ML030850493
| ML030850493 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee  |
| Issue date: | 03/31/2003 |
| From: | Dick M, Schrull E General Electric Co |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| DRF 0000-0007-1043 NEDO-33089 | |
| Download: ML030850493 (75) | |
Text
Docket No. 50-271 BVY 03-23 Attachment 6 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 257 APRM/RBM/Technical Specifications /
Maximum Extended Load Line Limit Analysis (ARTS/MELLLA)
Non Proprietary Version - for Public Disclosure
GE Nuclear Energy NEDO-33089 Class I DRF 0000-0007-1043 March 2003 Vermont Yankee Nuclear Power Station APRM/RBM/Technical Specifications I Maximum Extended Load Line Limit Analysis (ARTS/MELLLA)
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NEDO-33089 Class I DRF 0000-0007-1043 March 2003 Vermont Yankee Nuclear Power Station APRM/RBM/Technical Specifications/
Maximum Extended Load Line Limit Analysis (ARTS/MELLLA)
Prepared by: E. D. Schrull Approv al__ __ __ _
M. J.Dic, Project Manager BWR Asset Enhancement Services
NEDO-33089 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The only undertakings of the General Electric Company (GE) respecting information in this document are contained in the contract between Entergy Nuclear Operations, Inc. (ENOI) and GE, Contract Order No. VY014791, effective September 25, 2002, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than ENOI, or for any purpose other than that for which it is intended, is not authorized; and, with respect to any unauthorized use, GE makes no representation or warranty, express or implied, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.
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NEDO-33089 TABLE OF CONTENTS 1.0 INTROD UCTION ............................................................................................................ 1-1 1.1 Background ................................................................................................................... 1-2 1.2 A RTS/M ELLLA Bases ................................................................................................ 1-2 1.2.1 Analytical Bases .................................................................................................... 1-2 1.2.2 APRM High Flux (Flow Bias) Scram and Rod Block Design Bases .................... 1-4 1.3 A PRM Im provem ents ................................................................................................... 1-5 2.0 O VERA LL AN ALY SIS APPROACH ............................................................................. 2-1 3.0 FUEL THERM A L LIM ITS .............................................................................................. 3-1 3.1 Lim iting Core-W ide A O O Analyses ............................................................................ 3-1 3.1.1 Elimination of APRM Trip Setdown and DTPF Requirement .............................. 3-2 3.2 Input A ssum ptions ........................................................................................................ 3-3 3.3 Analyses Results ........................................................................................................... 3-4 3.3.1 Power-Dependent M CPR Lim it ............................................................................. 3-4 3.3.2 Power-Dependent LHG R and M APLHGR Lim its ................................................ 3-5 3.3.3 Flow -D ependent M CPR Lim it ............................................................................... 3-6 3.3.4 Flow -D ependent LHG R and M APLHG R Lim its .................................................. 3-6 3.3.5 Safety Lim it M CPR A djustm ent Procedure .......................................................... 3-7 3.3.6 Single Loop Operation A djustm ent Procedure ...................................................... 3-7 3.4 Rod W ithdraw al Error A nalysis ................................................................................... 3-7 3.5 Conclusion .................................................................................................................... 3-8 4.0 VESSEL O VERPRESSURE PROTECTION .................................................................. 4-1 5.0 TH ERM AL-HYD RAULIC STA BILITY ......................................................................... 5-1 5.1 Stability Option I-D ...................................................................................................... 5-1 6.0 LO SS-O F-CO OLANT A CCIDENT AN A LY SIS ............................................................ 6-1 6.1 Conclusions .................................................................................................................. 6-2 7.0 CON TA INM EN T RESPON SE ........................................................................................ 7-1 8.0 REACTO R INTERN A LS INTEG RITY .......................................................................... 8-1 8.1 Reactor Internal Pressure D ifferences .......................................................................... 8-1 iv
NEDO-33089 8.2 A coustic and Flow-Induced Loads ............................................................................... 8-1 8.2.1 Approach/M ethodology ......................................................................................... 8-1 8.2.2 Input A ssum ptions ................................................................................................. 8-2 8.2.3 Results .................................................................................................................... 8-2 8.3 Structural Integrity Evaluation ..................................................................................... 8-2 8.3.1 Conclusion ............................................................................................................. 8-4 8.4 Reactor Internals V ibration .......................................................................................... 8-4 8.4.1 Approach/ M ethodology ........................................................................................ 8-4 8.4.2 Inputs/A ssumptions ............................................................................................... 8-4 8.4.3 Analyses Results .................................................................................................... 8-5 8.4.4 Conclusion ............................................................................................................. 8-6 9.0 ANTICIPATED TRANSIENT WITHOUT SCRAM .................................................. 9-1 9.1 Approach/M ethodology ................................................................................................ 9-1 9.2 Input Assum ptions ........................................................................................................ 9-2 9.3 Analyses Results ........................................................................................................... 9-3 9.4 Conclusions .................................................................................................................. 9-3 10.0 STEAM DRYER AND SEPARATOR PERFORMANCE ............................................ 10-1 11.0 TESTIN G ........................................................................................................................ 11-1 12.0 REFEREN CES ............................................................................................................... 12-1 V
NEDO-33089 LIST OF TABLES Table Title 1-1 Computer Codes Used for ARTS/MELLLA Analyses 2-1 Analyses Presented in this Report 2-2 Applicability of Analyses 3-la Base Conditions for ARTS/MELLLA Rated Transient Analyses 3-lb Base Conditions for ARTS/MELLLA Off-rated Transient Analyses 3-2 MELLLA Transient Analyses Peak Values, Cycle 23 ARTS Transient Analysis Results - Generic K(P) Confirmation Above P Bypass 3-4 ARTS Transient Analysis Results - MCPR(P) Below P-Bypass ARTS Transient Analysis Results - Generic MAPFAC(P) Confirmation Above P-Bypass 3-6 ARTS Transient Analysis Results - MAPFAC(P) Below P-Bypass 3-7 Summary of Unblocked OLMCPR Values for the RWE Event ARTS Generic Design RFI Rod Line for MCPR(F) Determination Compared to VYNPS RFI Results 3-9 ARTS Transient Analysis Results - MAPFAC(F) 4-1 VYNPS Cycle 23 Sensitivity of Overpressure Analysis Results to Initial Flow 6-1 DBA LOCA Initial Conditions for VYNPS ARTS/MELLLA 6-2 DBA LOCA Results Comparison for VYNPS ARTS/MELLLA 8-1 Flow-induced Loads on Shroud and Jet Pumps for VYNPS 8-2 Maximum Acoustic Loads on Shroud and Jet Pumps Maximum Acoustic Loads on Shroud Support (MELLLA condition, 64%P /
36%F) 8-4 Summary of Structural Evaluation for ARTS/MELLLA Operating Conditions and Equipment Performance Characteristics for ATWS 9-1 Analyses 9-2 Summary of Key Parameters for Short-term ATWS Calculation 9-3 Summary of Key Parameters for Long-term ATWS Calculation 9-4 ATWS Calculation Sequence of Events (Time, in seconds) vi
NEDO-33089 LIST OF FIGURES Figure Title 1-1 ARTS/MELLLA Power/Flow Map 3-1 Power-Dependent MCPR Limits, MCPR(P) 3-2 Power-Dependent LHGR or MAPLHGR Multiplier, LHGRFAC(P) or MAPFAC(P) 3-3 Flow-Dependent MCPR Limits, MCPR(F) 3-4 Flow-Dependent LHGR or MAPLHGR Multiplier, LHGRFAC(F) or MAPFAC(F) 5-1 Option I-D APRM Flow-Biased High Flux Scram Line (MELLLA versus ELLLA Operation) vii
NEDO-33089 ACRONYMS Term Definition AW Difference between two loop and single loop cffectivc drive flow at the same core flow AL Analytical Limit AOO Anticipated Operational Occurrence APRM Average Power Range Monitor ARI Alternate Rod Insertion ARTS APRM/RBM/Technical Specifications ATWS Anticipated Transient Without Scram BOC Beginning-of-Cycle BT Boiling Transition BWR Boiling Water Reactor CLTP Current Licensed Thermal Power COLR Core Operating Limits Report CPR Critical Power Ratio CRGT Control Rod Guide Tube DBA Design Basis Accident DIVOM Delta CPR over Initial MCPR Versus the Oscillation Magnitude DTPF Design Total Peaking Factor ECCS Emergency Core Cooling System ELLLA Extended Load Line Limit Analysis ENOI Entergy Nuclear Operations, Inc.
Entergy Entergy Nuclear Northeast EOC End-of-Cycle FCL Flow Control Line FCTR Flow Control Trip Reference FIV Flow-Induced Vibration RFI Recirculation Flow Increase FWCF Feedwater Controller Failure GE General Electric viii
NEDO-33089 Term Definition HCOM Hot Channel Oscillation Magnitude ICA Interim Corrective Action ICF Increased Core Flow ICGT Incore Guide Tube IORV Inadvertent Opening ofa Relief Valve IRLS Idle Recirculation Loop Start-up JPSL Jet Pump Sensing Line LCO Limiting Condition for Operation LCRP Limiting Control Rod Pattern LFWH Loss of Feedwater Heating LHGR Linear Heat Generation Rate LHGRFAC LHGR Multiplier LOCA Loss-Of-Coolant Accident LOOP Loss Of Offsite Power LPRM Local Power Range Monitor LRNBP Load Rejection with No Bypass MAPFAC MAPLHGR multiplier MAPLHGR Maximum Average Planar Linear Heat Generation Rate MCHFR Minimum Critical Heat Flux Ratio MCPR Minimum Critical Power Ratio MELLLA Maximum Extended Load Line Limit Analysis MNGP Monticello Nuclear Generating Plant MOP Mechanical Over-Power MSIV Main Steam Line Isolation Valve MSIVC Main Steam Line Isolation Valve Closure MSIVF Main Steamline Isolation Valve Closure with a Flux Scram MTPF Maximum Total Peaking Factor N/C Not Calculated N/R Not Reported NBP No Bypass ix
NEDO-33089 Term Definition NFI New Fuel Introduction NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System OLMCPR Operating Limit Minimum Critical Power Ratio OLTP Original Licensed Thermal Power OOS Out-of-Service PCT Peak Cladding Temperature PRFO Pressure Regulator Failure Open RBM Rod Block Monitor RCF Rated Core Flow RFI Recirculation Flow Increase RIPD Reactor Internal Pressure Difference RPT Recirculation Pump Trip RPV Reactor Pressure Vessel RSLB Recirculation Suction Line Break RWE Rod Withdrawal Error SER Safety Evaluation Report SLCS Standby Liquid Control System SLMCPR Safety Limit Minimum Critical Power Ratio SLO Single Loop Operation SR Surveillance Requirement SRV Safety-Relief Valve SRVOOS Safety-Relief Valve Out of Service SSV Spring Safety Valve TOP Thermal Over-Power TRM Technical Requirements Manual TS Technical Specification TTNBP Turbine Trip with No Bypass UFSAR Updated Final Safety Analysis Report VPF Vane Passing Frequency x
NEDO-33089 Term Definition VYNPS Vermont Yankee Nuclear Power Station Wc, % Rated Core Flow W % Recirculation Drive Flow xi
1.0 INTRODUCTION
Many factors restrict the flexibility of a Boiling Water Reactor (BWR) during power ascension from the low-power/low-core flow condition to the high-power/high-core flow condition. Once rated power is achieved, periodic adjustments must also be made to compensate for reactivity changes due to xenon effects and fuel burnup. Some of the factors currently existing at the Vermont Yankee Nuclear Power Station (VYNPS) that restrict plant flexibility in quickly achieving rated power are:
- 1. The currently licensed allowable operating power/flow map; and
- 2. The Average Power Range Monitor (APRM) flow-biased flux scram and flow-biased rod block setdown requirements.
Maximum Extended Load Line Limit Analyses (MELLLA) corresponds to plant operation above the current licensed VYNPS Extended Load Line Limit Analysis (ELLLA) boundary.
The current APRM and Rod Block Monitor (RBM) flow-biased rod block trips restrict the power ascension capability of certain BWRs, such as VYNPS. These operating restrictions are further compounded by the existing setdown requirements for these trips. The operating restrictions resulting from the existing APRM and RBM systems can be significantly relaxed or eliminated by the implementation of a series of APRM/RBM/Technical Specifications (ARTS) improvements. These improvements increase plant operating efficiency by updating the thermal limits administration. For the VYNPS application, the ARTS program will not include the modification of the RBM system from a flow-dependent to a power-dependent system.
Therefore, the existing flow-dependent RBM system setpoint is relaxed so that the potential for RBM interference when operating in the MELLLA region can be avoided or minimized. The operating flexibility associated with the ARTS activities complement those of the MELLLA mode of operation. The improvements associated with ARTS, along with the objectives attained by each improvement, are as follows:
- 1. A power-dependent Minimum Critical Power Ratio (MCPR) thermal limit similar to that used by BWR6 plants is implemented as an update to reactor thermal limits administration.
- 2. The APRM trip setdown and design total peaking factor are replaced by more direct power-dependent and flow-dependent thermal limits to reduce the need for manual setpoint adjustments and to allow more direct thermal limits administration. This improves human/machine interface, updates thermal limits administration, increases reliability, and provides more direct protection of plant safety.
- 3. The Rod Withdrawal Error (RWE) analysis was performed assuming no credit for the RBM control rod block functions. The new RBM setpoints will be based on providing operational flexibility in the MELLLA region.
This report presents the results of the safety analyses and system response evaluations performed for operation of VYNPS in the region above the rated rod line for a representative core of GEl 3 1-1
NEDO-33089 and GEl4 fuel-types (Cycle 23 core design). The current operating envelope is modified to include the extended operating region bounded by the rod line which passes through the 100% of current licensed thermal power (CLTP) / 75% of rated core flow (RCF) point, the rated power line, and the rated load line, as shown in Figure 1-1. Plant operational boundaries as shown in Figure 1-1 that are beyond the current licensed allowable operating power/flow map are referred to as the MELLLA region. Operation in the MELLLA region is intended to enhance the plant operational flexibility and increase plant capacity factor.
1.1 Background The power/flow operating map (Figure 1-1) includes the operating domain changes for ARTS/MELLLA consistent with approved operating domain improvements for other BWRs.
This performance improvement application expands the operating domain along the 120.8% rod line to 100% of CLTP at 75% of RCF. This operating domain is defined by the following boundary and basis:
" The MELLLA boundary line, extended up to the existing maximum CLTP of 1593 MWt.
The MELLLA boundary is defined as the line that passes through the 100% of CLTP /
75% of RCF state point.
" The CLTP of 1593 MWt, which is the same as the original licensed thermal power (OLTP).
" The currently analyzed Increased Core Flow (ICF) condition of 107.0% of RCF.
The MELLLA boundary line defines an increase in the extent of the current operating domain above the current boundary. The current boundary is the ELLLA, corresponding to the 108%
APRM Rod Block setpoint, and allows operation to approximately the 108% of CLTP rod line shown in Figure 1-1.
The currently analyzed power/flow point for Single Loop Operation (SLO), remains unchanged from its current absolute value of 1239 MWt for MELLLA. SLO is not extended into the MELLLA region.
1.2 ARTS/MELLLA Bases 1.2.1 Analytical Bases A modified power/flow curve has been derived to provide relief from the operating restrictions inherently imposed during ascension to power by the existing power/flow curve. [
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The ARTS/MELLLA application is determined on a plant-specific basis via a safety and impact evaluation for meeting thermal and reactivity margins for BWR plants. When compared to the existing power/flow operating domain, operation in the MELLLA region results in plant operation along a higher constant flow control line, which at off-rated operation allows for higher core power at a given core flow. This increases the fluid subcooling in the reactor vessel downcomer and alters the power distribution in the core, which can potentially affect steady state operating thermal limit and transient/accident analyses results. The effect of this operating mode has been evaluated to support compliance with the Technical Specification fuel thermal margins during plant operation. This report presents the results of the safety analyses and system response evaluations performed for operation of VYNPS in the region above the ELLLA and up to the MELLLA boundary line. The scope of the analyses performed covers the initial application for VYNPS operation with ARTS/MELLLA. For subsequent reload cycles, VYNPS will include the ARTS/MELLLA operating condition in the plant-specific reload licensing basis.
The safety analyses and system evaluations performed to justify operation in the MELLLA region consist of a non-fuel dependent portion and a fuel dependent portion that is fuel cycle dependent. In general, the limiting anticipated operational occurrences (AOOs) MCPR calculation and the reactor vessel overpressure protection analysis are fuel dependent. These analyses, as discussed in this report, are based on the assumption of a representative GE9, GEl 3, and GEl4 core (Cycle 23 core design). Subsequent cycle-specific analyses will be performed by VYNPS in conjunction with the reload licensing activities. The non-fuel dependent evaluations such as containment response are based on the current hardware design and plant geometry, and as such they are applicable to VYNPS. The limiting AOOs, as identified in Reference I, were reviewed for the MELLLA region based on a review of existing thermal analysis limits at plants similar to VYNPS and use of generic power-dependent and generic flow-dependent MCPR, Linear Heat Generation Rate (LHGR), and Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits/setpoints. For the fuel-dependent evaluations of reactor pressurization events, these reviews indicate that there is a small difference in the operating limit minimum critical power ratio (OLMCPR) for operation in the MELLLA region and the CLTP condition (100% of CLTP / 100% of RCF). The actual operating limit is calculated on a cycle specific basis to bound the entire operating domain. The analyses results also indicate that performance in the MELLLA region is within allowable design limits for overpressure protection, loss-of coolant accident (LOCA), containment dynamic loads, flow-induced vibration, and reactor internals structural integrity, and meets Anticipated Transient Without Scram (ATWS) licensing criteria.
The analyses which justify operation in the MELLLA region under the stated conditions are discussed in this report and its supporting references. These analyses include fuel performance event evaluations, mechanical evaluations of the reactor internals, structural vibration assessment, LOCA evaluations, and containment loads evaluations. NRC-approved or industry accepted computer codes and calculational techniques are used in the ARTS/MELLLA analyses.
A list of the Nuclear Steam Supply System (NSSS) computer codes used in the evaluations is 1-3
NEDO-33089 provided in Table 1-1.
1.2.2 APRM High Flux (Flow Bias) Scram and Rod Block Design Bases VYNPS employs long-term thermal hydraulic stability Option I-D that credits the APRM High Flux (Flow Bias) scram line in the low flow region of the power/flow map (Reference 2) for Safety Limit MCPR (SLMCPR) protection. Outside of the region of possible instability, the APRM High Flux (Flow Bias) scram line is conservatively not credited in any VYNPS licensing analyses. In addition, the APRM Upscale (Flow Bias) rod block line is conservatively not credited in any VYNPS safety licensing analyses, although it is part of the VYNPS design configuration. The purpose of this section is to discuss the setpoint changes for these systems for operational flexibility purposes and to provide inputs to the VYNPS Technical Specifications mark-up process. Further discussion of the APRM High Flux (Flow Bias) scram as it applies to the VYNPS Option I-D stability solution is included in Section 5.0.
For the current licensed power/flow map, the APRM High Flux (Flow Bias) scram line analytical limit (AL) is defined as: 0.66 (W - AW) + 54%, clamp at 120% of CLTP. W is defined as the recirculation drive flow for two loop operation (TLO) in percent of rated, where 100% drive flow is that required to achieve 100% core power and flow. AW is the difference between the TLO and the SLO drive flow at the same core flow; AW = 0 for TLO operation.
At the current ELLLA conditions, a single APRM High Flux (Flow Bias) scram equation is adequate for both the stability and non-stability related portions of the power/flow map. The APRM Upscale (Flow Bias) rod block line limit is currently set at: 0.66 (W - AW) + 42%, with no maximum. The RBM Upscale (Flow Bias) rod block line limits are currently set at: 0.66 (W AW) + N, with N defined in the core operating limits report (COLR).
With the current power/flow map, the operational margin between the APRM Upscale (Flow Bias) rod block line and the ELLLA Boundary line is significantly reduced, in comparison to the operational margin originally available with respect to the 100% rod line.
With the proposed power/flow map expansion to include the MELLLA region, the upper boundary of the licensed operating domain is extended to approximately the 120.8% rod line.
To accommodate this expanded operating domain, to restore the original margin between the MELLLA boundary line and the APRM Upscale (Flow Bias) rod block line, and to ensure compliance with the VYNPS long-term thermal-hydraulic stability solution (see Section 5.0), the following setpoints are redefined:
APRM High Flux (Flow Bias) scram ALs for TLO (Technical Specification 2.1 .A. l.a.) are:
0.4 W + 64.4% for 0 < W5 <31.1%
1.28 W + 37.0% for 31.1 < W5<54.0%
0.66 W + 70.5% for 54.0 < W < 75.0%
clamp at 120% of CLTP for W > 75.0%
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NEDO-33089 APRM High Flux (Flow Bias) scram ALs for SLO (Technical Specification 2.1 .A. L.a.) are:
0.4W+61.2% for 0< W<39.1%
1.28 W + 26.8% for 39.1 < W:< 61.9%
0.66 W + 65.2% for 61.9 < W < 83.0%
clamp at 120% of CLTP for W > 83.0%
APRM Upscale (Flow Bias) rod block limits for TLO are:
0.4 W + 52.4% for 0 < W:<31.1%
1.28 W + 25.0% for 31.1 < W:< 54.0%
0.66 W + 58.5% for 54.0 < W < 80.4%
maximum at 111.59% of CLTP for W > 80.4%
APRM Upscale (Flow Bias) rod block limits for SLO are:
0.4 W + 49.2% for 0 < W5 <39.1%
1.28 W+ 14.8% for 39.1 <W<61.9%
0.66 W + 53.2% for 61.9 < W < 88.4%
maximum at 111.59% of CLTP for W > 88.4%
The above AL calculations were performed using current GE ARTS/MELLLA methodology.
The RBM Upscale (Flow Bias) line limit equation remains as: 0.66 (W - AW) + N. However, because the RWE analysis does not take credit for any of these functions, the values of N will be specified in the COLR for operational flexibility in the MELLLA region. A bounding AW value of 8% for SLO was determined from previous analytical results (existing VYNPS-specific BILBO calculations). This bounding value for AW is sufficiently conservative for use in the safety analysis.
1.3 APRM Improvements The functions of the APRM system are to:
- 1. Generate trip signals to automatically scram the reactor during core-wide neutron flux transients before the actual core-wide neutron flux level exceeds the safety analysis design bases. This prevents exceeding design bases and licensing criteria from single operator errors or equipment malfunctions.
- 2. Block control rod withdrawal whenever operation occurs in excess of set limits in the operating map and before core power approaches the scram level.
- 3. Provide an indication of the core average power level of the reactor in the power range.
The VYNPS APRM system calculates an average of the in-core Local Power Range Monitor (LPRM) chamber signals using digital electronics. The LPRMs are averaged such that the APRM signal is proportional to the core average neutron flux and can be calibrated as a means of measuring core thermal power. The APRM signals are compared to a recirculation drive flow-referenced scram trip and a recirculation drive flow-referenced control rod withdrawal block trip.
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NEDO-33089 The current VYNPS APRM flow-biased setpoints are presently implemented by an analog Flow Control Trip Reference (FCTR) card installed in each of the APRM channels. The current VYNPS FCTR cards only have the ability to accommodate a single flow-biased scram equation.
The multiple APRM flow-biased equations defined in Section 1.2.2 will be implemented for VYNPS by use of digital FCTR cards as described in Reference 3 (Sections 2.1, 5.0, and 6.0).
The plant currently operates such that the core Maximum Total Peaking Factor (MTPF) does not exceed the Design Total Peaking Factor (DTPF), which limits the maximum local power at lower core power and flows to a fraction of that allowed at rated power and flow. If the MTPF exceeds the DTPF, the flow-referenced APRM trips must be lowered (setdown)' to limit the maximum power that the plant can achieve. The basis for this "APRM trip setdown" requirement originated under the original BWR design Hench-Levy Minimum Critical Heat Flux Ratio (MCHFR) thermal limit criterion and provides conservative restrictions with respect to current fuel thermal limits.
The change to the GE Thermal Analysis Basis critical power correlation, with its emphasis on bundle critical power rather than local critical heat flux allows for a more direct determination of fuel thermal limits.
The VYNPS ARTS/MELLLA application utilizes the results of the AOO analyses to define initial condition operating thermal limits, which conservatively assure that all licensing criteria are satisfied without DTPF and setdown of the flow-referenced APRM scram and rod block trips.
The objective of the APRM improvements is to justify removal of the APRM trip setdown and DTPF requirement. Two licensing areas that can be affected by the elimination of the APRM trip setdown and DTPF requirement are: (1) fuel thermal-mechanical integrity and (2) LOCA analysis.
The following criteria assure satisfaction of the applicable licensing requirements. They were applied to demonstrate the acceptability of elimination of the APRM trip setdown requirement:
- 2. All fuel thermal-mechanical design bases shall remain within the licensing limits described in the GE generic fuel licensing report GESTAR-I1 (Reference 4).
- 3. Peak cladding temperature and maximum cladding oxidation fraction following a LOCA shall remain within the limits defined in 10 CFR 50.46.
I The APRM scram and rod block trips are not actually setdown at VYNPS. Rather, an equivalent result is accomplished by APRM gain increases. It is understood that references to trip setdown in this report are literally implemented as gain increases at VYNPS.
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NEDO-33089 The safety analyses used to evaluate the OLMCPR such that the SLMCPR is satisfied and to ensure that the fuel thermal-mechanical design bases are satisfied are documented in Section 3.0 of this report. These analyses also establish the power-dependent and flow-dependent MCPR, LHGR, and MAPLHGR curves for VYNPS. The effect on the LOCA response due to the ARTS program implementation is discussed in Section 6.0 of this report.
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NEDO-33089 Table 1-1 Computer Codes Used for ARTS/MELLILA Analyses TakCode Computer Version or NRC AprvdComments Revision Approved Reactor Heat Balance ISCOR 09 Y (1) NEDE-2401 1-P Rev. 0 SER Reactor Core and Fuel TGBLA 04 Y NEDE-30130-P-A Performance PANAC 10 Y NEDE-30130-P-A ISCOR 09 Y (1) NEDE-2401 1-P Rev. 0 SER Transient Analysis PANAC 10 Y NEDE-30130-P-A (4)
ODYN 10(5) Y NEDE-24154P-A NEDC-24154P-A, Vol 4, Sup 1 ISCOR 09 Y (1) NEDE-2401 1-P Rev. 0 SER Stability Analysis ISCOR 09 Y (1) NEDE-2401 1-P Rev. 0 SER PANAC 10 Y NEDE-30130-P-A (4)
TRACG 02 Y NEDO-32465-A NEDE-32177P, Rev. 1 ODYSY 05 Y NEDC-32992P-A ECCS-LOCA LAMB 08 Y NEDE-20566P-A GESTR 08 Y NEDE-23785-1P-A, Rev. 1 SAFER 04 Y (7) (8) (9) (10) (11)
ISCOR 09 Y (1) NEDE-2401 1-P Rev. 0 SER TASC 03 Y NEDC-32084P-A Reactor Internal LAMB 07 (2) NEDE-20566P-A Pressure Differences TRACG 02 (3) NEDE-32176P, Rev. 2 NEDC-32177P, Rev. 2 NRC TAC No. M90270 ISCOR 09 Y (1) NEDE-2401 1-P Rev. 0 SER RPV Internals SAP4GO7V 01 NA NEDO-10909 (12)
Structural Integrity Evaluation Anticipated Transient PANAC 10 Y NEDE-30130-P-A (4)
Without Scram ODYN 10 (5) Y NEDC-24154P-A, Vol 4, Sup 1 STEMP 04 (6)
NA - Not Applicable Notes:
(1) The ISCOR code is not approved by name. However, the Safety Evaluation Report (SER) supporting approval of NEDE-2401 1-P Rev. 0 by the May 12, 1978 letter from D. G.
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NEDO-33089 Eisenhut (NRC) to R. Gridley (GE) finds the models and methods acceptable, and mentions the use of a digital computer code. The referenced digital computer code is ISCOR. The use of ISCOR to provide core thermal-hydraulic information in reactor internal pressure differences, Transient, ATWS, Stability, and LOCA applications is consistent with the approved models and methods.
(2) The LAMB code is approved for use in Emergency Core Cooling System (ECCS) LOCA applications (NEDE-20566P-A), but no approving SER exists for the use of LAMB for the evaluation of reactor internal pressure differences (RIPDs) or containment system response. The use of LAMB for these applications is consistent with the model description of NEDE-20566P-A.
(3) NRC has reviewed and accepted the TRACG application for the flow-induced loads on the core shroud as stated in NRC SER TAC No. M90270.
(4) The physics code PANACEA provides inputs to the transient code ODYN. The improvements to PANACEA that were documented in NEDE-30130-P-A were incorporated into ODYN by way of Amendment 11 of GESTAR II (NEDE-2401 1-P-A).
The use of PANAC Version 10 in this application was initiated following approval of Amendment 13 of GESTAR II by letter from G.C. Lainas (NRC) to J.S. Charnley (GE),
MFN 028-086, "Acceptance for Referencing of Licensing Topical Report NEDE-2401 1-P A Amendment 13, Rev. 6 General Electric Standard Application for Reactor Fuel,"
March 26, 1998.
(5) Version 10 of ODYN is applicable to plants that use variable pump speed (i.e., motor generator set, induction motor drive, or internal pumps) for recirculation flow control.
(6) The STEMP code uses fundamental mass and energy conservation laws to calculate the suppression pool heatup. The use of STEMP was noted in NEDE-24222, "Assessment of BWR Mitigation of ATWS, Volume I & II (NUREG-0460 Alternate No. 3) December 1, 1979." The code has been used in ATWS applications since that time. There is no formal NRC review and approval of STEMP or the ATWS topical report.
(7) NEDE-30996P-A, "SAFER Model for Evaluation of Loss-of-Coolant Accidents for Jet Pump and Non-Jet Pump Plants," October 1987.
(8) NEDE-23785-1-PA, Vol. III, Rev. 1, "GESTR-LOCA and SAFER Models for Evaluation of Loss-of-Coolant Accident (Volume III), SAFER/GESTR Application Methodology,"
October 1984.
1-9
NEDO-33089 (9) NEDE-23785P-A, Vol. III, Supplement 1, Rev. 1, "GESTR-LOCA and SAFER Models for Evaluation of Loss-of-Coolant Accident Volume III, Supplement 1, Additional Information for Upper Bound PCT Calculation," March 2002.
(10) NEDC-32950P, "Compilation of Improvements to GENE's SAFER ECCS-LOCA Evaluation Model," January 2000.
(11) Letter, S.A. Richards (NRC) to J.F. Klapproth (GE), "General Electric Nuclear Energy Topical Reports NEDC-32950P and NEDC-32084P Acceptability Review," May 24, 2000.
(12) Not a safety analysis code that requires NRC approval. The code application is reviewed and approved by GENE for "Level-2" application and is part of GENE's standard design process. Also, the application of this code has been used in previous power uprate submittals.
1-10
NEDO-33089 Figure 1-1 ARTS/MELLLA Power/Flow Map Core Flow (Mlb/hr) 0 10 20 30 40 50 120 A: Natural Circulation 1800 110 B: Minimun Puop Speed MELLLA Upper Boundary Line C: 64 01 Power/ 36.0. Flow D: 100.0 Fower/ 75.0: Flow 100 E: :00.0% Power/ 87.0% Flow 1593 MWt 1600 108% CLTP Rod F: 100.0% Power/ 100.0% Flow G. :00.01 Power/ 10.0% Flow
':20.0% Power! 107.0% Flow (1) 90 1400
- 1. 20.0% Power/ 100.0% Flow (I)
J. 20.0% Power/ 32.7% Flow 1)
Note (1) : VY does not contain puep 80 cavitation protection line, this line A
is the minimum power line. 1200 70
- i. 100% CLTP Rod Line C
U.
L.) 0 0 60 CU 800 2 50 a
C.. Minimum Pump Spee, 40 600 Natural Circulation Minimum Power Line .
30 20
/ J ---------------------- --.- 6 H 400 100% CLTP - 1593 MWt 200 l00% OLTP = 593 MWt 10 100% Core Flow : 4 0 MIb/hr 0 0 0 10 20 30 40 50 60 70 80 90 100 110 120 Core Flow (%)
1-11
NEDO-33089 2.0 OVERALL ANALYSIS APPROACH This section identifies the analyses that may be affected by the proposed MELLLA region. The analyses performed in the following sections assume the current plant operating parameters. For the transient and stability tasks, the VYNPS Cycle 23 core design was utilized (Reference 1),
and these tasks will be revalidated as part of the subsequent cycle-specific reload licensing analyses. For the remainder of the ARTS/MELLLA scope of work, the results are applicable to VYNPS, unless a plant configuration affecting this analysis is changed.
Table 2-1 identifies the safety and regulatory concerns that are potentially affected as a result of ARTS/MELLLA. Each applicable safety and regulatory concern implied in the listed items was reviewed to determine the acceptability of changing the power/flow map to include the MELLLA range. In addition, the characteristics of each analyses, whether generic or plant specific, and cycle-dependent or cycle-independent, are identified in Table 2-2.
2-1
NEDO-33089 Table 2-1 Analyses Presented In This Report Section Item Result 3.0 Fuel Thermal Limits Acceptable - Bounded by Current Results 4.0 Vessel Overpressure Protection Acceptable - Bounded by Current Results 5.0 Thermal-Hydraulic Stability Acceptable - New Region for ARTS/MELLLA 6.0 LOCA Analysis Acceptable - Bounded by Current Results 7.0 Containment Response Acceptable - Bounded by Current Results or Design Criteria 8.0 Reactor Internals Integrity Acceptable - Bounded by Current Results 9.0 ATWS Acceptable with the installation of an additional spring safety valve (SSV).
10.0 Steam Dryer and Separator Acceptable - Bounded by Current Results Performance 11.0 Testing Acceptable with the performance of the identified tests.
2-2
NEDO-33089 Table 2-2 Applicability of Analyses Task Description Generic or Plant-Specific Cycle-Independent or Cycle-Dependent Power-dependent MCPR, Generic, with plant-specific Cycle-independent unless LHGR, and MAPLHGR limits confirmation for initial change in plant configuration (between rated power and application from licensing analysis basis 30% of CLTP)
Power-Dependent MCPR, Plant-specific Cycle-specific review LHGR, and MAPLHGR limits (between 30% and 25% of CLTP)
Flow-dependent MCPR, Generic, with plant-specific Cycle-independent unless LHGR, and MAPLHGR limits confirmation for initial change in plant configuration application from licensing analysis basis ECCS-LOCA Plant-specific Cycle-independent unless change in plant configuration from licensing analysis basis 2-3
NEDO-33089 3.0 FUEL THERMAL LIMITS The potentially limiting AOOs and accident analyses were evaluated to support VYNPS operation with ARTS off-rated limits, as well as operation in the MELLLA region. The power/flow state points chosen for the review of AOOs indicated in Table 3-1 bound the current licensed operating domain for VYNPS and the MELLLA region. These evaluations are discussed in Sections 3.1 through 3.3. Section 3.5 discusses the governing MCPR, LHGR, and MAPLHGR limits that also include consideration of the RWE analyses (Section 3.4) and the LOCA analyses (Section 6.0).
3.1 Limiting Core-Wide AOO Analyses The potentially limiting AOOs analyzed for the current VYNPS Cycle 23 reload licensing analysis (Reference 1) were examined for operation with ARTS off-rated limits and in the MELLLA region.
The following events were considered potentially limiting and were evaluated for the ARTS program generic off-rated limits development:
(1) Generator Load Rejection with No Bypass (LRNBP) event; (2) Turbine Trip with No Bypass (TTNBP) event; (3) Feedwater Controller Failure (FWCF) maximum demand event; (4) Loss of Feedwater Heating (LFWH) event; (5) Inadvertent High Pressure Coolant Injection (HPCI) Startup event; (6) Idle Recirculation Loop Start-up (IRLS) event; and (7) Recirculation Flow Increase (RFI) event.
The analytical methods and input assumptions used for the VYNPS-specific evaluations were consistent with the bases used in Reference 1. The LFWH, HPCI, and IRLS events were not re evaluated here for the following reasons.
The LFWH event is not limiting for VYNPS and the effect of MELLLA on the LFWH severity is sufficiently small that the LFWH remains not limiting for MELLLA. The required MCPR for VYNPS Cycle 23 LFWH is 1.21 (Reference 1) based on an 87% initial core flow, compared to an end-of-cycle (EOC) GE14 Option B OLMCPR of 1.44 from the LRNBP. At 75% initial core flow, the OLMCPR from the LFWH would still show a large margin to the LRNBP, TTNBP, and FWCF events as shown in Reference 1. Consequently, the LFWH was not analyzed at the 3-1
NEDO-33089 MELLLA point. However, it should be noted that the LFWH event is analyzed on a cycle specific basis.
Finally, considering that the LFWH event tends to become less limiting as the power decreases (less feedwater to be affected by loss of heating), the LFWH event was not considered in the determination/validation of the off-rated limits.
The HPCI evaluation at 100% CLTP for VYNPS Cycle 23 (Reference 1) also showed a large margin for OLMCPR (1.29 for HPCI versus 1.44 for LRNBP and 1.42 for FWCF). This conclusion will not be altered when considering the MELLLA region. The HPCI event tends to be more severe as the initial power decreases (ratio of HPCI flow to initial feedwater flow increases). However, at low initial powers, the subcooling due to FWCF bounds the subcooling due to HPCI. Consequently, the HPCI event was not considered in the determination/validation of the off-rated limits.
The IRLS and RFI events are most limiting at off-rated conditions. Even when originated from their most limiting off-rated condition, the IRLS and RFI events are typically less limiting than the fast pressurization events (TTNBP, LRNBP, or FWCF) at rated power conditions. Thus, the IRLS and RFI events were not considered for the MELLLA application. As previously stated, these events were considered generically in the development of the ARTS flow-dependent limits.
[
] Consequently, the IRLS event was not performed for VYNPS for the ARTS/MELLLA application. However, the VYNPS licensing analysis basis contains an atypical two recirculation pump flow runout event (this event is not considered a credible AOO at most other GE BWRs, which do consider a one recirculation pump runout). Therefore, a VYNPS specific RFI analysis was performed to generate plant specific flow-dependent MCPR, LHGR, and MAPLHGR limits.
3.1.1 Elimination of APRM Trip Setdown and DTPF Requirement Extensive transient analyses at a variety of power and flow conditions were performed during the original development of the ARTS improvement program. These evaluations are applicable for operation in the MELLLA region. The analyses were utilized to study the trend of transient severity without the APRM trip setdown. A database was established by analyzing limiting transients over a range of power and flow conditions. The database included evaluations representative of a variety of plant configurations and parameters such that the conclusions drawn from the studies would be applicable to all BWRs. The database was utilized to develop a method of specifying plant operating limits (MCPR, LHGR, and MAPLHGR), which assures that margins to fuel safety limits are equal to or larger than those applied currently.
The generic evaluations determined that the power-dependent severity trends must be examined in two power ranges. The first power range is between rated power and the power level (PBypass) where reactor scram on turbine stop valve closure or turbine control valve fast closure is bypassed. The analytical value of PBypass for VYNPS is 30% of CLTP. The second power range 3-2
NEDO-33089 is between PBypass and 25% of CLTP. No thermal monitoring is required below 25% of CLTP, per VYNPS Technical Specification 3.11.
Generic power-dependent MCPR, LHGR, and MAPLHGR limits (in terms of multipliers on the plant's rated operating limits) were developed for use in the first power range (above PBypaSS).
VYNPS specific analyses of limiting transients were performed to confirm the applicability of the generic power-dependent limits above PBypss.
Between PBypams and 25% power, VYNPS specific evaluations were performed to establish the plant-unique MCPR, LHGR, and MAPLHGR limits in the low power range (below PBypas.;).
These plant-specific limits include sufficient conservatism to remain valid for future VYNPS reloads of GE fuels through the GE14 fuel design, utilizing the GEXL-PLUS correlation and the GEMINI analysis methods, except that the power-dependent MCPR limits below PBypass must be adjusted in accordance with Section 3.3.5 if the SLMCPR exceeds 1.10.
VYNPS specific evaluations were performed to establish the plant-unique flow-dependent MCPR, LHGR, and MAPLHGR limits. These plant-specific limits include sufficient conservatism to remain valid for future VYNPS reloads of GE fuels through the GEl4 fuel design, utilizing the GEXL-PLUS correlation and the GEMINI analysis methods, as long as the core flow corresponding to the maximum two recirculation pump runout is <5 109.5% of RCF, except that the flow-dependent MCPR limits must be adjusted in accordance with Section 3.3.5 if the SLMCPR exceeds 1.10.
3.2 Input Assumptions The limiting power/flow state condition for the operating region analysis is the rated power and maximum flow point (100%P / 107%F). Figure 1-1 shows the power/flow map used in the AOO analyses. Plant heat balance, core coolant hydraulics, and nuclear dynamic parameters corresponding to the rated and off-rated conditions were used for the analysis and reflect the VYNPS Cycle 23 core configuration (Reference 1). The initial heat balance conditions for the AOO analyses at rated and off-rated conditions are presented in Tables 3-la and 3-lb.
Because of the fuel cycle-independent nature of the ARTS thermal limits (for both above and below P1ypass power ranges), the ARTS transients analyses assumed rated core thermal power as 1593 MWt. All AOO analyses were performed using the standard reload licensing methodology (Reference 4). The following assumptions and initial conditions were used in the AOO analyses:
Analytical Assumptions Bases/Justifications Initial core flow range of 75% to 107% flow Bounding power/flow state points for MELLLA for thermal limits transients at 100% of CLTP Conservative end-of-Cycle 23 nuclear Consistent with VYNPS current licensing bases dynamic parameters I 3-3
NEDO-33089 Analytical Assumptions Bases/Justifications The lowest opening setpoint safety-relief Consistent with VYNPS current licensing bases valve (SRV) declared Out-of-Service (OOS)
SLMCPR = 1.10 Consistent with VYNPS current licensing bases The LFWH, HPCI, and IRLS events are not Consistent bases of the ARTS program limiting at off-rated conditions.
3.3 Analyses Results In summary, the operating limits associated with operation in the MELLLA region are presented in Table 3-2. The MELLLA region will also be incorporated into subsequent cycle specific reload licensing analyses.
The GE9 fuel in the VYNPS Cycle 23 core and future cores is not limiting for MCPR, LHGR, or MAPLHGR considerations. Consequently, GE9 is not included in the tables of results. All off rated limits are satisfactory for application to GE9 fuel.
3.3.1 Power-Dependent MCPR Limit As stated previously, the generic evaluations indicate that the power-dependent severity trends are to be examined in two power ranges, above and below Paypass. Above PBypaSS, bounding power-dependent trend functions have been developed. These trend functions, K(P), are used as multipliers to the rated MCPR operating limits to obtain the power-dependent MCPR limits, MCPR(P), or OLMCPR(P) = K(P) x OLMCPR (P = 100% of CLTP)
In the high power range (between rated power and PBypass), the trend for the power-dependent MCPR responses for the FWCF event is more severe than all other fast pressurization transient severity trends. As power is reduced from the rated condition in this power range, the LRNBP and TTNBP events become relatively less severe because the reduced steam flow rate at low power results in milder reactor pressurization. However, for the FWCF event, the power decrease results in greater mismatch between runout and initial feedwater flow, resulting in an increase in reactor subcooling and more severe changes in thermal limits during the event.
The results used to verify the generic MCPR(P) limits analyses are summarized in Table 3-3. As previously stated, the MCPR(P) is derived from the generic K(P) multiplied by the rated power OLMCPR. For power levels above PBypans, the formula for calculating the generic K(P) is given in Figure 3-1. A comparison of the plant-specific calculated values with the generic power-dependent MCPR limits verifies the applicability of the generic limits to VYNPS.
Below Paypas,, the transient characteristics change due to the bypass of the direct scram on the closure of the turbine stop valve and turbine control valve. Consequently, the scram signal is delayed until the vessel pressure reaches the high pressure scram setpoint. The extensive 3-4
NEDO-33089 transient analyses database shows a significant sensitivity to the initial core flow for transients initiated below PBypass. Therefore, both high and low core flow sets of power-dependent limits are determined for power levels above 25% and below PBypass.
Below PBypass, the MCPR(P) limits are actual absolute OLMCPR values, rather than multipliers on the rated power OLMCPR. These absolute MCPR limits were chosen with sufficient conservatism such that they remain applicable to future operating cycles provided the SLMCPR is less than or equal to 1.10 (Technical Specification 1.I.A.1.). The VYNPS specific analyses results used to establish the MCPR(P) at power levels below PBypass are summarized in Table 3-4.
The VYNPS below-Pnyps MCPR(P) limits are given in Figure 3-1.
3.3.2 Power-Dependent LHGR and MAPLHGR Limits In the absence of the APRM trip setdown requirement, power-dependent LHGR and MAPLHGR limits, expressed in terms of multipliers, LHGRFAC(P) and MAPFAC(P) are substituted to assure adherence to the fuel thermal-mechanical design bases. The power-dependent LHGRFAC(P) and MAPFAC(P) multipliers were generated using the same database as used to determine the MCPR multiplier, K(P). These factors are also applied in a similar manner.
Specifically, MAPLHGR(P) = MAPFAC(P) x (rated MAPLHGR limits) and LHGR(P) = LHGRFAC(P) x (rated LHGR limits).
The LHGRFAC(P) and MAPFAC(P) multipliers are identical. However, the LHGR(P) limits do not apply to PANAC 10 applications.
For GE fuel designs, both incipient centerline melting of the fuel (thermal over-power (TOP))
and plastic strain of the cladding (mechanical over-power (MOP)) are considered in determining the power-dependent MAPLHGR limits [
]
Similar to the MCPR(P) limits, VYNPS-specific transient analyses were performed to demonstrate the applicability of the generic LHGR(P) and MAPLHGR(P) limits. The transient and initial condition selection are identical to that previously described for MCPR(P). The applicable results of these analyses for power levels above PBypass are shown in Table 3-5. The generic LHGRFAC(P) and MAPFAC(P) above PBypas are shown in Figure 3-2.
As previously discussed, a significant sensitivity to initial core flow exists below PBypass.
Therefore, below PBypass, both high and low core flow sets of power-dependent LHGR and MAPLHGR multipliers are provided. To prevent the situation where the limits are more restrictive after increasing power above PBypass, the extrapolation of the generic above PByass limits are taken as the upper bound for the below PBypass limits. Appropriate LHGRFAC(P) and MAPFAC(P) multipliers are selected based on plant-specific transient analyses with suitable margin to assure applicability to future VYNPS reloads, including exposure ranges of GE fuels through the GEl4 fuel design. These limits are derived to assure that the peak transient LHGR 3-5
NEDO-33089 or MAPLHGR for any transient is not increased above the fuel design basis values. The results of plant-specific transient analyses below PBypass are presented in Table 3-6. The plant-specific LHGRFAC(P) and MAPFAC(P) below PByp.sq are shown in Figure 3-2.
3.3.3 Flow-Dependent MCPR Limit Flow-dependent MCPR limits, MCPR(F), are necessary to assure that the SLMCPR is not violated during recirculation flow increase events. The design basis flow increase event is a slow-flow power increase event which is not terminated by scram, but which stabilizes at a new core power corresponding to the maximum possible core flow. This event was also used to determine the current MCPR flow multiplier Kr. For the VYNPS two recirculation pump runout event, it was previously determined that the Kr was valid for runout rates up to I %/sec, as long as the rated OLMCPR __SLMCPR + 0.20.
The generic MCPR(F) basis flow runout events were analyzed [
] For the VYNPS flow runout analysis, a 1%/sec runout rate was slow enough to allow the use of steady state methods, but too fast to allow equilibrium feedwater temperature. Consequently, a VYNPS-specific flow runout was analyzed to determine plant-specific flow-dependent limits. The flow runout was performed similar to the generic basis, except that the feedwater temperature is held constant at the low initial value. This results in a more severe power increase than would occur with an equlibrium feedwater temperature assumption. The ARTS generic MCPR(F) boundary for a runout to a maximum core flow of 112% of RCF is sufficient to bound the VYNPS specific flow runout to 109.5% of RCF (See Table 3-8 for a comparison of the generic design rod line and the values calculated for VYNPS with constant feedwater temperature). The flow-dependent MCPR limit, as shown in Figure 3-3, is cycle-independent, except that it must be adjusted in accordance with Section 3.3.5 if the SLMCPR is greater than 1.10.
To verify the applicability of the original ARTS generic flow-dependent MCPR limits, the RFI and IRLS events were re-performed in a generically applicable manner for the GE14 New Fuel Introduction (NFI). For the application of ARTS, the IRLS basis is an initial 50'F AT between the idle and operating loops. This is an appropriate assumption for thermal limits calculations and it is consistent with Technical Specification requirements.
3.3.4 Flow-Dependent LHGR and MAPLHGR Limits Flow-dependent LHGR and MAPLHGR limits were designed to assure adherence to all fuel thermal-mechanical design bases. The same transient events used to support the MCPR(F) operating limits were analyzed, and the resulting overpowers were statistically evaluated as a function of the initial and maximum core flow. From the bounding overpowers, the LHGRFAC(F) and MAPFAC(F) multipliers were derived such that the peak transient LHGR would not exceed fuel mechanical limits. The LHGR(F) and MAPLHGR(F) limits are 3-6
NEDO-33089 cycle-independent and are specified in terms of multipliers, LHGRFAC(F) and MAPFAC(F), to be applied to the rated LHGR and MAPLHGR values. Specifically, MAPLHGR(F) = MAPFAC(F) x (rated MAPLHGR limits) and LHGR(F) = LHGRFAC(F) x (rated LHGR limits).
The LHGRFAC(F) and MAPFAC(F) multipliers are identical. However, the LHGR(F) limits do not apply to PANAC I0 applications.
The LHGRFAC(F) and MAPFAC(F) multipliers are shown in Table 3-9 and Figure 3-4.
3.3.5 Safety Limit MCPR Adjustment Procedure The MCPR limits, provided in Figures 3-1 and 3-3 assume a dual-loop SLMCPR of 1.10. The off-rated MCPR(P) is defined by Figure 3-1. Only adjustment of the P < PBypass portion of the MCPR(P) curve may be required because, at P > PBypass, the K(P) applies the rated power OLMCPR adjustment to the MCPR(P). The off-rated MCPR(F) is defined by Figure 3-3. When necessary, adjustment to the entire MCPR(F) limit is required.
Should a future cycle SLMCPR exceed 1.10, the MCPR(F) and below PBypass MCPR(P) limits must be increased by the following factor:
Cycle specific SLMCPR']
1.10 )
Should a future cycle SLMCPR be less than 1.10, the MCPR(F) and below-PBypa"s MCPR(P) limits may optionally be reduced by the above factor.
3.3.6 Single Loop Operation Adjustment Procedure When operating in SLO, an adjustment will be made to the rated power OLMCPR as well as the off-rated OLMCPR. The off-rated MCPR(F) is defined by Figure 3-3. The off-rated MCPR(P) is defined by Figure 3-1. Only adjustment of the P < Paypas portion of the MCPR(P) curve is required because, at P > PBypass, the K(P) applies the rated power OLMCPR adjustment to the MCPR(P). The equation for the adjustment is as follows when operating in SLO:
SLO OLMCPR = OLMCPRdual-Ioop + SLMCPRsLo - SLMCPRduaII.oop 3.4 Rod Withdrawal Error Analysis The VYNPS application of ARTS is a partial application. VYNPS will not be implementing the hardware changes that are usually installed to the RBM system. The hardware changes to the RBM system would typically provide the required protection for an off-rated RWE event.
Therefore, the evaluation of the RWE event was performed without taking credit for the mitigating effect of the flow-biased RBM setpoints. The results of four analyses for VYNPS 3-7
NEDO-33089 Cycle 23 (30% power / 35% flow, 30% power / 107% flow, 60% power / 47% flow, and 60%
power / 107% flow) are summarized in Table 3-7 and provide the OLMCPR results at off-rated conditions for the unblocked RWE event.
The ARTS definition of a limiting control rod pattern (LCRP) is one for which the RBM is required to prevent violating a thermal limit in the event of an RWE. Based on a statistical analysis, the LCRP concept has been implemented for standard ARTS plants as a requirement on the operability of the RBM for MCPR < 1.40 (typical) for > 90% power, or MCPR < 1.70 (typical) for < 90% power. Because the VYNPS ARTS RWE basis does not credit the RBM, the LCRP concept is no longer meaningful for VYNPS.
For Cycle 24 and subsequent reloads, analyses will be performed to provide a statistically-based power-dependent RWE OLMCPR. The analyses will follow the approach utilized to establish the standard ARTS basis RBM operability Technical Specification limits.
3.5 Conclusion The rated OLMCPRs, LHGRs, and MAPLHGRs are determined by the cycle-specific fuel reload analyses. Then, at any given power/flow state (P,F), all applicable off-rated limits are determined: MCPR(P), MCPR(F), LHGR(P), LHGR(F), MAPLHGR(P) and MALHGR(F). The most limiting MCPR (maximum of MCPR(P) and MCPR(F)), the most limiting LHGR (minimum of LHGR(P) and LHGR(F)), and the most limiting MAPLHGR (minimum of MAPLHGR(P) and MAPLHGR(F)) will be the governing limits. Finally, the limits should be adjusted for SLMCPRs > 1.10 or SLO, as applicable.
3-8
NEDO-33089 Table 3-1a Base Conditions for ARTS/MELLLA Rated Transient Analyses Normal 75%F MELLLA 107%F ICF Power (MWt/% of CLTP) 1593/100 1593/100 1593/100 Flow (Mlb/hr/ % rated) 48.0/ 100 36.0/75 51.36/ 107 Steam Flow (MIb/hr) 6.458 6.448 6.461 FW Temperature (*F) 376.0 375.9 376.0 Core Inlet Enthalpy (Btu/lb) 521.1 512.0 522.9 Dome Pressure (psig) 1010 1010 1010 Table 3-lb Base Conditions for ARTS/MELLLA Off-rated Transient Analyses 80%P/107%F 60%P/107%F 60%P/47%F 45%P/107%F 30%P/107%F Power (MWt) 1274.4 955.8 955.8 716.85 477.9 Flow (Mlb/hr) 51.36 51.36 22.56 51.36 51.36 Steam Flow (Mlb/hr) 5.038 3.670 3.653 2.682 1.731 FW Temperature (F) 356.1 332.0 331.7 309.7 2804 Core Inlet Enthalpy (Btu/ib) 523.3 524.7 502.6 526.5 528.9 Dome Pressure (psig) 989 971 971 960 950 30%P/60%F 30%P/35%F 25%P/107%F 25%P/60%F 50.4P%/30%F Power (MWt) 477.9 477.9 398.25 398.25 802.9 Flow (Mlb/hr) 28.8 16.8 51.36 28.8 14.40 Steam Flow (MIb/hr) 1.717 1.714 1.424 1.410 3.015 FW Temperature ('F) 279.9 279.8 268.0 267.4 317.9 Core Inlet Enthalpy (Btu/lb) 521.0 508.3 529.8 523.0 487.0 Dome Pressure (psig) 950 950 948 948 963 3-9
NEDO-33089 Table 3-2 MELLLA Transient Analyses Peak Values, Cycle 23 Peak Initial Peak Peak Steam Peak Power / Flow Neutron Heat Line Vessel
(% Rated) Flux Flux TOP (%) ACPR(b) OLMCPR(c) Pressure Pressure Transient(') (% Initial) (% Initial) GEI4/GE13 GE14/GEI3 GE14/GE13 (psig) (psig) 100 / 100 - RWE(d) 0.48 /0.48 1.58/ 1.58 100 / 107-EOC LRNBP 316.45 116.92 29.18 / 29.18 0.29/0.24 1.44/1.36 1176.5 1208.4 TTNBP 322.99 117.27 28.34 / 28.34 0.28 / 0.23 1.44 / 1.35 1179.2 1209.7 FWCF 293.18 119.39 26.49 / 26.49 0.26/0.23 1.42/1.35 1090.2 1131.3 100 / 75- EOC LRNBP 207.88 107.20 20.25 / 19.95 0.22 / 0.16 1.37 / 1.27 1172.4 1195.5 TTNBP 205.21 109.54 21.77 / 21.43 0.23/0.17 1.38/1.29 1176.0 1198.5 FWCF 177.13 111.00 19.80/ 19.36 0.20/0.15 1.35 / 1.27 1087.6 1116.1 Notes:
(a) Under-Bum power shape for MELLLA and ICF transients.
(b) ACPR calculated, uncorrected.
(c) OLMCPR Option B.
(d) Unblocked.
3-10
NEDO-33089 Table 3-3 ARTS Transient Analysis Results - Generic K(P) Confirmation Above P-Bypass Initial Transient ACPR (a) OLMCPR-A (b) OLMCPR-B (') Limiting Generic Power I Flow GEI4IGE13 GEI4/GEI3 GE14/GE13 Calculated K(P)
(%Rated) EOC EOC EOC K(P) (d) 100/107 LRNBP 0.29/0.24 1.61/1.47 1.44/1.36 TTNBP 0.28 /0.23 1.61 / 1.46 1.44 / 1.35 FWCF 0.26/0.23 1.59/1.46 1.42/1.35 1.0 1.0 80/ 107 LRNBP 0.28/0.24 1.51 / 1.39 1.40/1.34 TTNBP 0.28 / 0.23 1.50 / 1.38 1.39 / 1.33 FWCF 0.28/0.25 1.51/1.40 1.40/1.35 1.012 1.075 60 / 107 LRNBP 0.25 / 0.22 1.48 / 1.37 1.37 / 1.32 TTNBP 0.25/0.22 1.48/1.37 1.37/1.32 FWCF 0.30/0.28 1.53/1.43 1.42/1.38 1.038 1.150 45 / 107 LRNBP 0.20 / 0.18 1.43 / 1.33 1.32 / 1.28 TTNBP 0.20/0.18 1.43/1.33 1.32/1.28 FWCF 0.33 / 0.31 1.56 / 1.46 1.45 / 1.41 1.062 1.280 30/107 LRNBP 0.13/0.12 1.35/1.27 1.24/1.22 TTNBP 0.13/0.12 1.35/1.27 1.24/1.22 FWCF 0.38/0.38 1.61 / 1.53 1.50/1.48 1.119 1.4825 Notes:
(a) ACPR based on initial CPR which yields MCPR 1.10, uncorrected for Options A and B.
(b) OLMCPR for Option A.
(c) OLMCPR for Option B.
(d) The calculated K(P) considers the maximum OLMCPR calculated for any transient in that category divided by the operating limit for that category including exposure dependence. [
I 3-11
NEDO-33089 Table 3-4 ARTS Transient Analysis Results - MCPR(P) Below P-Bypass Option A Initial ACPR (b) OLMCPR(c)
Power / Flow GEI4/GEI3 GE14/GEI3 Calculated Limiting
(%Rated) Transient(") EOC EOC MCPR(P) (d) MCPR(P) 30/ 107 LRNBP 0.90/0.84 2.26/2.18 2.38 TTNBP 0.89 / 0.84 2.26 / 2.18 2.37 FWCF 0.40/0.40 1.58/ 1.59 1.64 2.38 30/60 LRNBP 0.66/0.61 1.92/1.86 2.00 TTNBP 0.66 / 0.61 1.92 / 1.86 2.00 FWCF 0.26/0.24 1.41 / 1.38 1.44 2.00 25 / 107 LRNBP 1.00 / 0.95 2.42 / 2.34 2.55 TTNBP 1.00 / 0.95 2.42 / 2.34 2.55 FWCF 0.47/0.47 1.67/ 1.68 1.73 2.55 25 / 60 LRNBP 0.74 / 0.69 2.03 / 1.97 2.13 TTNBP 0.74 / 0.69 2.03 / 1.97 2.12 FWCF 0.26/0.24 1.41 / 1.39 1.44 2.13 Notes:
(a) For these cases, only the high pressure scram is functional. For the FWCF events, the vessel pressure did not reach the pressure scram setpoint, therefore, no scram occurs.
(b) ACPR based on initial CPR which yields MCPR = 1.10, uncorrected for Option A.
(c) Option A OLMCPR = 1.10 / (1.0 - A/1 (95/95)).
(d) [ ]
3-12
NEDO-33089 Table 3-5 ARTS Transient Analysis Results - Generic MAPFAC(P) Confirmation Above P-Bypass Limiting Limiting Initial GE14/GEI3 GEI4/GEI3 Calculated Calculated Power / Flow TOP MOP TOP-Based MOP-Based Generic
(%Rated) Transient EOC EOC MAPFAC(P)(2) MAPFAC(P)(b) MAPFAC(P) 100/107 LRNBP 29.18/29.18 29.28/29.18 TTNBP 28.34 /28.34 28.45 / 28.34 FWCF 26.49 / 26.49 26.78 / 26.49 1.000 1.000 1.000 80 / 107 LRNBP 27.97 / 27.97 27.97 / 27.97 TTNBP 27.36 / 27.36 27.36 / 27.36 FWCF 27.09 / 27.09 27.40 /27.09 0.980 1.000 0.896 60/107 LRNBP 24.31 / 24.43 24.40 / 24.43 TTNBP 24.32 / 24.43 24.41 / 24.43 FWCF 28.99 / 28.44 29.39 / 30.46 0.966 1.000 0.791 45 / 107 LRNBP 17.84 / 17.93 17.87 / 17.93 TTNBP 17.85 /17.91 17.86/17.91 FWCF 30.96 / 29.45 33.55 / 34.48 0.951 1.000 0.713 30 / 107 LRNBP 10.59 / 10.59 10.59 / 10.59 TTNBP 10.41 /10.41 10.41 / 10.41 FWCF 38.45 / 37.72 40.73 / 38.54 0.900 0.870 0.634 Notes:
[
I 3-13
NEDO-33089 Table 3-6 ARTS Transient Analysis Results - MAPFAC(P) Below P-Bypass Limiting Initial GE14/GEI3 Calculated Power I Flow TOP TOP-Based
(%Rated) Transient EOC MAPFAC(P)(b),(c) 30/107 LRNBP 81.05 /77.75 TTNBP 80.45 / 77.15 FWCF 41.79 / 40.93 0.628 30 / 60 LRNBP 53.53 / 53.67 TTNBP 52.83 / 52.97 FWCF 27.37 /26.61 0.739 25/107 LRNBP 93.81/90.91 TTNBP 93.44 /90.55 FWCF 48.76 / 48.13 0.586 25 / 60 LRNBP 61.33 / 54.90 TTNBP 60.89 / 54.47 FWCF 26.31 /25.39 0.704 Notes:
[
I 3-14
NEDO-33089 Table 3-7 Summary of Unblocked OLMCPR Values for the RWE Event Initial OLMCPR (Generic K(P))
- Power I Flow Values OLMCPR(100) (a)
(%Rated) 100/100 1.58 1.58 60/107 1.58 1.82 60/47 1.69 1.82 30/107 1.75 2.34 30/35 1.71 2.34 Note:
(a) Calculated based on the 100/100 unblocked RWE OLMCPR.
3-15
NEDO-33089 Table 3-8 ARTS Generic Design RFI Rod Line for MCPR(F) Determination Compared to VYNPS RFI Results Flow (% Rated) Generic Design Power VYNPS Calculated Power
(% Rated) (% Rated) 30 50.4 40 58.7 50 66.9 60 74.6 70 82 80 89.3 90 96.8 100 104.2 109.5 111.6 112 Note:
(a) Determined by linear interpolation between the powers corresponding to the 100% and 112% flow points.
3-16
NEDO-33089 Table 3-9 ARTS Transient Analysis Results - MAPFAC(F)
Flow Runout Calculated Calculated Calculated Calculated VYNPS (Initial -> Final, MOP95 /95(a) MOP-Based TOPgs/95 (c) TOP-Based MAPFAC(F)
%Rated) (%) MAPFAC(F)(b) (%) MAPFAC(F)(d) Limit(e) 30-+ 109.5 110.81 0.560 103.59 0.614 0.540 40 -> 109.5 81.71 0.646 76.53 0.708 0.627 50 --- 109.5 61.24 0.735 57.04 0.796 0.715 60 -- 109.5 46.04 0.977 41.86 0.881 0.802 70-* 109.5 34.11 1.000 30.92 0.955 0.889 80-- 109.5 24.39 1.000 22.74 1.000 0.977 90-- 109.5 15.38 1.000 15.68 1.000 1.000 100 -> 109.5 7.25 1.000 7.77 1.000 1.000 Notes:
[
I 3-17
NEDO-33089 Figure 3-1 Power-Dependent MCPR Limits, MCPR(P) 2460
>*60% Flow 240 2.20 Z:5 60%Flow Operating Limit MCPR (P) = K(P) x Operating Limit MCPR (100) 2.00 For P < 25%: No Thermal Limits Required For 25%:5 P < 30%, > 60% Flow.
OLMCPR(P) = 2.55- 0 034x(P - 25%)
For 25%:5 P < 30%, *5 60% Flow:.
1 OLMCPR(P) = 2.13 - 0.026x(P -25%)
a.
a- For 30% : P < 45%: K(P) = 1.55 -0 0135x(P- 25%)
- .For 45% < P < 60%: K(P)= 1.28 - 0.008667x(P- 45%)
0 160 For 60% 5 P < 100%: K(P) = 1.15 - 0.00375x(P - 60%)
For P;> 100%: K(P) = 1.00 1.40 ____
1.20 100 .
25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 Power (% Rated) 3-18
NEDO-33089 Figure 3-2 Power-Dependent LHGR or MAPLHGR Multiplier, LHGRFAC(P) or MAPFAC(P) 1.10 1.00 0 90 0.80 0.70 0.5 LHGR(P) = LHGRFAC(P)xLHGRstd MAPLHGR(P) MAPFAC(P)xMAPLHGRstd o 0.60 where:
-J MAPLHGRstd = Rated MAPLHGR limits IL For P< 25%: No Thermal Limits Required O, 0.50 For 25%:5 P < 30%, > 60% Flow:
LHGRFAC(P) = MAPFAC(P) = 0.586+ 0.0084x(P - 25%)
For 25%:5 P < 30%, *.. 60% Flow.
040 LHGRFAC(P) = MAPFAC(P) = 1.0 + 0.0052239x(P - 100%)
For30% :5 P < 100%:
LHGRFAC(P) = MAPFAC(P) =1.0 + 0.0052239x(P - 100%)
0.30 For P > 100%:
2 LHGRFAC(P) = MAPFAC(P) = 1.0 0.20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 Power (% Rated) 3-19
NEDO-33089 Figure 3-3 Flow-Dependent MCPR Limits, MCPR(F) 1.80 For W (% Rated Core Flow) > 30% AND Max Runout Flow < 109.5%:
1.70 MCPR(F) = (1.10/1.07)xMAX{1.20, [1.747 - (0.602)x(W/100)]}
1.60 N C
a..
0
- 011.50 N
S1.40 CL 0
1.30 1.20 1.10 - 4 - 4. - 4 d 4- C- C- 9- 9- C-30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 Core Flow (% Rated)
Note: For core flow > 100% of RCF, MCPR(F) remains equal to approximately 1.234.
3-20
NEDO-33089 Figure 3-4 Flow-Dependent LHGR or MAPLHGR Multiplier, LHGRFAC(F) or MAPFAC(F) 1.100 r1 1.000 0.900 0.800
" 0.700 C.,
~0.600 0
L) 0.500 L0.
"=" 0.400
-J !t MAPLHGR(F) = MAPFAC(F)xMAPLHGRstd LHGR(F) = LHGRFAC(F)xLHGRstd 0 300 MAPLHGRstd are the standard MAPLHGR limits LHGRstd are the standard LHGR limits 0 200 For W (% Rated Core Flow) > 30% AND Max Runout Flow < 109.5%:
MAPFAC(F) = LHGRFAC(F) = MIN{1.0, [0.2779 + 0.8737x(W/100)]}
0.100 n noto 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 Core Flow (% Rated)
Note: For core flow > 100% of RCF, LHGRFAC(F) and MAPFAC(F) remain equal to 1.0 3-21
NEDO-33089 4.0 VESSEL OVERPRESSURE PROTECTION The MSIV closure with a flux scram (MSIVF) event is used to determine the compliance to the ASME Pressure Vessel Code. This event was previously analyzed at the 102%P / 107%F state point for the VYNPS Cycle 23 reload licensing transient analysis. This is a cycle-specific calculation performed at 102% of CLTP and the maximum licensed core flow (maximum flow is limiting for this transient for VYNPS). Because the implementation of ARTS/MELLLA does not change the maximum core flow, ARTS/MELLLA does not affect the vessel overpressure protection analysis. However, the sensitivity of operation at the MELLLA condition (102%P /
75%F for this analysis) for VYNPS Cycle 23 is provided in Table 4-1.
4-1
NEDO-33089 Table 4-1 VYNPS Cycle 23 Sensitivity of Overpressure Analysis Results to Initial Flow Initial Peak Steam Peak Vessel Peak Steam Power / Flow Dome Pressure Pressure Line Pressure
(%Rated) (psig) (psig) (psig) 102/107 1279.4 1302.7 1278.2 102/75 1246.3 1264.3 1245.9 4-2
NEDO-33089 5.0 THERMAL-HYDRAULIC STABILITY The stability compliance of GE fuel designs with regulatory requirements of the NRC is documented in Section 9 of Reference 4. The NRC approval of the stability performance of GE fuel designs also includes operation in the MELLLA region of the power/flow map.
The above NRC acceptance of thermal-hydraulic stability includes the condition that the plant has systems and procedures in place, supported by Technical Specifications, as appropriate, which provide adequate instability protection.
5.1 Stability Option I-D VYNPS has implemented long-term thermal hydraulic stability solution Option I-D. Option I-D is only applicable to plants which can demonstrate that core wide mode instability is the predominant mode and regional mode instability is not expected. Generally, a smaller core size produces higher eigenvalue separation between oscillation modes and tighter core inlet orifice coefficients make regional mode oscillations unlikely. Option I-D has: 1. "Prevention" elements (Exclusion and Buffer Regions), and 2. A "detect & suppress" element (SLMCPR protection provided by the flow-biased APRM flux trip for the dominant core wide mode of coupled thermal-hydraulic/neutronic reactor instability). Solution application consists of calculating an administratively controlled exclusion region (per Reference 5) and demonstrating that the existing APRM High Flux (Flow Bias) scram line, considered an analytical limit, provides adequate SLMCPR protection (per Reference 2). The Option I-D exclusion region is core and fuel cycle dependent and represents a curved line of constant stability margin. The APRM High Flux (Flow Bias) trip protection is also fuel cycle dependent.
The NRC-approved ODYSY methodology (Reference 6) was applied to the Cycle 23 stability analysis. ODYSY applications offer the benefit of more accurate simulations of BWR stability events and conditions. The exclusion region demonstration is affected by operating conditions.
The actual region demonstration will be performed using the ODYSY code when MELLLA operation is introduced. Typical Exclusion and Buffer Regions are illustrated in Figure 5-1 for MELLLA operation.
Based on the Cycle 23 core wrapup, an ODYSY analysis was performed at the most limiting condition on the power/flow map (i.e., the intersection of the natural circulation line and the MELLLA line, corresponding to 59.3% of CLTP, 31.3% of RCF, and 0% drive flow).
[
The detect and suppress calculation consists of: 1. Calculation of a 95% probability / 95%
confidence level statistically-based hot bundle oscillation magnitude for anticipated core-wide mode reactor instability, and 2. Calculation of the stability-based OLMCPR which provides 95/95 SLMCPR protection. The detect and suppress calculation requires the use of the Delta 5-1
NEDO-33089 CPR over Initial MCPR Versus the Oscillation Magnitude (DIVOM) curve. Recent TRACG evaluations have shown that the generic core-wide DIVOM curve specified in Reference 2 may not be conservative for current plant operating conditions for plants which have implemented Stability Option I-D. Specifically, a non-conservative deficiency has been identified for high power-to-flow ratios in the generic core-wide mode DIVOM curve. The deficiency results in a non-conservative slope of the associated core-wide DIVOM curve so that the APRM flux trip setpoint is too high. A Part 21 Notification was made on this issue (Reference 7). For Option I D plants, the applicability of the core wide mode DIVOM curve may be determined by comparing the core average power-to-flow ratio following a simulated flow runback on the rated rod line to approximately 30% of RCF to a value of 66 MWt/Mlbm/hr. If the core average power-to-flow ratio exceeds this value, then the generic core-wide mode DIVOM curve is not applicable and appropriate corrective actions should be taken. Based on the Cycle 23 wrapup and MELLLA operating conditions, the calculated core average power-to-flow ratio is 58.0 MWt/Mlbm/hr and the generic DIVOM slope is valid for MELLLA operation.
The SLMCPR protection demonstration is affected by operating conditions. The new APRM High Flux (Flow Bias) scram line for the ARTS/MELLLA operation was determined with the following additional conservatisms in the evaluation:
- 1. Cycle 23 MELLLA operating condition was used in this analysis;
- 2. [
- 4. [
The detect and suppress evaluation shows that adequate SLMCPR protection is provided by the new APRM High Flux (Flow Bias) scram line. The equation of the revised APRM High Flux (Flow Bias) scram line is: 1.6204 Wc+ 13.68 (where W, is % rated core flow), which is applicable for both TLO and SLO operation. Because of the conservatism in the detect and suppress analysis, an approximation that does not exceed this flow-biased flux scram line by more than 1% in rated power is acceptable. Reload stability evaluations, as described in Reference 5, will continue to validate this equation as adequate for the VYNPS Stability Option I-D. Therefore, MELLLA operation is justified for plant operation with Stability Option I-D.
The APRM trip system utilizes recirculation drive flow as the input variable for calculating the APRM High Flux (Flow Bias) scram setpoints. In addition, the relationship between recirculation drive flow and reactor core flow is non-linear at the low core flow area of the power-flow map. Therefore, two equations, in terms of recirculation drive flow, were derived to map the reactor core flow based APRM High Flux (Flow Bias) scram equation (1.6204 Wc+
13.68). These equations are as follows:
5-2
NEDO-33089 0.4 W + 64.4% For 0 < W:< 31.1%
1.28 W + 37.0% For 31.1 < W5 <54.0%
where W is percent of rated recirculation drive flow The stability based APRM High Flux (Flow Bias) scram line intercepts the MELLLA operating domain, non-stability based APRM High Flux (Flow Bias) scram line (0.66 W + 70.5 as defined in Section 1.2.2) at 57% core flow (54% recirculation drive flow). Therefore, it is conservative that the APRM trip system utilizes the non-stability based APRM flow-biased equation at recirculation drive flows greater than 54%, and up to 75%. The maximum AL clamps at 120%
power for recirculation drive flows greater than 75%. The APRM High Flux (Flow Bias) scram lines are shown in Figure 5-1. This figure also provides a comparison between the ELLLA and MELLLA APRM High Flux (Flow Bias) scram lines.
The replacement of the analog FCTR card installed in each of the APRM channels by digital FCTR cards is described in Section 1.3.
5-3
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NEDO-33089 6.0 LOSS-OF-COOLANT ACCIDENT ANALYSIS The current licensing basis SAFER/GESTR-LOCA analysis for VYNPS (Reference 8) has been reviewed to determine the effect on the ECCS performance resulting from plant operation in the MELLLA domain. The Reference 8 analysis considered VYNPS operation in the ELLLA domain.
The two major parameters that affect the fuel peak cladding temperature (PCT) in the design basis LOCA calculation which are sensitive to the higher load line in the operating power/flow map are the time of boiling transition (BT) at the high power node of the limiting fuel assembly and the core recovery time. Initiation of the postulated LOCA at lower core flow may result in earlier BT at the high power node, compared to the 100% of RCF results, resulting in a higher calculated PCT. Similarly, initiation of the postulated LOCA at lower core flow affects break flow rate and core reflooding time, compared to the 100% of RCF results, which can also result in a higher calculated PCT. The effect on the calculated PCT is acceptable as long as the results remain less than the Licensing Basis PCT limits.
The current VYNPS licensing basis also specifies a requirement in maximum LHGR as a function of drive flow, known as the APRM setdown requirement. This lower LHGR requirement is applicable to core flows lower than 87% of RCF and has the effect of a less limiting PCT calculation for LOCA analyses at less than the 100% of RCF conditions. However, this lower LHGR requirement is being replaced with direct core power and flow fuel thermal limits by the ARTS improvement option, and is therefore not credited in the LOCA analysis for ARTS/MELLLA.
An evaluation was performed with GE14 and GEl3 fuel to determine the ECCS-LOCA analysis effects of VYNPS operation in the MELLLA region. The SAFER/GESTR-LOCA methodology was first applied to VYNPS in Reference 8. The limiting Design Basis Accident (DBA) was evaluated to show that the estimated PCT remains below the acceptance limits. The initial conditions for the VYNPS LOCA analysis that were used in this determination are listed in Table 6-1. A comparison of the SAFER/GESTR-LOCA results between MELLLA conditions and RCF operation is presented in Table 6-2 for the limiting GE14 and GE13 fuel types. The MELLLA results are shown for both nominal and Appendix K assumptions, because these results are the key contributors to the Upper Bound and Licensing Basis PCT, respectively.
The previously calculated Licensing Basis PCT from the VYNPS Cycle 23 / GE14 NFI analysis was < 1950'F. For MELLLA, a Licensing Basis PCT of 1960'F was determined for GE14 fuel; a Licensing Basis PCT of 1910°F was determined for GE13 fuel, which also bounds GE9 fuel.
This new bounding GE14 Licensing Basis PCT of 1960OF provides adequate margin to the 2200'F PCT limit. The maximum local oxidation is less than 3%. The core-wide metal-water reaction is less than 0.1%.
Further, the VYNPS Cycle 23 / GE14 NFI analysis provides justification for the elimination of the 1600'F Upper Bound PCT limit. Thus, based on the MELLLA results in Table 6-2, the 6-1
NEDO-33089 conclusions in Reference 9 are applicable, which state that the Licensing Basis PCT is conservative with respect to the Upper Bound PCT. Compliance with the coolable geometry and long-term cooling acceptance criteria were demonstrated generically for GE BWRs (Reference 10). Therefore, the requirements of 10 CFR 50 Appendix K are satisfied.
6.1 Conclusions The evaluation of the sensitivity of the ECCS-LOCA evaluation to operation in the MELLLA domain will meet all of the ECCS-LOCA acceptance criteria. Therefore, there are no ECCS LOCA analysis related plant operating restrictions due to the incorporation of ARTS/MELLLA.
6-2
NEDO-33089 Table 6-I DBA LOCA Initial Conditions for VYNPS ARTS/MELLLA Plant Parameter Nominal Appendix K Core Thermal Power (MWt / % CLTP) 1665.0 / 104.52 1698.3 / 106.61 Vessel Steam Output (Mlbm/hr) 6.768 6.925 Core Flow (% of RCF) 75.0 75.0 Vessel Steam Dome Pressure (psia) 1055 1070 Maximum RSLB Area (ft2) (a) 4.16 4.16 Note:
- a. The DBA LOCA break area includes the maximum Recirculation Suction Line Break (RSLB) area and bottom head drain flow path area.
Table 6-2 (a,b)
DBA LOCA Results Comparison for VYNPS ARTS/MELLLA Core Flow Nominal Appendix K Nominal Appendix K GE14 GE14 GE13 GE13 (OF) (OF) (OF) (OF) 100%F (Rated) (c) 1287 1839 1333 1858 75%F 75% ) 1381 1943 1396 1884 (MELLLA)
Notes:
- a. The effect of operation in the MELLLA domain on the ECCS-LOCA analysis PCT for both GE14 and GE13 is conservatively applicable to GE9 fuel.
- b. The assumed single failure is the failure of a DC power source (i.e., battery failure).
- c. The rated flow PCT results are from the VYNPS Cycle 23 / GE14 NFI analysis, which presents updated analysis inputs with respect to Reference 8.
6-3
I NEDO-33089 7.0 CONTAINMENT RESPONSE The changes in reactor vessel operating conditions due to MELLLA can affect the short-term containment system response to the DBA-LOCA. This covers the blowdown period (less than 30 seconds into the event) during which the maximum drywell pressure occurs. The long-term containment responses are not affected by the MELLLA conditions, because the decay heat does not change and there is a negligible difference in the vessel sensible heat in the MELLLA operating domain.
Analyses of the short-term containment response to the DBA-LOCA were previously performed for various cases that cover the full extent of the MELLLA power/flow boundary for VYNPS.
The pressure and temperature response results from these analyses (Reference 11) have been incorporated into the Updated Final Safety Analysis Report (UFSAR) (Sections 5.2.4 and 14.6.3). Therefore, the current licensing basis for the containment analysis bounds the MELLLA conditions. As discussed in the UFSAR, the highest peak drywell pressure was determined to be 38.2 psig, which is well below the design pressure of 56 psig. Also, as documented in Reference 11, the containment hydrodynamic loads (pool swell, vent thrust, condensation oscillation, and chugging loads) were evaluated using the short-term DBA-LOCA pressure and temperature response results that cover the full extent of the MELLLA power/flow boundary.
The evaluation concluded that the existing DBA-LOCA hydrodynamic load definitions remain applicable. In addition, the SRV loads do not change because the SRV setpoints, and the capacity of the vacuum breakers installed on the SRV discharge lines, do not change for MELLLA operation.
Thus, based on the current licensing basis analysis for the containment, the containment system response at MELLLA conditions is acceptable.
7-1
NEDO-33089 8.0 REACTOR INTERNALS INTEGRITY 8.1 Reactor Internal Pressure Differences The increase in RIPDs across the reactor internal components and the fuel channels in the MELLLA condition are bounded by the ELLLA (87% of RCF) and the ICF (107% of RCF) conditions due to the lower core flow condition in MELLLA (75% of RCF). Thus, no new RIPDs or fuel bundle lift are generated for the MELLLA condition. The current RIPD basis in Reference 12 remains applicable to the MELLLA condition. In addition, the VYNPS load definition basis does not include the control rod guide tube (CRGT) load condition.
8.2 Acoustic and Flow-Induced Loads The acoustic and flow-induced loads are contributing factors to the VYNPS design basis load combination in the Faulted condition. The acoustic loads are imposed on the reactor internal structures as a result of the propagation of the decompression wave created by the assumption of an instantaneous RSLB. The acoustic loads affect the core shroud, core shroud repair components, core shroud support, and jet pumps. The flow-induced loads are imposed on the reactor internal structures as a result of the fluid velocities from the discharged coolant during an RSLB. The flow-induced loads affect the shroud and jet pumps.
The acoustic and flow-induced loads in the MELLLA condition are slightly higher than the current ELLLA condition due to the increased subcooling in the downcomer associated with the MELLLA condition. From ELLLA to MELLLA, the downcomer subcooling increases thereby increasing the critical flow and the mass flux out of the break in a postulated RSLB. As a result, the acoustic and flow-induced loads in MELLLA conditions increase slightly.
8.2.1 Approach/Methodology Major components in the vessel annulus region, the shroud, shroud support, and jet pumps were evaluated for the bounding RSLB acoustic and flow-induced loads representing the MELLLA conditions.
The flow-induced loads were calculated for an RSLB utilizing the specific VYNPS geometry and fluid conditions applied to a reference BWR calculation. The loads were calculated by applying scaling factors that account for plant-specific geometry differences (e.g., size of the shroud, reactor vessel, and recirculation line) and thermal-hydraulic condition differences (e.g.,
downcomer subcooling) from the reference plant. The reference calculation was based on the GE methods utilized to support NRC Generic Letter 94-03 that was issued to address the shroud cracks detected at some BWRs.
The acoustic loads on the jet pumps and shroud applied for VYNPS represent VYNPS-specific plant geometry configuration and operating conditions. The bounding subcooling and natural frequencies for the jet pumps and shroud are applied. For acoustic loads on the shroud support, 8-1
NEDO-33089 generic bounding BWR loads based on the GE methods were used for the flow-induced load calculation.
For VYNPS, the most limiting subcooling condition is at the intersection of the minimum pump speed and the MELLLA flow control line. The subcooling at this point is applied to the reference BWR calculation, along with the VYNPS geometry, to determine the specific flow induced loads.
8.2.2 Input Assumptions The following assumptions and initial conditions were used in the determination of the acoustic and flow-induced loads for the MELLLA operation. [
]
Initial Conditions Bases/Justifications 102%P / 100%F Consistent with the VYNPS current licensing basis.
102%P / 87%F ELLLA corner at rated power at nominal rated feedwater temperature.
102%P / 75%F MELLLA comer at rated power at nominal rated feedwater temperature.
61%P / 36%F Minimum pump speed point on the ELLLA line at nominal rated feedwater temperature.
64%P / 36%F Bounding power/flow state point for MELLLA; minimum pump speed point on the MELLLA line at nominal rated feedwater temperature 8.2.3 Results The flow-induced loads for the shroud and jet pumps are shown in Table 8-1. VY-specific flow induced load multipliers for off-rated conditions to be applied to the baseline loads are also documented. The maximum acoustic loads on the shroud and jet pumps are shown in Table 8-2.
The maximum acoustic loads on the shroud support are shown in Table 8-3. These loads were used to determine the structural integrity of these components.
8.3 Structural Integrity Evaluation The structural integrity of the reactor internals for the loads associated with MELLLA operation for VYNPS was evaluated considering the original design basis loading combinations. The RIPDs, vessel dome pressure, seismic, and fuel lift margin are either unchanged or bounded by those for the current (ICF / GEl3) condition. However, the RSLB LOCA loads are affected by the MELLLA conditions. The effect of MELLLA operation on the other loads such as flow and temperature were also considered in the evaluation, as appropriate. The load inputs (e.g.,
RIPDs) for the MELLLA evaluation incorporated operation with GE14 fuel.
8-2
NEDO-33089 The only components affected by acoustic and flow-induced loads are the shroud, shroud repair components, shroud support, and the jet pumps. An assessment was performed for the Faulted condition RSLB LOCA loads in combination with the seismic loads. The resultant stresses compared to allowables are shown in Table 8-4. The horizontal welds on the core shroud were assumed to have failed, thus the shroud repair was installed. Because the condition of the horizontal welds was unknown, an evaluation was also performed assuming all the welds were intact. This case was not limiting for maximum stresses. All other reactor internals remain unaffected by MELLLA or have insignificant effect on their structural integrity due to MELLLA (i.e., MELLLA loads remain unchanged or bounded by those of the current condition, or the change is insignificant).
The following key Reactor Pressure Vessel (RPV) internal components were reviewed:
"* Core Plate
"* Top Guide
"* Control Rod Drive Housing
"* Control Rod Guide Tube
"* Orificed Fuel Support
"* Fuel Channel
"* Jet Pumps
"* Core Spray Line and Sparger
"* Access Hole Cover
"* Shroud Head and Steam Separator Assembly
"* Shroud
"* Shroud Repair Components
"* Shroud Support VYNPS shroud effective loads were compared to the allowable stress limits for the shroud and shroud repair, and determined to remain within the limits.
The VYNPS core support structure and non-core support structure reactor internals are "non ASME code" components and there are no specific code requirements that apply. However, ASME Code,Section III criteria were used as a guide where applicable, consistent with the original design basis of the components.
A qualitative review of the existing flaw evaluation handbook for the core spray line and sparger was performed. The net temperature change for the shroud and the stilts for MELLLA operation was estimated to be about 1% and 3%, respectively, which is insignificant. Therefore, the existing flaw evaluation for the core spray line and sparger is valid for the MELLLA condition.
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NEDO-33089 A qualitative review of the existing flaw evaluation handbook for the jet pump was performed.
The change in the RSLB LOCA loads was about 2%, which is also insignificant. Therefore, the existing flaw evaluation for the jet pump is judged to be valid for the MELLLA condition.
8.3.1 Conclusion Based on the structural integrity evaluation of the reactor internals, all of the reactor internals listed in Section 8.3 remain qualified for the MELLLA condition. The existing flaw evaluations of the core spray line and sparger and the jet pump remain valid based on a qualitative evaluation for the MELLLA condition.
8.4 Reactor Internals Vibration The reactor internals vibration characteristics can be affected by the increased rod line associated with MELLLA operation.
8.4.1 Approach/ Methodology To ensure that the flow-induced vibration (FIV) response of the reactor internals is acceptable, a single reactor for each product line and size undergoes an extensively instrumented vibration test during initial plant startup. After analyzing the results of such a test and assuring that all responses fall within acceptable limits of the established criteria, the tested reactor is classified as a valid prototype. All other reactors of the same product line and size are classified as non prototype and undergo a less rigorous confirmatory test.
Monticello Nuclear Generating Plant (MNGP), a BWR3, 205-inch diameter reactor, was designated as the prototype plant for VYNPS. FIV tests were performed at MNGP and data was collected during plant start-up between December 1970 and July 1971. The critical reactor internals were instrumented with vibration sensors at MNGP and the reactor was tested up to 100% of RCF at the 100% rod line. This data was used in the current evaluation of VYNPS for MELLLA operation.
VYNPS is currently licensed to operate at an ICF of up to 107% of RCF (51.4 Mlb/hr) at 100%
of CLTP. For MELLLA operation, the rated power output remains the same, but core flow is reduced to 75% of RCF as shown in Figure 1-1.
8.4.2 Inputs/Assumptions The following inputs/assumption were used in the reactor internals vibration evaluation:
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NEDO-33089 Parameter Input Plant data selected for flow induced MNGP was designated as the prototype plant for vibration evaluation VYNPS, which is a BWR4, 205-inch diameter reactor.
FIV data collected during MNGP plant start-up between December 1970 and July 1971 was used.
During the startup, the reactor was tested up to 100%
of RCF at the 100% rod line.
Target plant conditions in the CLTP of 1593 MWt and 75% of RCF (120.8% rod MELLLA region selected for line) with balanced flow conditions.
component evaluation GE stress acceptance criterion of The GE limit is lower than the value allowed by the 10,000 psi is used for all stainless steel current ASME Section III design codes for the same components material and is bounding for all stainless steel material. The ASME Section III value is 13,600 psi for service cycles equal to 10'1.
8.4.3 Analyses Results Because the vibration levels generally increase as the square of the flow and MELLLA flow rates are lower than CLTP flow rates with power remaining unchanged, CLTP vibration levels bound those at MELLLA conditions.
The reactor internals vibration measurements report for MNGP was reviewed to determine which components are likely to have significant vibration at MELLLA conditions. Only the jet pump sensing lines (JPSLs) are affected by MELLLA. The vibration responses of JPSLs subjected to recirculation pump pressure pulsations at the vane passing frequency (VPF) were analyzed in detail. Based on the analysis results, the recirculation pump was determined not to have an adverse effect on the JPSLs during MELLLA operating conditions.
Because the vibration levels are generally proportional to the square of the flow, the lower plenum components (CRGT, Incore Guide Tube (ICGT)) and the jet pumps whose vibrations are dependent on the core flow, will experience reduced vibration due to the reduction in core flow during MELLLA operation. Hence, the vibration levels of those components at MELLLA conditions are bounded by those at CLTP conditions.
For the shroud and separator assembly and the steam dryer, the vibrations are a function of steam flow. The MELLLA steam flow conditions are bounded by the steam flow at CLTP conditions.
For the feedwater sparger, the vibrations are a function of the feedwater flow, which at MELLLA conditions is bounded by the feedwater flow at CLTP conditions.
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NEDO-33089 The jet pump riser braces were evaluated for possible resonance due to VPF pressure pulsations.
The jet pump riser braces natural frequencies are well separated from the recirculation pump VPF during MELLLA conditions and will not have any increased vibrations.
The FIV evaluation is conservative for the following reasons:
"* The GE criteria of 10,000 psi peak stress intensity is more conservative than the ASME allowable peak stress intensity of 13,600 psi for service cycles equal to 101 1;
"* The modes are absolute summed; and
"* The maximum vibration amplitude in each mode is used in the absolute sum process, whereas in reality the vibration amplitude fluctuates.
Therefore, the FIV will remain within acceptable limits.
8.4.4 Conclusion The analyses documented in this section demonstrate that, from an FIV viewpoint, the reactor internals structural mechanical integrity is maintained to provide VYNPS safe operation in the MELLLA domain.
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NEDO-33089 Table 8-1 Flow-induced Loads on Shroud and Jet Pumps for VYNPS Component Parameter Loads (a)
Shroud Baseline Force (lbf) 350,520 Baseline Moment at the Shroud Centerline (in-lbf) 22,508,000 Baseline Force (lbf) 25,724 Jet Pump Baseline Moment at the Jet Pump Centerline (in-lbf) 1,375,000 Load Component Operating Condition Multiplier o b) 102%P / 100%F [
102%P / 87%F (ELLLA)
Jet Pump 102%P / 75%F (MELLLA) 61%P / 36%F (ELLLA) 64%P / 36%F (MELLLA) ]
Component Operating Condition' Load Multiplier (c) 102%P / 100%F [
102%P / 87%F (ELLLA)
Shroud 102%P / 75%F (MELLLA) 61%P / 36%F (ELLLA) 64%P / 36%F (MELLLA) ]
Note:
(a) Loads at rated conditions.
(b) Loads multipliers [ ] in critical break flow assumption.
(c) Loads multipliers [ ] in critical break flow assumption.
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NEDO-33089 Table 8-2 Maximum Acoustic Loads on Shroud and Jet Pumps Effective Force Effective Moment Moment Component Conditions (Ibf).' Force (Ibf) * (in-lbf) (in-lbf)
MELLLA 64%P / A 2.389E+06 1.082E+06 2.873E+08 1.052E+08 64%P / 36%F Shroud ELLLA 1%P / F2.388E+06 1.080E+06 2.873E+08 1.049E+08 61%P / 36%F MELLLA 64%P / 36 3.526E+04 3.233E+04 1.872E+06 1.97 1E+06 ump Jet 64%P / 36%FI Jet Pump EL A 6 P /%F3.158E+04 2.934E+04 1.665E+06 1.698E+06 161%P /36%Fl Table 8-3 Maximum Acoustic Loads on Shroud Support (MELLLA condition, 64%P / 36%F)
Component Parameter Unit Loads Total Vertical Force lbf 2.20E6 Shroud Support Moment at the Shroud Support Plate in-lbf 3.24E8 Outside Edge Nearest the Break Half Period sec 0.037 8-8
NEDO-33089 Table 8-4 Summary of Structural Evaluation for ARTS/MELLLA Component Faulted Allowable Stress (psi) Stress (psi)
Shroud 43,800 60,000 Shroud Support 53,100 55,900 Jet Pump Diffuser 53,165 57,600 Component Faulted Code Allowable Load (Ibs) (lbs)
Shroud Repair Components 248,800 269,125 (Support - Buckling Load) 8-9
NEDO-33089 9.0 ANTICIPATED TRANSIENT WITHOUT SCRAM 9.1 Approach/Methodology The basis for the current ATWS requirements is 10 CFR 50.62. This regulation includes requirements for an ATWS Recirculation Pump Trip (RPT), an Alternate Rod Insertion (ARI) system, and an adequate Standby Liquid Control System (SLCS) injection rate. The purpose of the ATWS analysis is to demonstrate that these systems are adequate for plant changes associated with operation in the MELLLA region. This is accomplished by performing a plant specific analysis in accordance with the approved licensing methodology (Reference 13), to demonstrate that ATWS acceptance criteria are met for operation in the MELLLA region.
The ATWS analysis takes credit for ATWS-RPT and SLCS, but assumes that ARI fails. If reactor vessel and fuel integrity are maintained, then the ATWS-RPT setpoint is adequate for the proposed plant modifications. If containment integrity is maintained, then the SLCS injection rate is adequate for the proposed plant modifications.
MELLLA conditions provide the greatest effect on peak vessel pressure and peak long-term containment response (suppression pool temperature and containment pressure). The SRVOOS and SRV setpoint tolerance relaxation provide the greatest effect on the peak short-term reactor vessel pressure. The analysis assumed an initial power level of 1593 MWt (100% of CLTP) with the corresponding MELLLA minimum core flow of 75% of RCF. The limiting value of one SRVOOS and 3% SRV setpoint tolerance relaxation values were used in this analysis.
Two limiting ATWS events for VYNPS were re-evaluated at the most limiting MELLLA point (100% of CLTP and 75% of RCF) with ARI assumed to fail, thus requiring the operator to initiate SLCS injection for shutdown. These limiting events were:
(1) Closure of all MSIVs (MSIVC); and (2) Pressure Regulator Failure (Open) to Maximum Steam Demand Flow (PRFO).
The loss of offsite power (LOOP) and inadvertent opening of a relief valve (IORV) events were also considered, but found to be non-limiting. The following ATWS acceptance criteria were used to determine acceptability of the VYNPS operation in the MELLLA region:
(1) Fuel integrity:
"* Maximum clad temperature < 2200'F
"* Maximum local clad oxidation < 17%
(2) RPV integrity:
(3) Containment integrity:
- Peak suppression pool bulk temperature < 281 'F 9-1
- Peak containment pressure < 62 psig Fuel integrity is ensured by meeting the 2200'F PCT and the 17% local cladding oxidation acceptance criteria of 10 CFR 50.46. [
]
The adequacy of the margin to the SLCS relief valve lifting as described in NRC Information Notice 2001-13, "Inadequate Standby Liquid Control System Relief Valve Margin," was also assessed.
9.2 Input Assumptions Along with the initial operating conditions given in Table 9-1, the following assumptions were used in the analysis:
Analytical Assumptions Bases/Justifications Both beginning-of-cycle (BOC) and EOC Consistency with generic ATWS evaluation nuclear dynamic parameters were used in the bases.
calculations.
Dynamic void and Doppler reactivity are ATWS analyses are performed based on VYNPS Cycle 23 data. conservatively compared to a nominal basis, which bounds cycle to cycle variation. Thus, utilization of VYNPS Cycle 23 fuel parameters are appropriate.
One SRVOOS, specified as the valve with Consistency with the VYNPS Technical the lowest setpoint. Specifications.
SRV setpoints are adjusted to be consistent Consistency with the VYNPS Technical with the 3% setpoint tolerance relaxation. Specifications.
An additional SSV will be installed prior to In order to ensure compliance with the the implementation of MELLLA. reactor vessel overpressure criterion.
MSIV closure starts at event initiation (time Consistency with generic ATWS evaluation zero) for the MSIVC event, bases.
Maximum combined flow through the main Conservatively bounds current VYNPS steam line flow limiters is 125% of rated UFSAR PRFO analysis basis.
steam flow (PRFO event).
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NEDO-33089 Currently, VYNPS has four SRVs and two SSVs. Prior to the implementation of MELLLA, an additional SSV will be installed with the capacity listed in Table 9-1.
9.3 Analyses Results Tables 9-2 and 9-3 summarize the key transient responses for the MSIVC and PRFO events analyzed. As shown, the peak vessel bottom pressure result for both events analyzed is 1367 psig, which is below the ATWS vessel overpressure protection criterion of 1500 psig. The highest calculated peak suppression pool temperature is 183'F, which is well below the ATWS limit of 281IF. Also, the peak containment pressure of 11.1 psig is a small fraction of the 62 psig limit. Thus, the RPV and containment integrity criteria for ATWS are met.
However, compliance with the applicable acceptance criteria necessitates the installation of a third SSV. The additional SSV will be designed and installed in accordance with applicable codes and standards; the installation will be evaluated for jet impingement considerations. The heat load to the drywell by this additional SSV was evaluated and found to be acceptable.
The maximum SLCS pump discharge pressure during the limiting ATWS event is 1320 psig.
This value is based on a peak reactor vessel lower plenum pressure of 1290 psia that occurs during the PRFO BOC event at the time of SLCS initiation. Therefore, the required test pressure specified in the Technical Specification 4.4.A. 1. has been changed.
As discussed in Section 9.1, the PCT and local cladding oxidation criteria need not be evaluated to demonstrate compliance with the ATWS Rule for MELLLA operation.
Finally, there is adequate margin to prevent the SLCS relief valve from lifting (per NRC Information Notice 2001-13). With a nominal SLCS relief valve setpoint of 1400 psig, there is a margin of 80 psi between the peak SLCS pump discharge pressure and the relief valve nominal setpoint.
9.4 Conclusions The results of the ATWS analysis performed for VYNPS to support operation in the MELLLA region show that the maximum values of the key performance parameters (reactor vessel pressure, suppression pool temperature, and containment pressure) remain within the applicable limits. Therefore, VYNPS operation in the MELLLA region has no adverse effect on the capability of the plant systems to mitigate postulated ATWS events in the expanded operating region.
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NEDO-33089 Table 9-1 Operating Conditions and Equipment Performance Characteristics for ATWS Analyses Parameter Current Analysis Dome Pressure (psig) 1010 Core Flow (Mlb/hr / % rated) 36.0/75 Core Thermal Power (MWt /%NBR) 1593/ 100 Steam / Feed Flow (Mlb/hr / %NBR) 6.45 / 99.8 Feedwater Temperature ('F) 376.0 Initial Void Reactivity Coefficient - BOC / EOC (c/%) -16.8 / -12.0 Core Average Void Fraction - BOC / EOC (%) 58.7 / 46.5 Sodium Pentaborate Solution Concentration in the SLCS 10.42 Storage Tank (% by weight)
Nominal Boron 10 Enrichment (atom %) 43.0 SLCS Injection Location Lower Plenum Number of SLCS Pumps Operating One SLCS Injection Rate (gpm) 40.5 SLCS Liquid Transport Time (see) 33.3 Initial Suppression Pool Liquid Volume (ft3) 68,000 Initial Suppression Pool Temperature ('F) 90 Initial Suppression Pool Mass (Mlbm) 4.216 Number of RHR cooling loops 2 RHR heat exchanger effectiveness (Btu/sec-0 F) 176 Service Water Temperature (*F) 85 High Dome Pressure ATWS-RPT Setpoint (psig) 1150 SRV Capacity - per valve (lbm/hr) / Reference Pressure (psig) 800,000 / 1080 SSV Capacity - per valve (lbm/hr) / Reference Pressure (psig) 932,500 / 1240 SRV / SSV Configuration (a) 4/2 Note:
(a) Currently, VYNPS has four SRVs and two SSVs. Prior to the implementation of MELLLA, an additional SSV will be installed with the capacity listed above. The ATWS analysis assumed a four SRV / two SSV configuration, which is shown through evaluation to be representative of a three SRV / three SSV alignment, i.e., one SRVOOS.
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NEDO-33089 Table 9-2 Summary of Key Parameters for Short-term ATWS Calculation MSIVC PRFO BOC EOC BOC EOC Peak Vessel Bottom 1360 1357 1367 1362 Pressure (psig)
Time of Peak Vessel 10.4 9.9 38.9 42.2 Pressure (see)
Peak Neutron Flux (% 210 224 224 696 rated)
Time of Peak Neutron 4.1 4.1 29.9 33.0 Flux (see)
Peak Vessel Heat Flux 127 129 135 139
(% rated)
Time of Peak Heat 5.8 5.2 34.0 36.8 Flux (see)
Table 9-3 Summary of Key Parameters for Long-term ATWS Calculation MSIVC PRFO BOC EOC' BOC EOC Peak Suppression Pool 170 183 172 182 Temperature ('F)
Time of Peak 1259 1485 1672 1574 Temperature (see)
Peak Containment 8.5 11.1 8.8 10.8 Pressure (psig)
Time of Peak Pressure 1259 1485 1672 1574 (see) 9-5
NEDO-33089 Table 9-4 ATWS Calculation Sequence of Events (Time in seconds)
Event MSIVC PRFO BOC EOC BOC EOC Turbine control and bypass valves - - 0.1 0.1 start open MSIV isolation initiation 0.0 0.0 24.7 28.9 MSIVs closed 4.0 4.0 28.7 32.9 High pressure ATWS setpoint 4.5 4.5 29.9 33.0 Opening of the first relief valve 4.5 4.5 32.8 36.5 Recirculation pumps tripped 5.0 5.0 33.2 37.0 Boron injection initiation 80 80 108 112 temperature achieved SLCS pumps start 124.5 124.5 152.7 156.4 Boron solution reaches lower 157.8 157.8 186.0 189.7 plenum RHR cooling initiated 660 660 660 660 Hot shutdown achieved 1431 1650 1569 1740 9-6
NEDO-33089 10.0 STEAM DRYER AND SEPARATOR PERFORMANCE The ability of the steam dryer and separator to perform their design functions during MELLLA operation was evaluated. MELLLA decreases the core flow rate, resulting in an increase in separator inlet quality for constant reactor thermal power. These factors, in addition to core radial power distribution, affect the steam separator-dryer performance. Steam separator-dryer performance was evaluated to determine the effect of MELLLA on the steam dryer and separator operating conditions, the entrained steam (i.e., carryunder) in the water returning from the separators to the reactor annulus region, the moisture content in the steam leaving the RPV into the main steam lines, and the margin to dryer skirt uncovery.
The evaluation concluded that the performance of the steam dryer and separator remains acceptable (e.g., moisture content _ 0.1 weight %) in the MELLLA region. Therefore, no modifications are needed to the steam dryer and separator.
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NEDO-33089 11.0 TESTING Required pre-operational tests (i.e., APRM and recirculation system flow calibrations) will be performed in preparation for operation at the MELLLA conditions with the ARTS improvements. Routine measurements of reactor parameters (e.g., APLHGR, LHGR, MAPLHGR, MLHGR, and MCPR) will be taken within a lower power test condition in the MELLLA region. Core thermal power and fuel thermal margin will be calculated using accepted methods to ensure current licensing and operational practice are maintained.
Measured parameters and calculated core thermal power and fuel thermal margin will be utilized to project those values at the CLTP test conditions. The core performance parameters will be confirmed to be within limits to ensure a careful monitored approach to CLTP in the MELLLA region.
Initial MELLLA testing will be performed in Test Condition A (i.e., between the 50% of RCF line and the core flow line that results in 90% of CLTP on the MELLLA Boundary on a flow control line (FCL) within 5% of the MELLLA Boundary). Power increase beyond Test Condition A will be along this constant FCL to Test Condition B (i.e., between 95% and 100%
of CLTP within 5% of the MELLLA Boundary).
The APRMs will be calibrated prior to MELLLA implementation. The APRM flow-biased scram and rod block setpoints will be calibrated consistent with the ARTS/MELLLA implementation and all APRM trips and alarms will be tested. The flow-biased setpoints of the RBM will also be confirmed.
Acceptable plant performance in the MELLLA power-flow range will be confirmed by inducing small flow changes through the recirculation flow control system. Control system changes are not expected to be required for MELLLA operation, with the possible exception of tuning following evaluation of testing. Subsequently, the recirculation system flow instrumentation calibration will be confirmed within Test Condition B.
Steam separator and dryer performance will be evaluated by measuring the main steam line moisture content. The evaluation will be conducted within Test Condition B. Other test condition power/flow operating points may be tested as deemed appropriate prior to the Test Condition B test to demonstrate the test methodology or to confirm that acceptable steam moisture content at limiting operating conditions is achieved before MELLLA implementation.
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12.0 REFERENCES
- 1. GNF 0000-0006-1823-SRLR, "Supplemental Reload Licensing Report for Vermont Yankee Nuclear Power Station, Reload 22, Cycle 23," Revision 0. October, 2002.
- 2. NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.
- 3. NEDC-32339P-A, Supplement 2, Revision 1, "Licensing Topical Report, Reactor Stability Long-Term Solution: Enhanced Option I-A Solution Design," April 1998.
- 4. NEDE-2401 1-P-A-14, "General Electric Standard Application for Reactor Fuel,"
Revision 14. June, 2000.
- 5. NEDO-31960-A and NEDO-31960-A Supplement 1, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.
- 6. NEDO-32992P-A, "ODYSY Application for Stability Licensing Calculations," July 200 1.
- 7. MFN 01-046, J. S. Post (GE) to Document Control Desk, USNRC, "Stability Reload Licensing Calculations Using Generic DIVOM Curve," August 31, 2001.
- 8. NEDC-32814P, "VYNPS SAFER/GESTR-LOCA Analysis," March 1998.
- 9. NEDE-23785P-A, Vol. III, Supplement 1, Revision 1, "GESTR-LOCA and SAFER Models for Evaluation of Loss-of-Coolant Accident Volume I11,Supplement 1, Additional Information for Upper Bound PCT Calculation," March 2002.
- 10. NEDO-20566A, "General Electric Model for LOCA Analysis in Accordance with 10 CFR 50 Appendix K," September 1986.
- 11. Calculation VYC-2135, Revision 1, "GE Containment System Response Analysis."
September 2000.
- 12. NEDC-32791 P, "Vermont Yankee Nuclear Power Station Increased Core Flow Analysis,"
February 1999.
- 13. NEDC-24154P-A, "Qualification of the One Dimensional Core Transient Model (ODYN) for Boiling Water Reactors (Supplement 1 - Volume 4)", February 2000.
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