCES
- 1. XN-NF-85-67(P)(A) Revision 1, "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel," Exxon Nuclear Company, September 1986.
- 18
- 2. EMF-85-74(P) Revision 0 Supplement 1(P)(A) and Supplement 2(P)(A), "RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model," Siemens Power Corporation, February 1998.
- 3. XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, "RODEX2 Fuel Rod Thermal Mechanical Response Evaluation Model," Exxon Nuclear Company, March 1984.
- 4. ANF-89-98(P)(A) Revision 1 and Supplement 1, "Generic Mechanical Design Criteria for BWR Fuel Designs," Advanced Nuclear Fuels Corporation, May 1995.
- 5. BAW-1 0247(P)(A) Revision 0, "Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors," AREVA NP, April 2008 (ADAMS Accession No.ML081340383)
- 6. XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," Exxon Nuclear Company, February 1987.
- 7. R.A. Copeland, Siemens, to R.C. Jones, U.S. Nuclear Regulatory Commission, LeUer, "ATRIUM-10 Presentations," RAC:95:080, May 4, 1995 (38-9091703-000).
- 8. ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3 and 4, "COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses,"
Advanced Nuclear Fuels Corporation, August 1990.
- 9. ANF-524(P)(A) Revision 2 and Supplements 1 and 2, "ANF Critical Power Methodology for Boiling Water Reactors," Advanced Nuclear Fuels Corporation, November 1990.
1O. EMF-2158(P)(A), Revision 0, "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," Siemens Power Corporation, October 1999.
- 11. XN-NF-80-19(P)(A) Volume 3, Revision 2, "Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Thermal Limits Methodology Summary Description," Exxon Nuclear Company, January 1987 (ADAMS Accession No. ML081340305).
- 12. XN-NF-80-19(P)(A), Volume 1, "Exxon Nuclear Methodology for Boiling Water Reactors Neutronic Methods for Design and Analysis," Exxon Nuclear Company, March 1983 (ADAMS Accession No. ML081850204).
- 13. XN-NF-80-19(P)(A), Volume 4, Revision 1, "Exxon Nuclear Methodology Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," Exxon Nuclear Company, June 1986 (ADAMS Accession No. ML081700491).
- 14. Stuart Richards, U.S. Nuclear Regulatory Commission, to James F. Mallay, Siemens Power Corporation, "Siemens Power Corporation Re: Request for Concurrence on
- 19 Safety Evaluation Report Clarifications (TAC No. MA6160)," dated May 31,2000 (ADAMS Accession No. ML003719373).
- 15. Jerry Roberts, Entergy Operations, Inc. to U.S. Nuclear Regulatory Commission, "License Amendment Request, Main Turbine Bypass System, River Bend Station, Unit 1, Docket No.40-458, License No. NPF-47," dated April 14, 2009 (ADAMS Accession No. ML091110056)
- 16. EMF-2209(P)(A) Revision 1, "SPCB Critical Power Correlation," Siemens Power Corporation, July 2000.
- 17. J. C. Roberts, Entergy Operations, Inc. to U.S. Nuclear Regulatory Commission, "License Amendment Request, Main Turbine Bypass System, River Bend Station, Unit 1, Docket No. 50-458, License No. NPF-47," dated January 25,2008 (ADAMS Accession No. ML080440293).
- 18. J. C. Roberts, Entergy Operations, Inc. to U.S. Nuclear Regulatory Commission, "License Amendment Request, Main Turbine Bypass System, River Bend Station, Unit 1, Docket No. 50-458, License No. NPF-47," dated April 29, 2009 (ADAMS Accession No. ML091260596).
- 19. U.S. Nuclear Regulatory Commission, "Standard Technical Specifications General Electric Plants, BWR/6," NUREG-1434, Revision 3, dated June 2004 (ADAMS Accession No. ML041910204).
- 20. U.S. Nuclear Regulatory Commission, "Removal of Cycle-Specific Parameter Limits From Technical Specifications," Generic Letter 88-16, dated October 3, 1988.
Principal Contributors: P. Yarsky S.Sun Date: June 24 I 2009
June 24, 2009 Vice President, Operations Entergy Operations, Inc.
River Bend Station 5485 US Highway 61 N St. Francisville, LA 70775 SUB~IECT: RIVER BEND STATION, UNIT 1 - ISSUANCE OF AMENDMENT RE: MAIN TURBINE BYPASS SYSTEM (TAC NO. MD7966)
Dear Sir or Madam:
The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 163 to Facility Operating License No. NPF-47 for the River Bend Station, Unit 1. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated January 25, 2008, as supplemented by letters dated April 14 and 29, 2009.
The amendment revises TS 3.7.5, "Main Turbine Bypass System." The change provides an alternative to the existing Limiting Condition for Operation for the Main Turbine Bypass System (MTBS). The revised TS will require that the MTBS be operable or that the Average Planar Linear Heat Generation Rate, the Minimum Critical Power Ratio, and the Linear Heat Generation Rate limits for the inoperable MTBS be placed in effect as specified in the Core Operating Limits Report.
A copy of our related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
Sincerely,
/RN Alan B. Wang, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-458
Enclosures:
- 1. Amendment No. 163 to NPF-47
- 2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:
PUBLIC RidsNrrDorlLpl4 Resource RidsOgcRp Resource LPLIV r/f RidsNrrDssSrxb Resource RidsRgn4MailCenter Resource RidsAcrsAcnw_MailCTR Resource RidsNrrDssSnbp Resource S. Sun, NRRlDSS/SRXB RidsNrrDirsltsb Resource RidsNrrLAJBurkhardt Resource PYarsky, NRRIDSS/SNPB RidsNrrDorlDpr Resource RidsNrrPMRiverBend Resource ADAMS Accession No ML091540570 *SE memo dated OFFICE NRR/LPL4/PM NRR/LPL4/LA DJRSIITSB/BC DSS/SRXB/BC DSS/SNPB/BC OGC NRR/LPL4/BC NRR/LPL4/PM NAME ABWang JBurkhardt RElliott GCranston AMendiola MSpencer MMarkley ABWang DATE 6/10109 6/8/09 6/18/09 5/21/09 4/20/2009 6/15/09 6/24/09 6/24/09 OFFICIAL AGENCY RECORD