ML20116D792
| ML20116D792 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 12/31/1994 |
| From: | Kevin Folk, Hunt B, Kanellopoulos SOUTHERN NUCLEAR OPERATING CO. |
| To: | |
| Shared Package | |
| ML19311C144 | List: |
| References | |
| SNCH-9501, NUDOCS 9608020325 | |
| Download: ML20116D792 (75) | |
Text
_ _-
l zociooore 2 gg_gg BWR STEADY STATE AND TRANSIENT ANALYSIS METHODS BENCHMARKING LICENSING TOPICAL REPORT A
Southern Nuclear Operating Company a subsidiary of The Southern Company
- MS1888R328?8321~
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SOUTHERN NUCLEAR OPERATING COMPANY BWR STEADY STATE AND TRANSIENT ANALYSIS METHODS BENCHMARKING LICENSING TOPICAL REPORT BWR CORE ANALYSIS NUCLEAR FUEL SOUTHERN NUCLEAR OPERATING COMPANY DECEMBER 1994 PREFARED BY:
M
/J11 T.C.' Kanellbp6g6s, Project Engineer BWR Core Analysis APPROVED BY:
K.S. Folk, Manager BWR Core Analysis
)
APPROVED BY:
B.E. Hunt, Maniger Nuclear Fuel The contents of this document are believed to be accurate and complete to the best of Southern Nuclear Operating Company's (SNC) knowledge and information. This document is authorized for use by SNC and/or the appropriate divisions of the U.S. Nuclear Regulatory Commission for review purposes.
The officers, directors, and employees of SNC assume no liability with regard to the unauthorized use of this report.
t Southern Nuclear Opera ting Company P.
O.
Box 1295 Birmingham, AL 35201
ABSTRACT This topical report presents Southern Nuclear Operating Company's (SNC's) benchmarking analyses of General Electric Company's (GE's) steady-state and transient analysis codes and
- methods, and comparisons to GE prepared reload licensing documents. The report demonstrates SNC's qualification and ability to perform steady-state and transient design and safety analysis calculations for licensing applications.
General Electric's steady-state and transient analysis codes were previously reviewed and approved by the Nuclear Regulatory Commission. The procedures used by SNC in performing the licensing calculations meet the requirements set forth in NEDE-240ll-P-A,
" General Electric Standard Application for Reactor Fuel," (GESTAR II). The benchmarking results compare favorably to the original GE reload licensing calculations, and meet the criteria described in NEDO-32362.
_m_
SOUTHERN NUCLEAR OPERATING COMPANY BWR STEADY STATE AND TRANSIENT ANALYSIS METHODS BENCHMARKING LICENSING TOPICAL REPORT TABLE OF CONTENTS Page TABLE OF CONTENTS i
List of Tables ii List of figures iii i
1.0 INTRODUCTION
1 i
1.1 Furpose and Scope 1
1.2 Steady-State Methodology 2
1.3 Model Application (Steady-State) 2 y
1.4 Model Qualification overview (Steady-State) 3 1.5 Reload Transient Analysis (RTA) Methodology 3
i" 1.6 Model Application (RTA) 4 1.7 Model Qualification overview (RTA) 4 1.8 Thermal Limits Evaluation 4
2.0 MODEL DESCRIPTION 5
2.1 Lattice Physics 5
2.2 Core Simulator 6
2.3 Plant and Transient Model Description 9
3.0 SNC METHODS EXPERIENCE 11 J
4.0 SNC QUALIFICATION...................................
12 4.1 Lattice Physics Calculations 12 4.2 Core Steady-State Physics Calculations 24 4.3 Transient Analysis 35 5.O GETAB ANALYSIS OF HATCH 1 CYCLE 16 EVENTS 61 S.1 General Discussion 61 5.2 Getab Results 61 5.3 Summary of GETAB Calculations 65 6.0 SNC QUALITY ASSURANCE PROGRAM 66 4
7.0 CONCLUSION
S 67
8.0 REFERENCES
68 i
i
TABLE OF CONTENTS (Continued)
LIST OF TABLES Table No.
Title Page 4.1-1 Hot Uncontrolled Kur versus Exposure and Voids 13 4.1-2 Hot Uncontrolled L;F versus Exposure and voids 14 4.1-3 Hot Controlled Ka, versus Exposure and voids 15 4.1-4 Hot Controlled LPF versus Exposure and voids 16 4.1-5 Cold Uncontrolled Kur versus Exposure and Voids 17 4.1-6 Cold Uncontrolled LPF versus Exposure and voids 18 4.1-7 Cold Controlled Kur versus Exposure and Voids 19 4.1-8 Cold Controlled LPF versus Exposure and voids 20 4.1-9 Hot Uncontrolled Pin LPF at 12.5 GWd/st-0.40 VH 21 4.1-10 Hot Controlled Pin LPF at 12.5 GWd/st-0.40 VH 22 4.1-11 Rotated Bundle R-Factor Comparison 23 4.2-1 Hatch-1 Cycle 16 Unit Parameters and Reactor Rated Conditions 26 4.2-2 Hatch-1 Cycle 16 Hot Critical Eigenvalues 27 4.2-3 Hatch-1 Cycle 16 Thermal Limits Comparison MCPR 28 4.2-4 Hatch-1 Cycle 16 Thermal Limits Comparison MAPRAT 29 4.2-5 Hatch-1 Cycle 16 Thermal Limits Comparison MFLPD 30 4.2-6 Hatch-1 Cycle 16 Cold Shutdown Margin Comparison 31 4.2-7 Hatch-1 Cycle 16 Standby Liquid Control System Comparison 32 4.2-8 Hatch-1 Cycle 16 Rod Withdrawal Error Comparison 33 4.3-3 Hatch-1 Cycle 16 Turbine Trip Without Bypass (TTNBP) at 100%/100% Power / Flow - Peak value Comparisons for Neutron Flux, Heat Flux, Vessel Pressure
~
41 4.3-4 Hatch *. Cycle 16 Load Rejection Without Bypass (LRNBP) at 100%/100% Power / Flow - Peak value Comparisons for Neutron Flux, Heat flux, Vessel Pressure 47 4.3-5 Hatch-1 Cycle 16 Feedwater Controller Failure (FWCF) Maximum Demand at 100%/100% Power / Flow Peak value Comparisons for Neutron Flux, Heat Flux Vessel Pressure 53 4.3-6 Hatch-1 Cycle 16 MSIV Closure With Flux Scram (MSIVF) at 100%/100% Power / Flow - Peak value Comparisons for Neutron Flux, Vessel Pressure 59 5.2-1 Hatch-1 Cycle 16 TTNBP Delta-CPR Comparison 62 5.2-2 Hatch-1 Cycle 16 LRNBP Delta-CPR Comparison 63 5.2-3 Hatch-1 Cycle 16 FWCF Delta-CPR Comparison 64 11 I
l
TABLE OF CONTENTS (Continued)
LIST OF FIGURES Fiqure No.
Title Page 2.0-1 GE Steady-State Analysis Code Sequence 8
2.0-2 GE Transient Analysis Code Sequence 10 4.2-1 Hatch-1 Cycle 16 Core Configuration of the Reference Loading Pattern 34 4.3-3 Hatch-1 Cycle 16 Turbine Trip Without Bypass at 100%/100% Power / Flow - Parameter Response 37 4.3-4 Hatch-1 Cycle 16 Load Rejection Without Bypass at 100%/100% Power / Flow - Parameter Response 43 4.3-5 Hatch-1 Cycle 16 Feedwater Controller Failure Maximum Demand at 100%/100% Power / Flow Parameter Response 49 4.3-6 Hatch-1 Cycle 16 MSIV Closure With Flux Scram at 100%/100% Power / Flow - Parameter Response 55 iii
1.0 INTRODUCTION
1.1 Purpose and Scope
The purpose of this licensing topical report (LTR) is to meet the requirements of the Nuclear Regulatory Commission's (NRC's) Generic Letter 83-11 by demonstrating Southern Nuclear Operating Company's (SNC's) ability to apply the' steady-state analysis model of the reactor fuel and core, and the transient analysis model of the nuclear steam supply system (NSSS) in performing design and safety analyses in support of licensing actions; e.g.
reload designs and applications, Technical Specification amendments, Supplemental Reload Licensing Reports (SRLRs), Core Operating Limits Reports (COLRs), and Final Safety Analysis Report (FSAR) updates for the Edwin I.
Hatch Nuclear Plant Units 1 and 2.
The analyses were performed using the NRC-approved computer methods developed by General Electric Company (GE).
The application procedures used by SNC in the analyses of cycle-specific licensing events meet the requirements specified in NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel"'" (CESTAR II), which i
has been reviewed and approved by the NRC.
The basic steady-state and transient modeling for Plant Hatch units 1 and 2 was developed by GE using their engineering computer programs (ECP)(*'3#. SNC has obtained all GE computer programs and databases associated with steady-state and transient analyses to use in performing in-house design, licensing, and operation support calculations for both Hatch units. SNC's ability to apply the existing GE models with appropriate plant and cycle-specific information and perform plant and cycle-specific stead-state and transient analyses is demonstrated through:
Modeling the neutronic characteristics of a GE9B (8X8NB) bundle; e.g.,
lattice K-infinity (Kw), local peaking factor (LPF) and comparing the data to GE's results.
Calculating the value of several licensing parameters contained in the Hatch-1 Cycle 16's COLR or SRLR; e.g.
cold shutdown margin and standby liquid control, and comparing to GE's results for the same cycle.
Independently performing the rodded burnup of Hatch 1 Cycle 16 and comparing to GE's results.
Independently performing the Hatch 1 Cycle 16 reload licensing calculations and comparing to GE's results.
This LTR contains the comparisons of steady-state and transient calculations as enumerated above. The transient events represent those typically encountered in reload licensing applications.
Since this LTR is demonstrating SNC's ability to use GE methods by comparing analysis results to vendor calculations, the comparison approach discussed in Criterion 4 of NEDO-32362"" will be utilized when appropriate. Specifically, the criterion 4 parameters (shown below) calculated by SNC will be compared to GE's results and the deviations will be shown to be within the following acceptance criteria:
Parameter Acceptable Deviation 3-D Analysis -
Critical Eigenvalue Delta-K - 1 0.001 Reactivity Margins Delta-K - 1 0.001 Thermal Margins 1 2.0%
- Delta-CPR 1 0.01
- Peak Vessel Pressure 15 psi
- Peak Power (Heat flux) 1 1.0%
- Neutron Flux 1 20.0%
1.2 Steady-State Methodolocry Steady-state analysis work in support of licensing actions is performed with technical procedures meeting the requirements of NEDE-240ll-P-A (GESTAR II) and NRC-approved GE computer methods.
The engineering computer programs include the NRC-approved lattice physics ECP (TGBLA), steady-state 3D core simulator ECP (PANACEA),
and various other linking codes. These ECPs/ methodologies are currently used by GE to perform design and licensing analyses for Plant Hatch Units 1 and 2.
As stated previously, the purpose of this LTR is to demonstrate SNC's ability to apply the subject ECPs to licensing calculations, 3
not to requalify GE models.
1.3 Model Application (Steady-State)
Southern Nuclear will use the NRC-approved GE steady-state methods to perform safety-related analyses in support of reload design and licensing actions; i.e.,
reload applications, Technical Specifications amendments, COLRs, SRLRs, and FSAR updates. The governing process methodology that will be employed by SNC is described in NEDE-240ll-P-A (GESTAR II).
~
1.4 Model Qualification Overview (Steady-State)
TGBLA and PANACEA have undergone extensive qualification by GE"*"
and have obtained favorable safety evaluations from the NRC for application or reference by GE in licensing submittals for their customer's plants. These qualifications consisted of GE comparisons between PANACEA /TGBLA calculated results and actual plant data (e.g., plant instrument data or gamma scan results).
To demonstrate proficiency in applying the steady-state codes SNC prepared lattice physics models for a recently utilized bundle, and the Hatch 1 Cycle 16 reference loading pattern. These models and resulting calculated parameters were compared to GE calculations.
These comparisons support the conclusion that SNC is qualified and proficient in the use of GE methods for licensing analyses.
Historically, SNC has performed steady-state analyses in parallel with GE. By incorporating the same calculational approach and using the same codes and methods, documentation of SNC's proficiency in using the codes may, therefore, be simplified by direct comparison to vendor calculations.
1.5 Reload Transient Analysis (RTA) Methodology Reload transient analysis work in support of licensing actions is performed with technical procedures meeting the requirements of NEDE-240ll-P-A (GESTAR II) and NRC-approved GE computer methods.
These methods generate initial conditions (e.g.,
thermal-hydraulic data, collapsed cross sections) that, together with the generic plant model form the plant-and cycle-specific transient analysis model used by ODYN. The most important engineering codes used in the licensing calculations are:
TGBLA
- Lattice physics methods PANACEA
- Steady-state 3D core simulator ODYN
- 1D systems transient ECP TASC
- Calculation of delta-critical power ratio (DCPR)
GETAB
- Delta-CPR calculation methodology These ECPs/ methodologies are currently used by GE to perform safety analyses for Plant Hatch Units 1 and 2.
s '
1.6 Model Application (RTA)
Southern Nuclear will use NRC-Lpproved GE computer methods to perform safety-related calculations in support of licensing actions (i.e., reload applications, Technical Specifications amendments, COLRs, SRLRs, and FSAR updates) or to evaluate actual operational transient events.
The transients selected for inclusion in this LTR are the same as the ones GE analyzed for the Hatch 1 Cycle 16 reload licensing campaign. They include:
Feedwater controller failure Turbine trip without bypass Load rejection without bypass Main steam isolation valve closure Rod withdrawal error (PANACEA) 1.7 Model Qualification Overview (RTA)
ODYN has undergone extensive qualification by GE, and has obtained a f avorable safety evaluation by the NRC for applications to a wide range of operational transients "). The qualification consisted of comparisons of the ODYN results to actual plant data and an assessment of modeling uncertainties and their effect on thermal limit calculations based on ODYN results. The basic ODYN model, as developed by GE, will be unchanged at SNC, except for the inclusion of necessary plant-and cycle-specific information. To demonstrate proficiency in applying the RTA codes, SNC analyzed the transient events listed in section 1.6 and compared results to those obtained by GE.
These comparisons support the conclusion that SNC is qualified and proficient in the use of GE methods for licensing analyses.
1.8 Thermal Limits Evaluation The transient response to an event (determined by ODYN and presented in section 4.0) is used to assess thermal limits performance using the GEXL correlation"3 The variation with time of some key core parameters such as power, flow, pressure, etc.,
form the boundary conditions for the hot rod calculation and the determination of the critical power ratio. The results of thermal limit calculations for Hatch 1 Cycle 16, and their comparison to GE's results are presented in section 5.0.
/ b
I 2.0 MODEL DESCRIPTION In this section a brief description of the steady-state and system transient models will be presented.
Steady-state analyses are typically separated into two distir.ct phases:
Individual lattice and bundle modeling for each fuel type Core modeling utilizing the lattice and bundle models, as well as appropriate core thermal hydraulic data.
Transient analyses model the nuclear steam supply system using a time dependent coupled neutronics and thermal hydraulics code. The main codes for steady-state (TGBLA and PANACEA) and transient (ODYN) analyses will be briefly presented below. More detailed discussions of the models and codes are contained in GE's topical reports in references 2, 3 and 4.
Figures 2.0-1 and 2.0-2 show a simplified diagram of ECP interactions for the steady-state and transient codes respectively.
2.1 Lattice Physics The GE lattice physics ECP is TGBLA. The purpose of TGBLA is to provide the 3D steady-state simulator code with diffusion parameters representing the neutronic behavior of a lattice under a variety of expected operating conditions.
TGBLA is a two dimensional diffusion theory based model which assumes an infinite and periodic lattice in the traverse direction, and an infinite and uniform lattice in the axial direction. The lattice dimensional parameters, as well as enrichment and burnable poison distributions, are input by the user. Nuclear cross sections based upon ENDF/B-V are input from a controlled data file. The solution technique in TGBLA is broken into fine and coarse mesh modeling and incorporates an iterative transport and diffusion theory approach.
The ultimate goal of TGBLA is to produce few group lattice average diffusion parameters suitable for use in the 3D core simulator.
This goal is accomplished by determining the few group coarse mesh model (pin-by-pin) diffusion parameters such that the lattice wide reaction rates are conserved. However, to determine the nodal diffusion parameters, the multi group cross sections are determined by collapsing the fine group cross sections. The fine group thermal fluxes required to collapse the fine group cross sections are determined by the leakage-dependent integral transport i
equations.
The leakage term is determined by the coarse mesh diffusion solution. Therefore, the fine group thermal integral transport calculation is coupled to the few group dif fusion theory calculation, and an iterative scheme is used to determine the converged thermal parameters. l l
i
l l
I To determine the fast group and epithermal group cross sections, a
simpler integral transport model is employed on an equivalent one-dimensional bundle where the hydrogen-to-uranium atom ratio is conserved. This method is sufficient because the intermediate and fast fluxes are essentially flat across the lattice.
In addition to the lattice average diffusion parameters, TGBLA includes a nuclide depletion calculation based upon nuclide chains, including both the fissionable material and the fission products.
Both regular fuel and burnable poison rods are included in the depletion calculation.
i The above calculations are performed at a predefined set of exposure points and void fractions for hot controlled and uncontrolled conditions, cold controlled and uncontrolled
{
conditions, instantaneous moderator density change from a base moderator density history, and fuel temperature change (Doppler).
The diffusion parameters for this collection of calculations are available for use in the 3D core simulator code.
2.2 Core Simulator The GE core simulator ECP is PANACEA. The purpose of PANACEA is to produce reactivity, power distribution, and thermal performance information, as functions of design and operational variables, in core design and operational calculations. PANACEA has an extensive set of analysis options and provides the engineer with information required for BWR core analysis.
PANACEA is a
steady-state, three-dimensional, coupled nuclear / thermal-hydraulic, diffusion theory based model.
The diffusion parameters are provided by the lattice physics code and parametrically fit as a function of exposure, control state, moderator density, moderator density history, and fuel temperature.
The core operating data (statepoint) such as core thermal power, core flow, bypass flow, core pressure,. inlet enthalpy, and control i
rod pattern, are also provided as input.
The ultimate goal of PANACEA is to provide the 3D nodal power distribution for a given core statepoint. PANACEA uses collapsed three group dif fusion parameters and core statepoint data to solve the one-group diffusion equation for the core.
The one-group diffusion equation is adequate because the thermal neutron mean free path in a light water reactor is small, and most diffusion is governed by the fast neutrons. The one group diffusion equation is discretized over the nodal mesh points which are 6 inches on each side. Therefore, each node consists of a 6-inch axial slice of a given fuel bundle. The solution methodology involves iterations on the one-group discretized diffusion equation, such that the flux distribution and effective multiplication factor K
are en determined.
The diffusion parameters are sensitive to the local moderator density, which is governed by the power distribution and hence, the flux solution. Therefore, the flux solution is coupled to the thermal-hydraulic solution. To ascertain the final solution for the core statepoint, a consistent power / flux and void distribution must be determined by performing an outer void loop af ter convergence of the inner flux loop and its associated power distribution. The flow distribution required to determine the void distribution is established by balancing the pressure drop across a
set of characteristic channels representing the different types of mechanical bundle designs.
Individual bundle flows are then obtained by interpolating from the characteristic channel flow based on radial and axial power shapes. Once the bundle flow is known, a nodal energy balance is used to determine nodal quality.
A void quality relation is used to determine the nodal void distribution. The converged solution results in flux, power, and void distributions that are consistent with respect to the coupled nuclear and thermal-hydraulic calculations.
Based on the power and flow distributions, thermal limits --
average planar linear heat generation rate (APLHGR) and minimum critical power ratio (MCPR) -- may be evaluated for conformance to design criteria. Additionally, a calculated traversing incore probe (TIP) response is determined for the instrument channel locations.
A sequential set of steady-state power distribution calculations may be performed at various exposure points throughout a cycle, so core reactivity behavior can be evaluated.
Other reactivity parameters which may be derived from PANACEA results, such as
- Doppler, void coefficient, and delayed neutron fraction, are provided to support transient calculations.
FIGURE 2.0-1 GR STEADY STATE ANALYSIS CODE SEQUENCE Product Line Mechanical Information Pin Enrichment GadDistribution !
TGBLA i X.Section Libraries 1r Core Statepcint i Thermal / Hydraulic *
...................3 o.1, PANACEA o,1, 4
= Input Process
= Computer Code
I 2.3 Plant and Transient Model Description The Edwin I. Hatch Nuclear Plant is a two unit 2436 Mwth, BWR/4, jet pump nuclear steam supply system (NSSS) utilizing a motor-generator (M-G) set for reactor flow control, an electro-hydraulic control (EHC) system for pressure control, and a three-element feedwater control system for level control. The two plant Hatch units have almost identical features, except for a few parameter i
and geometrical differences.
The transient behavior of the NSSS is modeled with ODYN. The system j
model encompasses the reactor pressure
- vessel, jet
- pumps, recirculation pumps and associated piping, feedwater system, and
)
steam lines. The ODYN model of the core is one dimensional and is coupled to the recirculation and major system control models. The vessel consists of the coupled upper and lower plenums, steam i
separator, vessel dome, bulkwater, and core models.
l The Hatch specific model is based on as-built drawings, vendor specifications on component performance, and certain plant-specific parameters. Plant-and cycle-specific data important for updating f
the ODYN model are defined each cycle through the use of the Operating Parameters for Licensing (OPL-3) form.
In ODYN the reactor core is represented in one axial dimension by one averagt channel for both neutronic and thermal-hydraulic calculatiers. The nuclear properties of that channel are generated by radially collapsing the nuclear properties of the 3-D core model at discreet axial core elevations.
ODYN calculates the time dependent axial flux distribution, assuming six groups of delayed neutrons and the one-group time dependent diffusion equation.
a 2
The hydraulic modeling of the channel accounts for single phase liquid, subcooled boiling, and bulk boiling. A five-equation model a
for mass, energy, and momentum balances is used to model liquid and j
vapor mass conservation and mixture momentum conservation.
i Heat transfer from the fuel is modeled assuming a cylindrical fuel pin with gap conductances that vary axially. The time dependent l
heat conduction equation is solved in the radial direction assuming no heat conduction axially.
a
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l a
4 I
4 i
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FIGURE 2.0-2 GE TRANSIENT ANALYSIS CODE SEQUENCE I' #^ #
LATTICEPHYSICS 3D BWR SIMULATOR STMY4 TATE TGBLA PANACEA T/H ISCOR 3D to 1D CROSS 4ECTION COLLAPSE CRNC i
i PLANTKYCLE 1D TRANSIENT DA TA ODYN DELTA-CPR TASCJGETAB w
3.0 SNC METHODS EXPERIENCE Methodology transfer from GE to SNC began in the mid 1980s in phases that started with the steady-state methodology, followed by the transient and GETAB methodology. Prior to this time SNC used in house models of CASMO/ SIMULATE for core follow and design, and RETRAN for analyzing FSAR and startup events.
Af ter GE methods became available, SNC personnel received extensive training from GE at their design facilities on the use of both steady-state and transient methods. This training involved actual design and licensing analyses for plant Hatch under GE's technical direction and procedures.
Southern Nuclear personnel have utilized GE methods to perform core management, parallel design and licensing
- work, and other ovaluation of operation to support plant Hatch.
l
l GE COMPANY PROPRIETARY 4.0 SNC QUALIFICATION To demonstrate technical competence to " set up an input deck, execute a code, and properly interpret the results", as discussed I
in Generic Letter 83-11, SNC prepared a representative set of I
calculations using the GE steady-state and transient analysis ECPs.
These calculations involve lattice physics calculations for the Hatch-1 Cycle 15 fresh bundle, the Hatch-1 Cycle 16 steady-state I
calculations associated with the reference loading pattern, and transient and GETAB analyses for the Hatch-1 Cycle 16 reload core.
The results of these analyses and their comparison to corresponding i
GE analyses presented in this report are considered the basis of SNC qualification to use GE methods, consistent with NEDO-32362.
4.1 Lattice Physics Calculations 4.1.1 Lattice Calculations For this LTR, SNC modeled the reload GE9B bundle used in Hatch-1 Cycle 15:
1 The modeling included all necessary steps to generate cross I
sections for the core simulator given the pin-by-pin U-235 and burnable poison enrichments. A standard set of calculations were performed to model the effects of void, void history, Doppler, control, cold conditions, and lattice burnup. Because the same I
codes, computer systems, and libraries were used, SNC's and GE's results are identical. Tables 4.1-1 through 4.1-10 show the SNC and GE comparisons for this bundle.
i 4.1.2 Rotated Bundle R-factors are used in the determination of a bundle's critical I
power ratio (CPR) performance. Rotated R-factors accordingly, are used to determine delta-CPR for the Fuel Loading Error rotated bundle event.
The rotated bundle configuration was modeled with TGBLA and the SNC results together with those of GE's, are shown in Table 4.1-11.
4.1.3 Summary Based on the presented comparisons between SNC and GE calculated lattice physics parameters, SNC has demonstrated proficiency in using the lattice physics ECPs to accurately model GE fuel. Because results were identical to GE's, no acceptable deviation criteria are necessary.
I u
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..... _....... -.. -. - ~..... _.. -..
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GE COMPANY PROPRIETARY TABLE 4.1-1 i
HOT UNCONTROL'.2D Ka, VERSUS EXPOSURE AND VOIDS l
t i
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GE COMPANY PROPRIETARY TABLE 4.1-2 HOT UNCONTROLLED LPF VERSUS EXPOSURE AND VOIDS
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GE COMPANY PROPRIETARY TABLE 4.1-3 HOT CONTROLLED Ka, VERSUS EXPOSURE AND VOIDS
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GE COMPANY PROPRIETARY TABLE 4.1-4 HOT CONTROLLED LPF VERSUS EXPOSURE AND VOIDS i -..
GE COMPANY PROPRIETARY
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COLD UNCONTROLLED Kins VERSUS EXPOSURE AND VOIng 1
l 1
l
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l l i
GE COMPANY PROPRIETARY TABLE 4.1-6 COLD UNCONTROLLED LPF VERSUS EXPOSURE AND VOIDS _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ - _ _ - _ - _
a d
4 i
I GE COMPANY PROPRIETARY TABLE 4.1-7 COLD CONTROLLED Ka, VERSUS EXPOSURE AND VOIDS 4
4 4
4 1
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e 4
1 I
5 k
i 4
b 1
i t
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4
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5 ii I
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1
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GE COMPANY PROPRIETARY a
TABLE 4.1-8 I
COLD CONTROLLED LPF VERSUS EXPOSURE AND VOIDS l
i 4
4 1
1 1
i s
2 i
1 1
4 h
4 a
4 7
1 l
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GE COMPANY PROPRIETARY TABLE 4.1-9 HOT UNCONTROLLED PIN LPFs
(*'
EXPOSURE = 12.5 GWD/ST VOID HISTORY = 0.40
/
a.
SNC values are in bold - _ _
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._.. _ _ _ _ _ ~.. _... _ _ _.. _. _
i i
i 1
1 I
i j
GE COMPANY PROPRIET1.RY 1
1 TABLE 4.1-10 HOT CONTROLLED PIN LPFs
(*3 1
i i
8 i
i i
1 1
1 1l' i
t i
- i..
t
.r l
t ij l
I i
l 4
l.
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1 4
1 1
i i
i l
1 i
t e
- a. SNC values are in bold l
GE COMPANY PROPRIETARY TABLE 4.1-11 ROTATED BUNDLE R-FACTOR COMPARISON d
4 5
4.2 Core Steady-State Physics Calculations 4.2.1 Introduction For this
- LTR, SNC modeled the Hatch-1 Cycle 16 burn of the reference loading pattern (RLP) to emulate GE's work. The unit's rated conditions and type are listed in Table 4.2-1. The starting point for SNC's Hatch-1 Cycle 16 calculations consisted of an SNC generated simulator wrapup containing the projected end-of-cycle 15 conditions. These conditions were obtained with a rodded burn of Cycle 15, based on SNC generated exposure accounting from the beginning of the cycle. The projected rodded burn of cycle 15 (from 7790.5 Mwd /st to EOC) was performed in the same manner as at GE.
The end of cycle 15 was then discharged, and the new core was loaded with fresh fuel and shuffled to obtain a beginning-of-cycle core simulator wrapup that matched GE's reference loading pattern for Hatch-1 Cycle 16, including all the appropriate neutronic and thermal-hydraulic core parameters.
The RLP thus developed was used at SNC for comparison of eigenvalues, thermal limits evaluations, shutdown margin j
evaluation, standby liquid control system evaluation, and rod j
withdrawal error analysis.
j Figure 4.2-1 shows the core configuration of the Hatch-1 Cycle 16 reference loading pattern.
)
4.2.2 Hot Critical Eigenvalues Hot critical eigenvalues were obtained from a rodded burn of the Hatch-1 Cycle 16 RLP. These were compared to those obtained at GE ar.d, as shown on Table 4.2-2, they agree very well, and meet the eigenvalue deviation criteria in NEDO-32362.
4.2.3 Thermal Limits The ability to accurately predict the power distribution in the core is a critical requirement for assurance a given core will operate within the Technical Specifications thermal limits throughout an operating cycle. GE's calculations of Hatch-1 Cycle 16 thermal limits were used as the benchmark. SNC's and GE's calculations of thermal limits are shown on Tables 4.2-3, 4.2-4, and 4.2-5.
The comparisons show excellent agreement between GE's and SNC's calculations, meeting the thermal margin deviation criteria in NEDO-32362.
i 4.2.4 Cold Shutdown Margin An important part of the licensing basis described in GESTAR II includes the determination of cold shutdown margin (CSDM) for a given core configuration. This type of calculation is performed with PANACEA at various points throughout a cycle to determine the minimum cold shutdown margin. Table 4.2-6 shows the comparison between the SNC and GE calculated CSDM for the Hatch-1 Cycle 16 RLP.
The results agree extremely well and meet NEDO-32362 OccrTrable reactivity margins deviations.
4.2.5 Standby Liquid Control System The standby liquid control system (SLCS) provides an independent and alternate means of reactor shutdown by injecting boron into the reactor coolant upon system initiation.
As with CSDM, PANACEA is used to analytically determine the degree of suberiticality provided by SLCS, by calculating the reactor's eigenvalue in the borated, all rods out, cold xenon free state.
Table 4.2-7 displays the comparison between the SNC and GE calculated SLCS shutdown margin, for the Hatch-1 Cycle 16 RLP. As it can be seen in that table, the NEDO-32362 reactivity margin acceptable deviations are met with a large margin.
4.2.6 Rod Withdrawal Error The rod withdrawal error (RWE) event assumes that the operator erroneously selects and withdraws a control rod. The intent of this calculation is to calculate the MCPR response to this event under certain assumptions on the operability of the rod block monitor (RBM), its setpoint to block, and statistical LPRM failures that feed into the RBM.
PANACEA is used for this calculation to determine the MCPR response at different increments of rod withdrawal assuming no RBM blocks.
Table 4.2-8 shows the comparison between the SNC and GE calculated evaluation of the RWE event for the Hatch-1 Cycle 16, and that it meets NEDO-32362 acceptable deviation criteria for delta-CPR.
4.2.7 Summary The calculations and comparisons shown in section 4.2 demonstrate SNC's proficiency in utilizing GE's steady-state methodology to perform licensing calculations. These comparisons include data taken from the Supplemental Reload Licensing Submittal, and other GE sources. The calculations performed at SNC are consistent with GE methodology, and meet all GESTAR II requirements.
./
TABLE 4.2-1 HATCH 1 CYCLE 16 UNIT PARAMETERS AND REACTOR RATED CONDITIONS f
Parameter Value Core Thermal Power (MWth) 2436.0 Core Flow (Mlb/hr) 78.5 Dome Pressure (psia) 1020.0 Inlet Enthalpy (M1b/hr) 523.9 f
BWR Type 4
Lattice Configuration D
No. of Fuel Assemblies 560 No. of Control Rods 137 No. of TIP locations 31 I
j t
[
GE COMPANY PROPRIETARY TABLE 4.2-2 HATCH 1 CYCLE 16 HOT CRITICAL EIGENVALUES
{
{
{
f
{
{
[
l __
GE COMPANY PROPRIETARY TABLE 4.2-3 HATCH - 1 CYCLE 16 THERMAL LIMITS COMPARISON MCPR DIFFERENCE
4 i
J 1
GE COMPANY PROPRIETARY i
i TABLE 4.2-4
)
HATCH - 1 CYCLE 16 THERMAL LIMITS COMPARISON i
i MAPRAT DIFFERENCE i
i e
4 I
l l
I I
4 i
i i
i i. _.,
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GE COMPANY PROPRIETARY TABLE 4.2-5 HATCH - 1 CYCLE 16 THERMAL LIMITS COMPARISON l
MFLPD DIFFERENCE
) i
i i
i 4
4 4
i GF COMPANY PROPRIETARY i
TABLE 4.2-6 j
HATCH
--1 CYCLE 16 i
COLD SHUTDOWN MARGIN COMPARISON 1
i l
r i
i 1
t r
o 6
1 4
k i
4 i
1 i
i 1
s i
t t.
i 4
i I
a i
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i
- J 1
I
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GE COMPANY PROPRIETARY TABLE 4.2-7 HATCH - 1 CYCLE 16 STANDBY LIQUID CONTROL SYSTEM COMPARISON i __
TABLE 4.2-8 HATCH - 1 CYCLE 16 ROD WITHDRAWAL ERROR COMPARISON SNC GE Difference i
i Event Delta-CPR 0.217 0.220
-0.003 i
i I
l i -
V"*Au, GE COMPANY PROPRIETARY FIGURE 4.2-1 HATCH - 1 CYCLE 16 CORE CONFIGURATION OF THE REFERENCE LOADING PATTERN i.
I
4.3 Transient Analysis 4.3.1 Introduction The core flow region that Hatch-1 Cycle 16 is licensed to operate in at 100% power, extends from 87% to 105% of rated core flow. This type of operation is justified by performing limiting transient analyses at the end points of this region i.e.,
extended load line limit (ELLL) for the 100%/87% power-flow point, and increased core flow (ICF) at the 100%/105% power-flow point. In addition, analyses at the 100%/100% power-flow point (STANDARD) were also performed.
The results of the GE analyses for Hatch-1 Cycle 16 are contained in the supplemental Reload Licensing Report") (SRLR) but only the ones for the STANDARD power flow case will be used as the benchmark to demonstrate SNC's qualifications in the use of GE's transient analysis methodology.
The Hatch-1 Cycle 16 specific model for transient calculations was independently developed at SNC based on:
~
1.
The reference loading pattern for that cycle, 2.
Cycle-specific design bases,
- 3. The operating parameters for licensing which establish important plant characteristic data used in transient analyses for reload applications, and, 4.
The independent modeling of Hatch-1 Cycle 16 described in Section 4.2.
In the process of developing this model, NRC-approved methods, and procedures meeting GESTAR II requirements were used. In addition, SNC QA requirements were followed.
I 4.3.2 General Discussion Various transient events starting from dif ferent operating and core exposure conditions, evolve dif f erently in terms of the severity in
{
challenging thermal or overpressurization limits.
From this perspective, the most limiting events for Hatch-1 Cycle 16 are:
Turbine Trip without Bypass (TTNBP)
Load Rejection without Bypass (LRNBP) j Feedwater Controller Failure (FWCF) l Rod Withdrawal Error (RWE)
MSIV Closure with Flux Scram (MSIVF)
Rotated bundle j
It should be noted that some graphical comparisons between SNC and GE results in this section are hard to distinguish because of their closeness which makes them appear as a single line.
j ;
(
4.3.3 Turbine Trip without Bypass The TTNBP pressurization transient assumes that the turbine stop valves close without opening of the bypass valves. A reactor scram signal as well as a recirculation pump trip (RPT) are initiated on
(
stop valve position. A typical reactor response is summarized below:
{
As a result of valve closure, a pressure wave travels through the steam lines to the vessel. There the pressure wave goes through the bulkwater and jet pumps to the core inlet, and through the separators and standpipes to the upper plenum,
[
with a timinc proportional to the sonic speed of the medium in which it travels. This manner of pressure wave propagation and time of arrival to the core at opposite ends is responsible
(
for the initial void collapse, power increase, flow increase, and water level drop. Scram eventually overtakes the positive reactivity from void collapse and the flux rise is followed by
[
a rapid decrease.
Pressure increase is terminated by the L
opening of the SRVs. Core flow increase is terminated by the wave arriving through the separator and is subsequently reduced by RPT.
Figure 4.3-3 shows the SNC/GE comparison of the TTNBP event for the rated power / flow conditions at EOC. Agreement in magnitude and timing is excellent. Table 4.3-3 shows the comparison of maximum heat flux, neutron flux, and peak vessel pressure between SNC and GE. The agreement meets the acceptable deviation criteria for these parameters in NEDO-32362 by a comfortable margin.
[
[
[
{
( :
1
(
FIGURE 4.3-3a s
j HATCH - 1 CYCLE 16 t
TURBINE TRIP WITHOUT BYPASS AT 1001b/100% POWER / FLOW 2.0 4.0 44
[
l j
e
'~
3-3 OO 2.0 e
t v
t 0
- 1.0 0
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o
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(
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[
j-2.0
'> 0.0 - d
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6 8
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2 4
6 8
10 Time (Seconds)
Time (Seconds)
{
20 0.0 W _10
{
1.0 I
1 X
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o o
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E; z
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FIGURE 4.3-3b HATCH 1 CYCLE - 16 TURBINE TRIP WITHOUT BYPASS AT 100%/100% POWER / FLOW 200 150
~1 i
T 1
l 3
1 150 125 g
f 5
I h
8 100 S
100 d e
z
\\
c 50.0 75 0 8
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$0.0 -
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10 Time (Seconds)
Time (Seconds) 400 150 Dv V
o E
^,
a*
n g
100 l
g i
x 3
200
[
x
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e a
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10 GE Time (Seconds)
Time (Seconcs)
\\ i f
l
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l FIGURE 4.3-3d MATCH 1 CYCLE - 16 TURBINE TRIP WITHOUT BYPASS AT 100%/100% POWER / FLOW l
l l
I 1
I i
i 1.0 150 3
V b
LL o
0.8 E
l 100 9
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10 l
Time (Seconds)
Time (Seconds) l i
l i
. J
TABLE 4.3-3 KATCH 1 CYCLE 16 TURBINE TRIT WITHOUT BYPASS AT 100%/100% POWER / PLOW PEAK VALUE COMPARISONS FOR NEUTRON FLUX, HEAT FLUX, VESSEL PRESSURE l
i Parameter SNC GE Difference Peak Neutron Flux (%)
351.87 351.88
-0.01 Peak Heat Flux (%)
116.48 116.45 0.03 Peak Vessel Pressure (psia) 1193.2 1193.2 0.00 I
l
)
j 4.3.3 Load Rejection without Bypass 9
The LRNBP pressurization transient assumes that the turbine control valves (TCVs) close without opening of the bypass valves. Actual reactor scram is initiated by pressure switches attached to the hydraulic oil components associated with the f ast closure mechanism of the TCVs. Load rejection also causes an RPT. A typical reactor response is summarized below:
j The typical reactor response to the LRNBP pressurization transient is almost identical to the TTNBP discussed earlier.
The main dif f erence lies in that TCVs close more rapidly than the turbine stop valves, thereby making the LRNBP event more severe. Otherwise the qualitative response between the two transients is identical.
I Figure 4.3-4 shows the SNC/GE comparison of the LRNBP event for the
)
rated power / flow conditions. Again agreement in magnitude and timing is excellent. Table 4.3-4 shows the comparison of maximum heat flux, neutron flux, and peak vessel pressure between SNC and GE.
Again the agreement is very good and meets the NEDO-32362 criteria for acceptable deviations of these parameters.
I -
FIGURE 4.3-da HATCH 1 CYCLE - 16 LOAD REJECTION WITHOUT BYPASS AT 100%/100% POWER / FLOW i
2.0 4.0 1
l 1.0 3.0 I
i b
'5 0'0 h
W
$_ 2'0 0
O V
f
- 1.0 1.0 g
L S
D S
'5 O -2.0 J
O 0.0
- 3.0 -
- 1.0 -
0 2
4 6
8 10 0
2 4
6 8
10 Time (Seconds)
Time (Seconds) 2.0 0.0 1.0 M -10 f
W I
30.0
-J x
T -20 5
z O
U O
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l 0 - 1.0 E -30 E
(
i;
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4 6
8 10 0
2 4
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10 i
GE Time (Seconds)
Time ($econds)
- i i
J FIGURE 4.3-4b HATCH 1 CYCLE 16 LOAD REJECTION WITHOUT BYPASS AT 100%/100% POWER / FLOW 200 150
[q t
C.
3 1 150 1
125 i
e n
.T C
I 8 100 k
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100 f
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Time (Seconds)
Time (Seconds) 400 150 D
"O3 f
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100
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x
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5 0.0 -
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2 4
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10 GE ice (Seconds)
Time (Seconds)
SNC FIGURE 4.3-4c HATCH 1 CYCLE 16 LOAD REJECTION WITHOUT BYPASS AT 100%/100% POWER / FLOW l
150 150 t
D o
2 O
o C
C N
100 100 g
I I
3 3
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50 0 0 50.0 e
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2 3
t 3
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e 12 0.0 -
0.0 -
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2 4
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10 1
Time (Seconds)
Time (Se,conds) 50.0 100 t
z 3
_E_ 40.07 oe 80.0 l
M M
l I
m 30.0 60.0 S
O t.L
.O< 20.0 e 40.0 I)>
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0.0 -
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4 6
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2 4
6 8
10 GE Time (Seconds)
Time (Seconds)
SNC l
[
FIGURE 4.3-4d
[
HATCH - 1 CYCLE - 16 LOAD REJECTION WITHOUT BYPASS AT 100%/100% POWER / FLOW
[
l
[
[
[
1.0 150 D
[
i 0.8 E
100 l
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.{
j 0.6
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l 3 0.4 E 0.0
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Time (Seconds)
[
[
[
[
as SNC r PC
1 TABLE 4.3-4 1 CYCLE HATCH 16 i
LOAD REJECTION WITHOUT BYPASS AT 100%/100% POWER / FLOW PEAK VALUE COMPARISONS FOR NEUTRON FLUX, HEAT FLUX, VESSEL PRESSURE Parameter SNC GE Difference Peak Neutron Flux (%)
368.63 370.14
-1.51 Peak Heat Flux (%)
117.91 117.91 0.00 Peak Vessel Pressure (psia) 1194.9 1194.9 0.00 i
i
< 1 A
1
l 4.3.5 Feedwater Controller Failure The FWCF pressurization transient assumes that the feedwater flow controller fails and feedwater flow increases instantaneously to the maximum runout capacity of the feedwater pumps. As a result, water level rises suf ficiently to cause feedwater pump and turbine trips, which in turn cause reactor scram and RPT. A typical reactor 1
response is summarized below:
As a result of FWCF, feedwater enters the downcomer at an increasing rate up to the maximum capacity of the feedwater l
pumps. The excess amount of feedwater causes water level to rise, while its temperature, which is lower than the bulkwater it mixes with, causes an increase in inlet subcooling and subsequent increase of the neutron flux level.
Water level and neutron flux increase at an almost linear rate until the water level rises high enough to trip the turbine.
At this point the FWCF event turns into a turbine trip with bypass initiated from a higher power level. From that point on the evolution of the event is similar to the TTNBP discussed earlier.
Figure 4.3-5 shows the SNC/GE comparison of the FWCF event for the rated power / flow conditions. Agreement in magnitude and timing is excellent.
Table 4.3-5 shows the comparison of maximum heat, neutron flux, and peak vessel pressure, between SNC and GE. The agreement is again very good and it meets the a :eptable deviation criteria in NEDO-32362 for these parameters.
i
! t
1 FIGURE 4.3-Sa 16 HATCH 1 CYCLE FREDWATER CONTROLLER FAILURE MAXIMUM DEMAND AT 100%/100% POWER / FLOW i
2.0 4.0 1.0 3.0 g
1 I
l b
)
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'5 0.0 32.0 W
5 i
0*
o
[
1.0 l
- 1.0 k
.?
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8 12 16 20 0
4 8
12 16 20 Time (Seconds)
Time (Seconds) l 2.0 0.0 1.0 M -10' W
I I
l x
30.0 T -20
)
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t o
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4 8
12 16 20 0
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12 16 20 GE Time (Seconds)
Time (Seconds)
SNC l l
l l
I J
J 4Hodg5
,*La = m'O 8
5 yeom. e go*tN Pm e
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ll jlIil' i]i 4Hoa E S.W*UO 9
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o s
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R 8
7 7
8 W
w w
lo lo F
F r
e 86 e g6 t
n a
ib w
r d
u e
T e
F o9 0
o m
c uo O
c
_8 wg. eai-t
=
eai3 8
8 se d
h e
c t
o In 86 R mS 7
S
+
S 86
\\
eS w
ev lo o
F b
k A eo
> eS v
le lo v
V e
L 8o oS f
ie re le t
R a
W o9 S
o m
c mo O
a c
g o.
ye 3al-sk mai3 su.n
,m"
!II llll
1 i
FIGURE 4.3-5d 1 CYCLE - 16 HATCH FERDWATER CONTROLLER FAILURE MARIMUM DEMAND AT 100%/100% POWER / FLOW 150 t
E 100 l%
g l
L l
50.0 4
O L.
E 0.0 o
u
- - 50 mm 0>
-100-0 4
8 12 16 20 Time (Seconds) 200 50.0 i
3 m
o 3
E o 180 40.0 x
-o 1 160 30.0
[b I
M
_oo l
O D 140 3 20.0 h
2
~ 120 10.0 0
0 0 100 l
0.0 -
0 4
6 12 16 20 0
4 8
12 16 20 GE Time (Seconds)
Time (Seconds)
SNC TABLE 4.3-5 16 1 CYCLE HATCH FREDWATER CONTROLLER FAILURE MAXIMUM DEMAND AT 100%/100% POWER / FLOW PEAK VALUE COMPARISONS FOR NEUTRON FLUX, HEAT FLUX, VESSEL PRESSURE Paranieter SNC GE Difference Peak Neutron Flux (%)
229.72 229.67 0.05 Peak Heat Flux (%)
115.29 115.28 0.01 Peak Vessel Pressure (psia) 1166.8 1166.8 0.00 4.3.6 MSIV closure with Flux Scram The MSIV closure with flux scram transient assumes all MSIV's close without the benefit of a valve position scram; instead, scram is initiated by the high APRM neutron flux. RPT occurs on high pressure to. Limit the severity of the event. A typical reactor response is summarized below:
As a result of the MSIV closure, a pressure front reaches the vessel and travels to the core along two separate paths similar to the TTNBP transient discussed earlier. Due to the timing of the pressure wave in reaching the bottom and top of the core, there is an initial increase in core flow and neutron flux while water level decreases. Core flow remains above its initial level until the time of RPT and opening of the SRVs. Neutron flux reaches a first maximum due to positive void reactivity, and then it turns around due to scram, fuel heat-up, and reduction in the positive void reactivity introduction. Continuing pressurization (and void collapse) results in a second flux maximum terminated by scram. Pressure rise is terminated by the opening of the SRVs.
Figure 4.3-6 shows the SNC/GE comparison of the MSIV event for the rated power / flow conditions.
Agreement in magnitude and timing is excellent, and it meets NEDO-32362 criteria for acceptable deviations - peak vessel pressure-as shown on Table 4.3-6.
I 4
m l J
1 l
1 FIGURE 4.3-6a i
1 CYCLE - 16 HATCH MSIV CLOSURE WITH FLUX SCRAM AT 100%/100% POWER / FLOW 2.0 4.0 1.0 3.0 g
l I
x
/
=
g
.0 2
j n 8
o
- 1 ~0 1.0 A
e
-g Nfw a
.2
]
a O
[-2.0
> 0.0
- 3.0 -
- 1.0 -
0 2
4 6
8 10 0
2 4
6 8
10 Time (Seconds)
Time (Seconds) j 2.0 0.0 1.0 M -10 I
A 1
30.0 3 -20 g
5 z
o O
O I
O O
O -1.0 1 - 30 E
(
i 1
E
.o
)
O t
Z o
- 2.0 y) -40 g
h
- l
- 3.0 -
(i,
0 2
4 6
8 10 0
2 4
6 8
10 Time (Seconds)
Time (Seconds)
l i
I FIGURE 4.3-6b i
HATCH - 1 CYCLE 16 MSIV CLOSURE WITH FLUX SCRAM AT 100%/100% POWER / PLOW i
200 150 U
0 I 150 E
125 D
bt i'
.T C
1 100 l
O i
m E
M I
5 a.
C I
50.0 75.0 e
{
f O
I 0.0 -
50.0-0 2
4 6
8 10 4
6 8
io Time (Seconds)
Time (Seconds) 500 150 to I
D o
- 400
-l C
f 0
o g
e
"?
gt M
100 T
l X
3 X
C j
2k 200 I 50.0 f
J [/
i a la c
Z Y)
I 0.0 I
0.0 -
0 2
4 6
8 10 0
2 4
6 8
10
}
Time (Seconds)
FIGURE 4.3-6c HATCH - 1 CYCLE 16 MSIV CLOSURE WITH FLUX SCRAM AT 100%/100% POWER / FLOW 150 150 t
o o
E o
o T
100 N
100 1
I 3
3 o
2 C
k 50 I
Y E
.0 o 50.0 5
N 3
g 3'
t 0
e L
\\
0.0 -
0.0 -
0 2
4 6
8 10 0
2 4
6 8
10 Time (Seconds)
Time (Seconds) 50.0 100 "I
en i
i N
_8 m
E N
40.0 e 80.0 1
M O,(/
I 30.0 60.0 V
t a
2
\\
o L
.O< 20.0 40.0 e
D O
\\
l J 10.0 g 20.0 e
.=
~
0 S
0.0 -
0.0 -
0 2
4 6
8 10 0
2 4
6 8
10 Time (Seconds) -
Time (Seconds)
k
FIGURE 4.3-6d 1 CYCLE 16 HATCH MSIV CLOSURE WITH FLUX SCRAM AT 100%/100% POWER / FLOW 1.0 150 3
7 2
E L
o 0.8 E
I 100 M
I o
0.6 50.0 M
3o k
g 0.4 E 0.0 l,
o O
a s
W 0.2 g -50 c.
>s D
0.0 -
-100-0 2
4 6
8 10 0
2 4
6 8
10 Time (Seconds)
Time (Seconds)
GE SNC TABLE 4.3-6 HATCH 1 CYCLE 16 MSIV CLOSURE WITH FLUX SCRAM AT 100%/100% POWER / FLOW PEAK VALUE COMPARISONS FOR NEUTRON FLUX, VESSEL PRESSURE Parameter SNC GE Difference Peak Neutron Flux (%)
468.19 467.69 0.50 Peak Vessel Pressure (psia) 1237.2 1237.2 0.00 4.3.7 Summary of Transient Calculations The calculations and comparisons shown section in 4.3 demonstrate SNC's proficiency in utilizing GE's transient analysis methodology j
to perform licensing calculations. These comparisons include data i
taken from the Supplemental Reload Licensing Submittal, and other GE sources.
The calculations performed at SNC are consistent with GE methodology, and meet the GESTAR II requirements.
As discussed throughout the comparisons, all criteria in NEDO-32362 for acceptable deviations were met.
i I
1 1
I I
I I
I I
I I
i
-eo-
5.0 GETAB ANALYSIS OF HATCH 1 CYCLE 16 EVENTS 5.1 General Discussion For Plant Hatch Units 1 and 2, delta-CPR due to a transient event, modeled with ODYN, is calculated using General Electric's thermal analysis basis (GETAB) methodology which combines the ODYN-calculated time profiles of
- power, flow,
- pressure, core inlet / outlet enthalpy, and bundle specific characteristics (e.g. R-factors) to calculate delta-CPR for an event using the NRC-approved GE critical quality boiling length (GEXL) correlation. Various uncertainties, such as scram speed, are accounted for in the transformation of raw delta-CPRs into the " Option A" and " Option B" delta-CPRs. This methodology is discussed in GESTAR II.
5.2 GETAB Results As discussed
- above, the transient response of key reactor parameters, as calculated by
- ODYN, are used by the GETAB 1
methodology to calculate delta-CPR.
In
- practice, this is accomplished by an iterative process involving several GE computer programs. The comparisons presented in this section reflect results independently obtained by SNC and GE.
Delta-CPRs for the feedwater controller failure, turbine trip, and load rejection transients analyzed in section 4.3, for the rated power / flow conditions are presented in Tables 5.2-1, 5.2-2, and 5.2-3 respectively. As it can be seen, the NEDO-32632 criteria for acceptable delta-CPR deviations were met.
TABLE 5.2-1 HATCH 1 CYCLE 16 TTNBP DELTA-CPR COMPARISON SNC GE Difference Uncorrected Delta-CPR GE7B 0.108 0.110
-0.002 GE9B 0.122 0.123
-0.001 Option A Delta-CPR GE7B 0.160 0.160 0.000 GE9B 0.174 0.175
-0.001 Option B Delta-CPR GE7B 0.120 0.120 0.000 GE9B 0.134 0.135
-0.001 TABLE 5.2-2
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HATCH - 1 CYCLE - 16 LRNBP DELTA CPR COMPARISON i
)
SNC GB Difference Uncorrected Delta-CPR GE7B 0.120 0.119 0.001 i
GE9B 0.134 0.135
-0.001 GE13 0.237 0.240
-0.003 Option A Delta-CPR GE7B 0.17 0.17 0.000 GE9B 0.19 0.19 0.000 GE13 0.28 0.28 0.000 Option B Delta-CPR GE7B 0.13 0.13 0.000 GE9B 0.15 0.15 0.000 GE13 0.25 0.25 0.000 i
l TABLE 5.2-3 HATCH 1 CYCLE - 16 FMCF DELTA-CPR COMPARISON SNC GE Difference Uncorrected Delta-CPR GE7B 0.108 0.107 0.001 GE9B 0.120 0.120 0.000 GE13 0.187 0.190
-0.003 1
Option A Delta-CPR GE7B 0.16 0.16 0.000 GE9B 0.18 0.18 0.000 GE13 0.23 0.23 0.000 Option B Delta-CPR
)
GE7B 0.12 0.12 0.000 GE9B 0.14 0.14 0.000 GE13 0.20 0.20 0.000 i
5.3 Summary of GETAB Calculations The calculations and comparisons shown section 5.2 demonstrate SNC's proficiency in utilizing GE's GETAB analysis methodology to calculate delta-CPRs. These comparisons include data taken from the Supplemental Reload Licensing Submittal, and GE databooks.
The calculations performed at SNC are consistent with GE methodology, and meet all GESTAR II requirements. All delta-CPR calculations comfortably meet the requirements of NEDO-32632.
6.0 SNC QUALITY ASSURANCE PROGRAM In Generic Letter 83-11, the NRC stated,"A large percentage of all errors or discrepancies discovered in safety analyses can be traced to the user rather than to the code itself." SNC addressed this concern in the following manner:
1.
The technical ability of SNC's staff is demonstrated by results contained in this topical report. The conclusive data presented herein confirm SNC's ability to competently apply GE's methodology and codes in reload licensing applications, achieving essentially the same results.
- 2. SNC will perform future safety-related licensing activities in accordance with SNC's quality assurance (QA) program and the technical requirements stated in NEDE-240ll-P-A (GESTAR II). SNC's QA program meets the requirements set forth in 10 CFR 50, Appendix B, and related ANSI standards and regulatory guidelines. The SNC QA program covers technical issues associated with steady-state and transient analyses, as well as documentation, review, approval, and revision of safety-related calculations and designs.
Since GE provides the engineering computer codes for installation on the computer system in executable form only.-
SNC uses GE's compiled coding for steady-state and transient analyses. Once the codes arc installed, a test case is executed and the results compared with GE-supplied results to assess the codes' proper execution on SNC's computer system.
The controlled codes, in conjunction with SNC's technical procedures and controlled input data, are used for the licensing analyses. l
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7.0 CONCLUSION
S General Electric received NRC approval to apply their codes and methodology in support of licensing actions. NRC approval was based on extensive review of GE's modeling and benchmark efforts. GE has since used these approved methods and procedures in preparing reload licensing submittals for the Edwin I.
Hatch Nuclear plant
{
Units 1 and 2.
As stated in Generic Letter 83-11, the NRC " encourages utilities to perform their own safety analyses since it significantly improves the understanding of plant behavior. " In submitting this licensing topical report, SNC meets the requirements of the Generic Letter 83-11 by demonstrating the ability to competently perform reload design and licensing steady-state analyses using approved GE methodology and codes. SNC does not intend to requalify the GE codes and methods, but rather to obtain NRC approval to utilize these methods for licensing calculations.
Based on comparisons between SNC and GE analyses contained in'this report, SNC has successfully demonstrated the ability to apply the GE steady-state and transient methods for reload design and licensing calculations, by meeting the requirements set forth by GE NEDO-32362"8-Upon receiving a favorable safety evaluation report from the NRC, SNC will be qualified to apply GE's steady-state and transient analysis methods, in performing design and safety analyses in support of licensing actions.
SNC plans to begin using this capability in Hatch-1 Cycle 17 which starts-up in the spring of 1996. (
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8.0 REFERENCES
1.
NEDE-24011-P-A-1,0, " General Electric Application for Reactor Fuel," (GESTAR), as supplemented.
2.
" Lattice Physics Methods," General Electric, June 1976.
3.
NEDE-30130-P-A,
" Steady State Nuclear Methods,"
General Electric, April 1, 1985.
4.
REFERENCE 4 INCLUDES THE FOLLOWING:
a.
NEDO-24154-A, " Qualification of the One-Dimensional Core Transient Model for BWRs," Volumes 1 and 2, General Electric, February 1, 1986.
- b. NEDO-24154-P-A, " Qualification of the One-Dimensional Core Transient Model for BWRs," Volume 3, General Electric, February 1, 1986.
c.
- Letter, J.
S.
Charnley (General Electric) to C.
O.
Thomas (NRC), " Amendment 11 to General Electric Licensing Topical Report NEDE-2 4 011-P-A, "MFN-27-85, dated February 27, 1985.
d.
- Letter, J.
S.
Charnley (General Electric) to C.
O.
Thomas (NRC), " Response to Request No. 1 for Additional Information on NEDE-24 011, Revision ti, Amendment 11", MFN-94-85, dated July 18, 1985.
e.
- Letter, C.
O. Thomas (NRC) to J.
S.
Charnley (General Electric), " Acceptance for Ref erencing of LTR NEDE-24 011-P-A, Revision 6, Amendment 11,
'GE Standard Application for Reactor Fuel,'" MFN141-85, dated November 5, 1985.
f.
- Letter, G.
C.
Lainas (NRC) to J.
S.
Charnley (General Electric), " Acceptance for Referencing of Licensing Topical Report NEDE-24 011-P-A, 'GE Generic Licensing Reload Report, Supplement to Amendment 11,'" MFN-29-86, dated March 22, 1986.
g.
- Letter, E.
Fuller (General Electric) to D.
Ross (NRC),
" Transient Model Margins-ODYN Model Comparison to Peach Bottom Test Data," MFN-461-77, dated Lccember 1, 1977.
h.
- Letter, E.
Fuller (General Electric) to D.
Ross (NRC),
" General Electric Proposal for ODYN Licensing Basis Criteria," MFN-58-78, dated February 7, 1978.
( -
5.
NEDO 20939, "Ltttice Physics Methods Verification," General Electric, June 1, 1976 6.
NEDO-20946, "BWR Core Simulator Methods Verification, " General Electric, May 1, 1976.
7.
REFERENCE 7 INCLUDES THE FOLLOWING:
a
- a. NEDE-10958-P-A, " General Electric BWR Thermal Analysis Basis (GETAB) : data, Correlation, and Design Application, General Electric, January 1, 1977.
- b. NEDO-10958-A, " General Electric BWR Thermal Analysis Basis (GETAB): data, Correlation, and Design Application, General Electric, January 1, 1977.
c.
- Letter, D.G. Eisenhut (NRC) to R.L.
Gridley (General Electric), " Safety Evaluation for the GE L'cR, Generic Reload Fuel Application, Original Document NEDE-24011,'
MFN-212-78, dated May 12, 1978.
d.
- Letter, A.C. Thadani (NRC) to J.S.
Charnley (General Electric), " Acceptance for Referencing of Application of Amendment 15 to GE LTR NEDE-24011-P-A,
'GE Standard Application for Reactor Fuel, '" MFN-32-88, dated Match 14, 1988 e.
- Letter, A.C. Thadani (NRC) to J.S.
Charnley (General Electric), " Acceptance for Referencing of Application of Amendment 18 to GE LTR NEDE-24011-P-A,
'GE Standard Application for Reactor Fuel,'" MFN-51-88, dated May 12, 1988 8.
NEDO-10958-a, " General Electric BWR Thermal Analysis Basis (CETAB): Data, Correlation, and Design Application," General Electric, January 1, 1977.
9.
- 24A5156, Rev.
O,
" Supplemental Reload Licensing Submittal Report for Edwin I. Hatch Nuclear Plant Unit 1 Reload 15 Cycle 16," Revision 0, General Electric, October 1994.
10.
NEDO-32362, " Utility Licensing of Vendor Methods," GE Nuclear Energy, May 1994..
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