ML20081G273
| ML20081G273 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 01/06/1995 |
| From: | Bond G, Irani A, Luoma J GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20081G272 | List: |
| References | |
| TR-078, TR-078-R00, TR-78, TR-78-R, NUDOCS 9503230082 | |
| Download: ML20081G273 (110) | |
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' i TMI-1 TRANSIENT ANALYSES USING THE RETRAN COMPUTER CODE TOPICAL REPORT 078 REY. O BA NO.: 135425 AUTHOR:
belc4e kd A. A. IRANI ENGINEER, TMI FUEL PROJECTS JANUARY 6, 1995 APPROVALS:
y-A k ke J.' D. LUOMA MANAGER, TMI FUEL PROJECTS G. R. BOND' DIRECTOR, NUCLEAR ANALYSIS & FUEL gj[2ggg(( M of P
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TR-078 Rev. O Page 2 ABSTRACT This report presents the methods for performing system transient analysis for Three Mile Island Unit l' Nuclear Power Station. A description of the RETRAN model and general code features have been discussed. The adequacy of the model and the associated accident analysis methodology has been demonstrated by comparison of representative analytical results to vendor calculations and to plant data. The overall good agreement in these comparisons and the conservative response for licensing applications demonstrates GPUN's ability to perform operational and licensing analysis of TMI-1, using RETRAN-02 MOD 5.1.
I TR-078 Rev. O Page 3 TABLE OF CONTENTS
1.0 INTRODUCTION
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 2.0 BRIEF DESCRIPTION OF TMI-l ...................... 16 3.0 BRIEF DESCRIPTION OF RETRAN ...................... 12 4.0 GENERAL DISCUSSION OF TMI-l RETRAN MODEL . . . . . . . . . . . . . . . . 14 4.1 Best Estimate Versus Licensing .................. 15 5.0 DETAILS OF TMI-l RETRAN ML9EL ..................... 17 5.1 Reactor Vessel .......................... 17 5.2 RCS Piping and Pumps ................-....... 18 5.3 Once Through Steam Generators . . . . . . . . . . . . . . . . . . . 18 5.4 Steam Lines and Steam Valves ................... 19 5.5 Pressurizer . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 5.6 Heat Conductors . . . . . . . . . . . . . . . . . . . . . . . . . . 20 5.7 Reactivity Feedback . . . . . . . . . . . . . . . . . . . . . . . . 21 5.8 Trips . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 5.9 Steady State Initialization . . . . . . . . . . . . . . . . . . . . 22 6.0 QUALIFICATION COMPARISONS ....................... 32 6.1 Verification Against Experimental Test / Operational Data . . . . . . 32 6.1.1 Pump Coastdown Tests .................... 32 6.1.2 Reactor / Turbine Trip .................... 33 6.1.3 TMI-2 Accident ....................... 34 6.2 Verification Against Licensing Analysis ............. 39 6.2.1 Complete Loss of Coolant Flow Accident ........... 42 6.2.2 TMI-2 Locked Rotor ..................... 42 6.2.3 Startup Accident ...................... 44 6.2.3.1 Maximum Worth Control Rod Group at Maximum Rod Speed . . . . 45 6.2.3.2 All Control Rod Assemblies at Maximum Speed ........ 46 6.2.4 Uncontrolled Control Rod Assembly Withdrawal at Power .. 48 6.2.5 Cold Water Accident .................... 50 6.2.6 Steam Line Break Accident ................. 52 7.0 MODEL JUSTIFICATION .......................... 96 7.1 Built-in Models ......................... 96 7.1.1 Pump Model for Reactor Coolant Pumps . . . . . . . . . . . . 96 7.1.2 Nonequilibrium Pressurizer Model for Pressurizer . . . . . . 97 7.1.3 Algebraic Slip Model . . . . . . . . . . . . . . . . . . . . 98 7.2 Computer Code Uncertainties ................... 98 7.3 Nodalization . . . . . . . . . . . . . . . . . . . . . . . . . . .
7.3.1 99 RCS and Steam Line Nodalization .............. 99 7.3.2 SG Nodalization ...................... 100
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t TR-078 Rev. O Page 4 7.3.3 Core Conductor Geometry . . . . . . . . . . . . . . .... 101 7.3.3.1 Clad Mesh Sensitivity .
. . . . . . . . . . . . . . . .... 102 ,
7.3.3.2 Fuel Mesh Sensitivity . . . . . . . . . . . . . . . .... 102 i 7.4 Time Step Convergence ....................... 102 l 1
8.0 CONCLUSION
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9.0 REFERENCES
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! LIST OF TABLES j l
. l Table 5-1 Volume Geometric Data ...................... 23 i Table 5-1 Volume Geometric Data'(continued) ................ 24 ]
Table 5-2 Junction Data ..........................
25 l l Table 5-2 Junction Data (continued) .................... 26 {
Table 5-3 Heat Conductor Data ....................... 27 i Table 5-4 Description of Trips . . . . . . . . . . . . . . . . . . . . . . . 28 !
l Table 5-5 RETRAN Initial Conditions .................... 29 i
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.TR-078 Rev. O Page 6 LIST OF FIGURES 2-1 TMI-1 Reactor Coolant System .................... ~ll 5 TMI-I RETRAN Two Loop Model Nodalization Diagram ........... 30 5-1 TMI-I RETRAN Two Loop Model Nodalization Diagram (cont'd.) ..... 31 6-1 Four Pump Coastdown Benchmark-Percent Normalized Flow . . . . . . . . 56 6-2 Two Pump Coasidown Benchmark-Percent Normalized Flow ........ 57 6-3 One Pump Coas.tdown Benchmark-Percent Normalized Flow ........ 58 6-4 Reactor / Turbine Trip Benchmark-Hot Leg Temperature . . . . . . . . . 59 6-5 . Reactor / Turbine' Trip Benchmark-Cold Leg Temperature . . . . . . . . . 60 6-6 Reactor / Turbine Trip Benchmark-Pressurizer Level .......... 61 6-7 Reactor / Turbine Trip Benchmark-RCS Pressure . . . . . . . . . . . . . 62 6-8 Reactor / Turbine Trip Benchmark-Main Steam Pressure ......... 63 6-9 TMI-2 Accident Benchmark-RCS Pressure . . . . . . . . . . . . . . . . 64
'6-10 TMI-2 Accident Benchmark-Hot Leg Temperature ............ '65 6-11 TMI-2 Accident Benchmark-Cold Leg Temperature . . . . . . . . . . . . 66 6-12 THI-2 Accident Benchmark-Pressurizer Level ............. 67 6-13 TMI-2 Accident Benchmark-Main Steam Pressure ............ 68 6-14 Complete loss of Coolant Flow Accident FSAR C Normalized Flow . . . . . . . . . . . . . . . . . . omparison-Percent ......... 69 6-15 TMI-2 Locked Rotor Benchmark-Percent Normalized Flow ........ 70 6-16 THI-2 Locked Rotor Benchmark-Percent Neutron Power ......... 71 6-17 TMI-2 Locked Rotor Benchmark-Pressurizer Pressure . . . . . . . . . . 72 6-18 Single Rod Group Startup Accident - Percent Neutron Power . . . . . . 73 6-19 Single Rod Group Startup Accident - Percent Thermal Power . . . . . . 74 6 ?0 Single Rod Group Startup Accident - RCS Pressure .......... 75
- f. 41 Single Rod Group Startup Accident - Fuel Temperature Change . . . . . 76 6-22 Single Rod Group Startup Accident - Average Core Moderator Temperature Change 6-23
................................ 77 All Rod Groups Startup Accident - Percent Neutron Power . . . . . . . 78 6-24 All Rod Groups Startup Accident - Percent Thermal Power . . . . . . . 79 6-25 All Rod Groups Startup Accident - RCS Pressure ........... 80 6-26 All Rod Groups Startup Accident - Fuel Temperature Change . . . . . . 81 6-27 All Rod Groups Startup Accident - Average Core Moderator Temperature Change 6-28
............................... 82 RWA FSAR Benchmark - Percent Neutron Power ............. 83 6-29 RWA FSAR Benchmark - Percent Thermal Power ............. 84 6-30 RWA FSAR Benchmark - RCS Pressure . . . . . . . . . . . . . . . . . . 85 l 6-31 RWA FSAR Benchmark - Fuel Temperature Change ............ 86 6-32 RWA FSAR Benchmark -Average Core Moderator Temperature Change . . . . 87 6-33 Cold Water Accident FSAR Comparison-Percent Neutron Power . . . . . . 88 6-34 Cold Water Accident FSAR Comparison-Percent Thermal Power ..... 89 6-35 Cold Water Accident FSAR Comparison-Average Core Moderator Temperature Change ............................... 90 6-36 Steam Line Break Accident FSAR Comparison-Percent Neutron Power . . . 91 6-37 Steam Line Break Accident FSAR Comparison-Percent Thermal Power . . . 92 6-38 Steam Line Break Accident FSAR Comparison-Total Reactivity ..... 93 6-39 Steam Line Break Accident FSAR Com Temperature . . . . . . . . . . . . .parison-Average Reactor Coolant 94 6-40 Steam Line Break Accident FSAR Comparison-Temperature of Isolated Steam l
Generator . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 95
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7-1 Reactor / Turbine Trip Sensitivity-Hot Leg Temperature . . . . . . . . . . 104 ;
7-2 Reactor / Turbine Trip Sensitivity-Cold Leg Temperature ..... ... 105
- 7-3 Reactor / Turbine Trip Sensitivity-Pressurizer Level . . . . . . . . . . . 106 j 7-4 Reactor / Turbine Trip Sensitivity-RCS Pressure ........... . lJ7 i 7-5 Reactor / Turbine Trip Sensitivity-Main Steam Pressure . . . . . . . . . . 108 _
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1.0 INTRODUCTION
f GPU Nuclear (GPUN) has developed the capability to perform system transient analyses of Three Mile Island Unit 1 (TMI-1) Nuclear ~ Power Station. This ;
l includes the capability for performing plant operational support applications,.
l l licensing calculations and reload analysis. In a previous submittal (1), the
- Nuclear Regulatory Commission (NRC), approsed GPUN's methods for reload analysis for the Oyster Creek Nuclear Generating Station. !
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i The principal analysis tool is the RETRAN computer code (2) which determines the l
transient thermal-hydraulic response of a Nuclear Steam Supply System (NSSS).
The NRC has reviewed RETRAN and issued a Safety Evaluation Report (SER) which allows the code to be referenced in a licensing submittal (3).
i A general purpose two loop " base" RETRAN model of TMI-1 was developed for this purpose. The nodalization of the model was based on extensive experience with the RETRAN-02 code so as to allow the analysis of a wide range of transients.
The model wa: 1esign verified by an independent consultant (4) and qualified by comparison of code predictions against plant transient data and vendor licensing analyses (non-LOCA). For best-estimate analysis, the measured initial conditions were modeled, while for licensing calculations, the assumptions were similar to those documented in the FSAR.
This report is organized in the following manner: Section 2 is a brief description of TMI-l and Section 3 is an overview of the RETRAN computer code.
Section 4 is a general discussion of the TMI-1 RETRAN model while Section 5 provides details of the model. The results of a range of comparative analyses
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using the model are documented in Section 6 and the results of sensitivity ,
studies which justify the model options, nodalization, etc. are contained in .
Section 7. The report conclusions and eferences are provided in Sections 8 and .
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a TR-078 Rev. O Page 10 2.0 BRIEF DESCRIPTIDN OF TMI-1 TMI-1, is a 2568 MWth pressurized water reactor (PWR) manufactured by Babcock and Wilcox (B&W). The reactor coolant system consists of the reactor vessel, two vertical once-through-steam-generators, four shaft sealed reactor coolant pumps, an electrically heated pressurizer und interconnecting piping, as shown on Figure 2-1.
Tha system is arranged in two heat transport loops consisting of one once-through-steam-generator and two reactor coolant pumps per loop. The reactor coolant is transported through piping connecting the reactor vessel to the steam generators tubes, transferring heat to the steam and water on the shell side of the steam generator. In each loop the reactor coolant is returned to the reactor in two lines each containing a reactor coolant pump.
Other components include core flood tanks, main steam system consisting of isolation valves, safety valves and turbine bypass valves, feedwater system, Engineered Safeguards Actuation System and Reactor Protection System. A detailed description of the plant may be found in Reference 5.
i RELIEF VALVE NOZZLES b PORV VENT LINE l ,
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NOZZLE SPRAY LINE PRESSURIZER
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D CA HEAT NOZZLE ll w ,
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,HPl NOIZLE MAIN l FEEDWATER TSG4 g V-HEADER , ?
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Figure 2-1. TMI-I Reactor Coolant System -
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i 3.0 BRIEF DESCki! # OF RETRAN l The RETRAN computer code is a one-dimensional thermal hydraulic code developed by EPRI as a tool for best estimate analysis of light water reactor systems.
The code can model a Reactor Coolant System (RCS) as an assembly of volumes.
connected by flow paths or junctions. The volumes specify a region of fluid within a given set of fixed boundaries, whereas junctions represent the common flow areas of connected volumes. A pump or a valve can be inserted in a flow path. An extensive trip logic model, a non-equilibrium pressurizer model and l
a bubble rise model are also included. Heat conductors can represent materials which conduct heat into the fluid of a volume or between the fluids in two I i
different volumes. Critical flow can be modeled by selecting one of several l ;
l critical flow options. A point kinetics model is generally used to calculate '
the normalized power.
i The code uses one-dimensional flow equations to predict the thermal hydraulic !
transient results. The program solves the mass, energy, momentum and optional phase slip equations for subcooled water, two phase steam water mixtures and superheated steam using finite difference methods. Low order polynomials provide the equation of state for steam / water. The one dimensional heat conduction equation provides coupling to the fluid energy equation. Energy produced in the core, for example, is removed by the flow of coolant through the core. This is explicitly calculated by the code using basic principles.
The application of RETRAN to operational transients is especially accommodated
- through a flexible scheme used to model interaction of control systems with the thermal-hydraulic model.
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l TR-078 Rev. O Page 13 j RETRAN-02 has been widely utilized by utilities and consultants on a variety of )
transient problems. The current code version used here is RETRAN-02 M005.1 l which has been accepted by the NRC for referencing in licensing applications i (6). A detailed description of RETRAN is available in the literature and a. [
knowledge of the code is assumed for the use of this report. The code is !
installed and controlled on GPUN computers using GPUN procedures which assures :
the required quality assurance / control. ;
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4.0 SENERAL' DISCUSSION OF TMI-1 RETRAN MODEL 4
The RETRAN model for TMI-I documented in this report is based on a model which ;
has been in use within GPUN since 1978. Thus, the nodalization chosen for this !
model is based on extensive experience with the RETRAN code. A general purpose, base model consisting of two loops was developed in such a way as to allow the analysis of a wide range of transients with only minor modifications to the -
model input. The two loops and four cold legs allow for the analysis of asymmetric transients while the nodalization is of sufficient detail to provide j accuracy of results. The base model is comprised of those parameters which would not ordinarily change from cycle to cycle. This includes system geometry i
such as volumes and flow areas, heat conductors, characteristics of valves and '
pumps, etc. l Thus, specific transient cases may be analyzed without altering the base deck, l but by using override cards which are placed at the bottom of the deck and allow '
t' the modification of some of the input to model a specific transient. This is i explained in more detail below.
The TMI-l RETRAN model was design verified by an independent consultant (4) and qualified by comparison of code predictions against plant transient data and !
vendor licensing analyses. These benchmarks are for a representative series of l
transients, which exercise the model through a wide variety of perturbations.
l The qualification against plant data includes pump coastdown tests, a reactor / turbine trip and the TMI-2 accident.
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The comparisons ',o vendor predictions is for almost all the transients presented l in the FSAR. Thus the model was subjected to variations in primary coolant flow J
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-1 primary to secondary heat transfer,-etc. j i
4.1 Best Estimate Versus Licensing- ..
i The TMI-1 RETRAN ' base' model was constructed using best-estimate values. ,
i The geometry of the nodalization such as fluid volumes, areas, elevations - i i
were developed from as-built drawings. Other inputs such as valve setpoints, !
trip setpoint, etc. are also input at their actual value as defined in the Technical Specifications. j i Initial conditions also correspond to actual plant conditions. Thus the l actual representation of the plant is defined and this allows the model to 1
l be used for predicting realistic response of the plant to postulated or j i
actual transients. To perform such analyses, change or replacement cards 'are l
maintained in a separate file. These represent the actual conditions of the trans'ient being simulated, the initiating event, and assumptions' concerning availability of system components. These changes are small in number and appended to the bottom of the base deck, thus replacing the values defined i in the base deck. This allows a strict configuration control of the base deck and easier verification of changes made for simulation of transients.
In analyzing licensing calculations, assumptions consistent with those l documented in the FSAR would be used in the " modification" deck, and no changes would be made to the base deck. The deterministic approach assumes that all components are simultaneously at their most adverse values. These especially include the more sensitive parameters such as initial conditions,
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s TR-078 Rev. O Page 16 reactivity parameters and system performance assumptions. )
i The initial system power, temperature and pressure would include errors which j minimize core thermal margin or margin to other plant design criteria. The, j reactivity parameters are chosen in a manner which tends to maximize the f nuclear power during the transient. In many instances the mitigating effect-of various system design features on postulated transients are ignored.
Conservative instrument errors and system response times are assumed so as i
to bound the expected values. These assumptions assure that the ;
deterministic- or licensing calculations conservatively " bounds" the actual expected system performance.
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Page 17 5.0 DETAILS OF TNI-1 RETRAN MODEL i
The model is shown in Figure 5-1, and consists of 92 volumes,120 junctions and )
27 heat conductors. The data required for setting up the model was obtained i P
from as built plant drawings, reload reports, equipment manuals, operational i procedures / manuals and the FSAR. The modelling philosophy was built on GPUN experience in simulating a wide . range of plant transients over the years. A description of the model follows, and details of the geometric input data and ;
options for volumes, junctions and heat conductors are provided in Tables 5-1, )
1 5-2, and 5-3. Table 5-4 gives descriptions of the trips used in the model, and L
Table 5-5 lists some model initial conditions.
l 5.1 Reactor Vessel In the reactor vessel, single volumes are used to represent the downconer and lower plenum. The lower plenum node consists of the lower spherical head and the lower grid. The lower spherical head volume consists of instrument nozzle penetrations at the vessel, instrument guide tubes, flow distributor, portions of the lower grid, the flow stabilizer plates and the guide lugs.
l The downcomer volume is considered from the top of the inlet nozzle to the beginning of the spherical head. The downcomer volume is made up of the volume between the reactor vessel and the core barrel.
The reactor vessel upper plenum modelling uses three volumes to represent the upper plenum, outlet plenum and upper head. This representation models the
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flow path from the core to the upper plenum to the outlet plenum and a l l
leakage path from the upper plenum to the upper head to the outlet plenum. j l The active core region is modelled with three axial core volumes, three heat.
conductors and a single volume representing the core bypass region.
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The flow area in the core is composed of the difference between the area within the core barrel and the area of the N el rods, instrumentation tubes i l
and guide tubes.
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l 5,2 RCS Piping and Pumps l
The primary system piping is represented by single volumes for the hot leg piping, pump suction and pump discharge piping. The reactor coolant pumps are also represented by single volumes. The TMI-l pumps are Westinghouse pumps and homologous curves were developed for input into RETRAN based on the
! pump characteristics. The RETRAN pump model calculates pump behavior through the use of these empirically developed curves such that head and torque response of the pump are uniquely defined as functions of volumetric flow and speed. The moment of inertia, friction factors and rated speed corresponding j to actual pump data were input into RETRAN.
l l 5.3 Once Through Steam Generators l
l The steam generators are modeled in some detail so as to correctly identify the changes in the heat transfer processes which occur. Consequently, twelve
s TR-078 Rov. O Page 19 primary volumes, twelve secondary volumes and twelve heat conductors are used to model the tube region, with a single node representing the OTSG inlet plenum, outlet plenum and upper and lower downcomers. As feedwater is sprayed into the lower annulus, it is heated to its saturation temperature by mixing with and condensing steam from the region between the upper and lower baffles of the steam generator. This aspiration process is represented by junctions in each OTSG and the aspirator flow rate was adjusted to obtain the initial steam generator mass specified by the vendor.
The steam generator downcomer is modelled with phase separation allowing the model to more accurately simulate low power and post trip situations when the downcomer will have a more definable mixture level. The feedwater flow is modelled as a fill junction, where the user specifies flow as a function of time.
5.4 Steam Lines and Steam Valves Steam lines from each generator to the steam chest header are modelled, however, the two steam lines for each generator are combined into one in the RETRAN model. These are further divided into three volumes to properly locate the bypass, atmospheric dump and safety valves and to also predict the dynamic behavior within the steam line after the stop valve closure.
The bypass, safety and atmospheric dump valves are connected to the steam line volume which corresponds to their approximate physical location in the piping. These valves are modelled as negative fills, with'the safety valves being opened and reset via trips and the bypass and atmospheric dump valves
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being controlled by the control system.
Each OTSG has 9 main steam safety valves, and all valves with the same ,
setpoints are combined in the RETRAN model. l 5.5 Pressurizer i
The pressurizer is modelled as a non-equilibrium volume with phase separation. The heaters, sprays, relief and safety valves are also modelled.
The relief and safety valves are represented by junctions connecting the pressurizer to a containment (sink) volume in the RETRAN model. When the valve is opened, the junction flow becomes choked and the Moody critical flow option is chosen for choked flow calculations. Contraction coefficients are used on valve junctions to get the specified flow at the reference pressure.
The spray junction parameters are defined to limit the flow to the prescribed maximum and allow the flow to vary with the pressure drop when less than the maxim'um.
5.6 Heat conductors !
Three axial heat conductors are used to represent the fuel rods such that ;
l there is one conductor per fluid volume. A standard, cylindrical, three l region representation is used with three nodes in the fuel, one node in the gap and three nodes in the cladding. The gap thermal conductivity is set to yield a constant heat transfer coefficient throughout the core.
Similarly, twelve axial heat conductors are used to represent the I
once-through steam generator tubes such that there is one conductor per fluid I
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volume. A cylindrical representation with three nodes is used. The heat !
transfer correlation chosen was the GE CHF option for all conductors. The. f
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material properties for U0, fuel, zircaloy cladding and inconel are taken from ' ,
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References 7 and '8 and input as temperature dependent . tables. The j description and geometric data for the heat conductors is provided on Table' Ii 5-3. l l
l 5.7 Reactivity Feedback j l;
t i The core power response is determined by the point kinetics model in {
conjunction with implicit reactivity forcing functions and thermal feedback !
I effects from moderator and fuel in the three' core regions. The point kinetics model specified incorporater one prompt neutron group and six delayed neutron groups with decay heat represented by 'll delayed gamma l emitters and the radioactive actinides, U-239 and Np-23g. Explicit l
react'ivity forcing functions represent reactor scram while constant temperature coefficients represent feedback effects'. Core. power is !
distributed axially among the three core conductors based on a cosine shape. l 5.8 Trips A detailed trip logic is included in the TMI-1 RETRAN model. This includes simulation of the reactor protection system by high pressure, high flux, high ;
temperature, low pressure and flux / flow scram trips. Also included are the opening and closing of valves such as pressurizer spray, relief and safety valves and steam line safety valves, MSIV's, and turbine stop valves. Other
TR-078 i Rev.-0 j Page 22 '
I features include operation of the pressurizer heaters, HPI and manual (time) trips to simulate equipment failure or an initiating event. A description of the trips used is shown on Table 5-4. l 5.9 Steady State Initialization
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The RETRAN steady state initialization option was used to initialize the I model. The boundary conditions specified were hot leg pressure, steam generator outlet pressure and lower plenum enthalpy. The system flow and j core inlet flow were obt'ained from the FSAR (5) and do not change from cycle to cycle. The steam generator aspirator flow was adjusted to get the correct initial steam generator liquid mass as specified by the vendor. The steady state initial conditions from RETRAN were compared to information transmitted by the vendor, the'FSAR and plant data, and found to be in close agreement.
Some of the initial conditions are shown on Table 5-5.
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Table 5-1 volume seemetric pote volume Description Fluid Flou Bottoe* notes
- Volume Aree Elevation (Ft') (F t') (Ft) 001 Lower Plenue 621.64 38.95 -24.58 none 002 Lower Core 196.86 49.22 -16.72 none 003 niddle Core 196.86 49.22 -12.72 none 004 upper core 196.86 49.22 -8.72 none 005 upper Plenum 877.9 81.4 -4.72 none 006 core Bypass 193.23 17.11 -16.72 none 007 W pounconer 689.37 35.13 -18.7 none 008 W Upper heed 492.57 104.58 7.00 none 009 Outtet Plenum 247.96 23.29 -3.4 none
- i 108(208) not tog piping 526.3 7.07 -1.5 Transport Deley t 199(209) se upper Plenue 308.21 16.66 30.78 none 113.99 26.25 26.4 pone 110(210) ss Tubes 111(211) se Tubes 113.99 26.25 22. None 26.25 17.75 none i 112(212) ss Tubes 113.99 113(213) 36 Tdws 113.99 26.25 13.4 none 114(214) SS Tubes 113.99 26.25 9.07 none 115(2151 ss Tubes 113.99 26.25 4.72 none 116(216) ss Tubes 113.99 26.25 0.38 none 117(217) ss Tubes 113.99 26.25 -3.96 none 118(218) ss Tthes 113.99 26.25 -8.3 none 119(219) ss Tubes 113.99 26.25 -12.65 none '
120(220) ss Tubes 113.99 26.25 -16.99 none 121(221) ss Tubes 113.99 26.25 -21.34 None 122(222) ss Lower Planus 308.21 17.4 -28.3 none 123(223) Puup suction Piping 154.97 4.28 -30.92 Trenoport Detey 124(224) Reector Cootent Puup 56.00 1.E+10 -2.36 Artificiotty 1.orge Flow Aree 125(225) Pump Discherge Piping 119.27 4.28 -1.17 Transport petey 126(226) Pump suction Piping 154.97 4.28 -30.92 Trenoport Detey 127(227) meector Cootent Puup 56.00 1.E+10 -2.36 , Artificially Large Flow Aree 128(228) -Pump Discherge Piping 119.27 4.28 -1.17 Tronoport Delay 130 -Pressurizer 1550.0 38.4845 -3.02 non-equilibrium 131 Pressurizer sprey 1.ine 2.0 0.03 4.665. None
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Tabla 5-1 Volume Geometric Data (continued)
Volume Description Fluid Flou Bottoe* Notes
- Volume Aree Elevation
( F t') (Ft') (Ft) 151(251) SG Lower Downcomer 758.9 23.33 -21.34 None 152(252) SG Secondary 187.41 43.15 -21.34 Mone 153(253) SG Secondary 187.41 43.15 -16.99 None 155(254) SG Secondery 187.41 43.15 -12.65 None 155(255) SG Secondary 187.41 43.15 -8.31 None 156(256) SG Secondary 187.41 43.15 -3.% None 157(257) SG Secondary 187.41 43.15 0.38 None 158(258) SG 5econdary 187.41 43.15 4.72 None 159(259) SG Secondary 187.41 43.15 9.07 None 160(260) SG Secondary 187.41 43.15 13.41 Mone
-161(261) SG Secondary 187.41 43.15 17.75 Mone 162(262) SG Secondary 187.41 43.15 22.09 None 163(263) SG Secondary 187.41 43.15 26.44 None 164(264) SG Upper Downconer 461.88 23.33 10.98 Mone 165 'A' Steen Line(s) to 784.55 5.3 14.1 None Reector Building (RS) 166 'A' Stees Line(s) R8 333.66 5.3 22.75 Mone To MSIVs 167 'A' Stese Line(s) MSIVs 840.71 5.3 32.16 None To TSV's 168 Noin Steem Heeder 95.72 4.9 34.0 Mone 169 'A' Turbine Bypess 330.6 0.737- 22.75 None 181 conteinment Sink Volume 1.E+6 1.E+6 -50.00 None 182 Turbine Sink Volume 1.E+6 1.E+3 -50.00 Time Dependent 265 'B' Steen Line(s) to R8 752.3 5.3 14.1 Mone 266 '8' Steen Line(s), R8 to 308.44 5.3 22.75 Mone NSIVs 267 '8' Stees Line(s), MstVs 1134.46 5.3 32.16 Mone To TSV's 269 'a' Turbine Bypass 330.6 0.737 22.75 None
- System Zero Elevation is the Hot Leg Centerline.
O "Il -4 as 1l g- 33 S
w b
o.
. _ _ _ __.____.____m___.___._____.__ _-__.__.__m_._____ _ _ . _ - _ .r ,, -y . - -~ .*
Table 5-2 Junction pets [
Junction f Description Connects Volumes Flou Aree Elevction Notes From To (f t') (ft) 001 Core Intet 1 2 49.22 -16.72 Enth. Trans.
002 Core Bottom to Core Middle 2 3 49.22 -12.72 Enth. Trans.
- 003 Core Middle to Core Top 3 4 49.22 -8.72 Enth. Trans. l 004 Core outlet 4 5 49.22 -4.72 Enth. Trans. !
005 Louer Plenuse to core bypess 1 6 17.11 -16.72 None 006 Core Bypass to Upper Plenue 6 5 17.11 -4.72 Mone 007 W Dounconer to Louer Plenue 7 1 35.13 -18.7 Mone 008 Upper Plenue to Upper Head 5 8 24.1 7.0 Mone
.009 Upper Head to Outlet Plenue 8 9 8.56 7.0 Mone 010 Upper Plenue to outlet Plenue 5 9 50.75 4.25 vertical 108 (208) W outlet 9 108(208) 7.07 0.0 verticet 109 (209) sG Intet 108(208) 109(209) 7.07 37.74 Mone 110 (210) SG Tubesheet to tubes 109(209) 110(210) 26.25 30.78 Enth.Trans.
111 (211) 3G Tubes 110(210) 111(211) 26.25 2C.44 Enth. Trans.
112 (212) SG Tubes 111(211) 112(212) 26.25 *2.09
. Enth. Trans.
113 (213) SG Tubes 112(212) 113(213) 26.25- 17.75 Enth. Trans.
114 (214) sG Tubes 113(21't) 114(214) 26.25 13.41 Enth. Trans.
115 (215) SG Tubes 114(214) 115(215) 26.25 9.07 Enth. Trans.
116 (216) sG Tubes 115(215) 116(2161 26.25 4.72 Enth. Trans.
117 (217) 3G Tubes 116(216) 117(217) 26.25 0.38 Enth. Trans.
118 (218) 56 Tubes 117(217) 118(218) 26.25 -3.% Enth. Trans.
119 (219) SG Tubes 118(218) 119(219) 26.25 -8.31 Enth. Trans.
120 (220) SG Tubes 119(219) 120(220) 26.25 -12.65 Enth. Trans.
121 (221) 3G Tubes 120(220) 121(221) 26.25 -16.99 Enth. Trans.
122 (222) 36 Tubes to tube sheet 121(221) 122(222) 26.25 -21.34 Enth. Trans.
123 (223) 3G to pump suction piping 122(222) 123(223) 4.28 -26.84 Mone 124 (224) Pump intet 123(223) 124(224) 4.28 -2.36 Mone 125 (225) Pump outlet 124(224) 125(225) 4.28 3.5 vertical 126 (226) 36 to pump suction piping 122(222) 126(226) 4.28 -26.84 None 127 (227) Pump intet 126(226) 127(227) 4.28 -2.36 None 128 (228) Pump outlet 127(227) 128(228) 4.28 3.5 verticat
-130 Mot tog to Pressurizer -108 130 0.42 -1.5 No slip 131 Sprey line to Pressurizer 131 130 0.01 36.7 Spray 132 Cold tog to sprey line 128 131 0.02 4.665 No aos flux 133 (233) Cold tog to W dounconer 125(225)- 7 4.28 0.0 verticet 134 (234) Cold tog to W dounconer 128(228) 7 4.28 0.0 Verticet 135 Pressurizer Relief volve 130 181 0.0065 36.7 Noody choking 136 Pressurizer safety vetve 130 181 0.0354 36.7 Noody Choking 142 (242) High Pressure injection 0 125(225) 10.0 0.0 Fitt 150 (250) Aspirator 159(259) 151(251) 10.33 11.02 No stip 151 (251) Feeduster O 151(251) 1.0 10.64 Filt 152 (252) 36 dounconer to SG secondary 151(251) 152(252) 4.28 -19.81 No slip E N g ? o.
_ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ ___.....m._.__.. _ _ _ . . _ _ _ . _ . _ _ _ _ _ _ - - _ _ _ . _ _ . _ _ _ _ . . - _ _ _ . _ _ _ . _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ . -_____m. -
- - . , ...--w ..~. , +. . _ - . . ._ - - .m. _
_.m .-
Tablo 5-2 Junction Data (continuei..
Junction # Description Connects Volumes Flow Aree Elevation Notes From To - (f t') (ft) 153 (253) SG secondary 152(252) 153(253) 43.15 -16.99 Bubble Flow Regime 154 (254) SG secondary 153(253) 154(254) 43.15 -12.65 Bubble Flow Regime 155 (255) '36 secondary 154(254) 155(255) 43.15 -8.31 Bubble Flow Regime 156 (256) SG secondery 155(255) 156(256) 43.15 -3. % Bubble Flow Regime 157 (257) SG secondary 156(256) 157(257) 43.15 0.38 Bubble Flow Regime 158 (258) SG secondary 157(257) 158(258) 43.15 4.72 Bubble Flow Regime 159 (259) SG secondary 158(258) 159(259) 43.15 9.07 . Subble Flow Regime 160(260) SG secondary 159(259) 160(260) 43.15 13.41 Bubble Flow Regime 161(261) SG Secondary 160(260) 161(261) 43.15 17.75 None 162(262) SG secondary 161(261) 162(262) 43.15 22.09 None 163(263) SG secondary 162(262) 163(263) 43.15 26.44 None 164(264) ss to upper downconer 163(263) 164(264) 23.33 30.03 None 165(265) et SG outlet 164(264) 165(265) 5.3 15.02 None
.166(266) et reector building- 165(265) 166(266) 5.3 23.67 None
-167(267) et N51V's 166(266) 167(267) 5.3 33.08 None 168(268) et Tsy's _167(267) 168 5.3 33.25 Mone 169(269) et turbine bypees 166(266) .169(269) 0.67 33.09 None 170(270) stoo.pheric dump vetves 0 169(269) 0.07 33.09 None 171(271) safety vetve benk #1 0 166(266) 0.11 33.09 Neg. Filt 172(272) safety velve bank #2 0 166(266) 0.22 33.09 Neg. Fitt ,
173(273) safety vetve bank #3 0 166(266) 0.22 33.09 Neg. Fill ;
174(274) safety vetve bank #4 0 166(266) 0.22 33.09 Neg. FILL i 175(275) safety volve bank #5 D 166(266) 0.11 33.09 Neg. FILL 176(276) Turbine bypees vetves 0 169(269) 0.26 33.09 None 180(280) smett safety valves 0 166(266) 0.03 33.09 Neg. Fill 300 At turbine 168 182 10.0 34.0 Mone
! .l i
er Wsa g =
.l
_ _ m._m_. ... . . _ _ . . . . _ .
Tebte 5-3 Heat Conductor Data Meet Volume Geometry Surface Aree (Ft') Conductor Conductor # Left Right Type Left Right Volume (Ft') Description 1 (13)- 121 (221) 152 (252) 'Cylindricat 9835.69 11036.45 29.105 SG Tubes 2 (14) 120 (220) 153 (253) tylindricot 9835.69 11036.45 29.705 SG Tubes 3 (15) 119 (219) 154 (254) Cylindrical 9835.69 11036.45 29.705 SG Tubes 4 (16) 118 (218) 155 (255) Cytindricet 9835.69 11036.45 29.705 SG Tubes 5 (17) 117 (217) 156 (256) CyLinoricet 9835.69 11036.45 29.705 se Tubes 6 (18) 116 (216) 157 (257) Cylindricet 9835.69 11036.45 29.705 SG Tubes 7 (19) 115 (215) 158 (258) Cylindricet 9835.69 11036.45 29.705 SG Tubes 8 (20) 114 (214) 159 (259) Cylindrice'. 9835.69 11036.45 29.705 SG Tubes 9 (21) 113 (213) 160 (260) Cylinaricet 9835.69 11036.45 29.705 SG Tubes 10 (22) 112 (212) 161 (261) Cylindricot 9835.69 11036.45 29.705 SG Tubes Cylindricet 9835.69 11036.45 U.705 SG Tubes '
11 (23) 111 (211)' 162 (262) 12 (24) 110 (210) 163 (263) Cylindricat 9835.69 11036.45 29.705 SG Tubes 25 0 2 CyLindriceL 0.0 16578.06 148.51 Fuel nods 26 0 3 CyLindrice1 0.0 16578.06 148.51 Fuel nods 27 0 4 CyLindricet 0.0 16578.06 148.51 Fuel Reds sRo m da N
b
+ , , , . , . . . - . . - , ,- , . . .. _- . . . _ . . . . _ _ . , . . , , .._r r. _m -. - . , - . . . . - , . . . , . ~ :e....,m, , , , - - . . . . . . , _ . . ,. ,. . .._ - . . , . .... _. . . . - . ,
TR 078 Rev. O
{' Pcge 28 i
i Table 5-4 Description of Trips J
i Description Setpoint 4
Reactor Scram RCS Pressure > 2370 PSIA RCS Temperature > 618.8'F ,
RCS Pressure < 1915 PSIA i Normalized flux > 1.051
! Flux / Flow, Control Block -403 < setpoint Op:n Pressurizer Relief Valve RCS Pressure > 2465 PSIA J
Close Pressurizer Relief Valve RCS Pressure < 2415 PSIA i Open Pressurizer Safety Valve RCS Pressure > 2515 PSIA 1
Close. Pressurizer Safety Valve RCS Pressure < 2290 PSIA
- Opsn small steam safety valve Steam Line Pressure > 1055 PSIA
, Close small steam safety valve Steam Line Pressure < 1012 PSIA Open Steam Safety Valve Bank #1 Steam Line Pressure > 1065 PSIA-Close Steam Safety Valve Bank #1 Steam Line Pressure < 1022 PSIA l Open Steam Safety Valve Bank #2 Steam Line Pressure > 1065 PSIA 1 Close Steam Safety Valve Bank #2 Steam Line Pressure < 1022 PSIA L Op:n Steam Safety Valve Bank #3 Steam Line Pressure > 1075 PSIA
] Close Steam Safety Valve Bank #3 Steam Line Pressure < 1031 PSIA
- Open Steam Safety Valve Bank #4 Steam Line Pressure > 1095 PSIA i' Close Steam Safety Valve Bank #4 Steam Line Pressure < 1050 PSIA Opsn Steam Safety Valve Bank #5 Steam Line Pressure > 1107 PSIA l Close Steam Safety Valve Bank #5 Steam Line Pressure < 1062 PSIA Cicse MSIV's Steam Line Pressure < 615 PSIA Close Turbine Stop Valves RCS Pressure < 1915 PSIA Initiate High Pressure Injection RCS Pressure < 1615 PSIA 3
Turbine trip'on Reactor trip Any Reactor Scram Signal.
j Opin/Close Turbine Bypass valves Controlled by Control Model j Open/Close Atmospheric Dump Valves Controlled by Control Model
- Trip any Reactor Coolant Pump User specified trip time
- Open Pressurizer Spray Valve RCS Pressure > 2220 PSIA-i Close Pressurizer Spray Valve RCS Pressure < 2170 PSIA i Turn on Pressurizer Heater RCS Pressere < 2150 PSIA
, Bank #1-3 j Turn off Pressurizer Heater RCS Pressure > 2170 PSIA ,
, Bank #1-3 !
- Turn on Pressurizer Heater RCS Pressure < 2135 PSIA Bank #4
- Turn off Pressurizer Heater RCS Pressure > 2155 PSIA J Bank #4
! Turn on Pressurizer Heater RCS Fressure < 2120 PSIA Bank #5 1 1 Turn off Pressurizer Heater RCS Pressure > 2140 PSIA j Bank #5 I
a t
- I TR-078 Rev. 0 l l Page 29 ;
l I i
l i
[
, Table 5-5 RETRAN Initial Conditions :
Power 2568 MWt t
Pressure 2200 psia i Coolant Inlet Temperature 556*F -[
Coolant Outlet Temperature 602.5*F SG Secondary Pressure 930 psia !
SG Secondary Temperature 571*F r RCS Flowrate 139.7 x 10' LBM/Hr !
Core Inlet Flowrate 127.4 x 10' LBM/Hr
[
i I l t
) l l
I I
i i
i t
l t
i 3
208 ggg 1 131 gog z., O STEAM J o V 13 1 L- STEAM CENERATOR J209 Q ggj Ak J109 GENERATOR B gn 4 PORY 13g A m ~ M J 136 %
209 Q HEATERS 109 26a 3264 zio Q PRESSURIZE f Q tio '63 316*
262 U 163 zii 111 130 264 261 g 2 12 112 161 161 260
& W 9 9 d . ,
16' 2:a 5 113 226s 259 2:4 g JL :4 159 3p
,[3[ j2so 25e ans 227 ##8I T 6 3 7 #28 127 T m m m*
2s7 6
216 J 2 M' 16 157 2s6 2 17 I224 RPAI fRA 124D 117 2s6 ass ans its Iss 251 h REACTOR REACTOR IS I X C00 M T ~
PUMPS PRESSURE VESSEL COOLANT ase 2ns nns 1s*
G pyypg asa azo 120 153 2s2 k 221 121 $ 152 X tu THREE MILE ISLAND UNIT 1 Q W '
\222 Q RETRAN TWO LOOP MODEL NODALIZATION DIAGRAM 8 122 l Q , -
226 126 g g 22s n 3 12s FIGURE 5-1 ygg g .< o g oM
_ - . . ~ _ - _ _ . _ - _ . _ - - _-
SAFETY VLV
,'71
., o o ,175 o OTSC A 168 167 165 166 167 N
164 0
169 C EO S Ass TURBINE CONTROL VALVE 168 k^ 182 vm -
300 271~275 o o o o o OTSC B 267 268 266 265 266 267 X
gy 280 w- 269 269 ~270 ADV
= 276 BYPASS ona FIGURE 5-1 (Cont'd) {?$
e8 b
___ ____._____________m ____.___.______._____.____.______-_____m____ _______ ______m_ _ _ _ _ _ . _ . _ - -- _ _ _ - - _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ .
TR-078 Rev. O Page 32 6.0 QUALIFICATION COMPARISONS As verification of the TMI-l RETRAN model, appropriate results and comparisons have been performed for a representative series of analyses of licensing and best estimate plant transients. The results of these i
predictions are intended to demonstrate the adequacy of the TMI-1 RETRAN model and GPUNs ability to apply the RETRAN code for performing o~perational support and licensing calculations.
6.1 Verification Against. Experimental Test / Operational Data l
l This section shows comparisons of a variety of the TMI-I RETRAN model i
l predicted results to experimental test / operational data. These l comparisons inc,1ude pump coastdown tests of various configurations, a reactor / turbine trip and the TMI-2 accident, and demonstrate that the model provides an accurate representation of the reactor plant and associated systems and components.
6.1.1 Pump Coastdown Tests l
Pump coastdown tests of various configurations (i.e., coastdown of all four pumps, two pumps and one pump) are performed as part l
of the initial startup test sequence for new nuclear units.
RETRAN comparisons for four pump, two pump and one pump coastdown have been perfo.ued and are shown on Figures 6-1, 6-2, and 6-3. The RETRAN predictions for the four pump coastdown l
l
TR-078 !
Rev. O Page 33 show close comparison to data. The RETRAN predictions for two I pumps and one pump coastdown show a slightly more rapid flow f coastdown than data (less than 5% difference). ,
^
6.1.2 Reactor / Turbine Trip After the initial :t:rt-up for Cycle 8 (March 4,1990), the ,
plant was power limited fot* a short period due to fouling in the OTSG's. Consequently, a planned reactor / turbine trip was performed in. order to exercise the OTSG's and attempt to re-distribute the fouling that was causing the power limitation.
l The transient was initiated by a manual turbine trip followed by a reacto,r trip by actuation of the anticipatory reactor trip on turbine trip. Primary thermal hydraulic response was as expected following the reactor trip, as shown on Figures 6-4 to 6-8 and which show RETRAN predicting slightly higher temperatures than data. The reason for this difference is explain d below.
I The RCS temperature drop resulted in inventory shrink and caused pressurizer level to dip to approximately 66 inches as shown on I Figure 6-6. Make-up flow was initiated to recover the level and the RETRAN response matched data closely. Figure 6-7 shows close pressure response comparisons between RETRAN and the data, and shows that the plant was in stable post trip window within l l
'TR-078' Rev. O Page 34 320 seconds. Figure 6-8 shows 'the main steam pressure comparisons, which show a higher RETRAN prediction-for a'short period of time after 40 seconds. This difference is attributed to a main steam safety valve remaining open (partially) during the test and the OTSG pressure being' manually lowered to have '
this valve reseat.
No attempt was made to simulate this valve behavior with RETRAN as this would result in a -trial and error solution, and consequently the RETRAN seco0dary pressures rerained higher than data until the stuck open valva was reset and then the pressure comparisons were very close. ~ As can be seen fret Figures 6-4 and 6-5 this had a small effect on the temperature comparisons with RETRAN predicting slightly higher temperatures due to the slightly higher secondary pressure. The overall RETRAN predictions showed close comparison to data, and due to the unknown stuck open safety valve flow, no acceptance criteria was established.
6.1.3 TNI-2 Accident The TMI-2 Accident of March 28, 1979 was initiated by a loss of normal feedwater to the steam generators resulting in a turbine trip. The reactor coolant system responded to this initiating event in a normal manner as follows: reactor coolant system pressure increased because heat was not being removed from the
i n.
TR-078 Rev. O Page 35 system at sufficient -rates by the steam generators; the electromatic relief valve operated to relieve pressure; the reactor shut down automatically because of.a high-pressure trip signal; heat generation from the reactor dropped to the decay )
heat level; within a fr seconds the system pressure dropped to )
i normal values. ;
Approximately forty seconds into the event, closed valves !
between the control valves and 'the steam generators prevented i l \
l emergency feedwater from being delivered to the steam generators !
(these valves were opened by the operators approximately 8 minutes after the accident was initiated).
l The electromatic relief valve which relieved excess pressure, as intended, should have closed when pressure was reduced sufficiently. Instead, it remained open, thereby allowing continued coolant discharge from the reactor coolant system, and causing a further decrease in reactor coolant system pressure.
This loss of the reactor coolant continued without interruption
- until approximately 2.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (142 minutes) -into the accident when the block valve, which is in series with the electromatic relief valve, was closed.
l
- Since this' analysis only benchmarks about the first 1500 seconds
! of the accident, details of the rest of the accident are not i.
l provided. The RETRAN predictions and comparison to data are
I l
TR-078 l Rev. O Page 36 shown on Figures 6-9 through 6-13.
~
i The trip of the main feedwater pumps and the turbine, resulted )
i' in an imbalance between heat added to the primary system fluid by the core and that removed via the steam generators. This " ,
resulted in an initial increase in primary system pressure and temperatures as shown on the very early portion of Figures 6-9, 6-10, and 6-11. The increase in pressure caused the PORV to j open in about I second as the system pressure- exceeded the I opening setpoint of 2265 PSIA. The pressure relief provided by )
l I l
the pressurizer was not sufficient to control the system heatup, I i
and pressure continued to rise resulting in a reactor scram at j 8 seconds on a high pressure trip (2370 PSIA). A few seconds after the reactor trip, power generation decreased to decay heat power and less heat was being generated than removed via l inventory boil-off in the two steam generators. This plus the open PORV caused a reduction in the primary system pressure.
When the primary system pressure decreased to the electromatic l relief valve closure setpoint, (2210 PSIA) the valve failed to close and remained stuck in the fully open position, thus continuing to depressurize the primary system. With no feedwater available, the steam generator secondary side continued to lose inventory and the water level quickly reached the setpoint for initiating auxiliary feedwater (AFW) flow.
However, auxiliary feedwater was delayed for approximately 8
l TR-078
! Rev. O Page 37 minutes until the operators discovered the isolation valves -
! closed.
I 4
Shortly after the reactor trip, letdown flow was stopped, and a f second makeup pump was started and a high pressure injection f isolation valve was opened (40 seconds). These actions were 1 '
taken by the operators to compensate for the expected reduction l of pressurizer level. This HPI flow held the primary system i.
{ temperatures at fairly constant values as seen on Figures 6-10 i and 6-11. This is the period between 40 seconds and 278 seconds. The RETRAN HPI flow for this. period is modelled as a-l table of system pressure versus flow for two makeup pumps for the TMI-l configuration. The difference between the TMI-1 and f TMI-2 HPI flows was not evaluated given- the good results 4
obtained with the TMI-1 HPI flow information. It appears from
! Figures 6-10 and 6-11 that the RETRAN flow was slightly less l
l; than the actual flow, as the RETRAN temperatures are increasing
- at a slightly faster rate than the data.-
I While the HPI flow was entering the RCS in _ response to the l
i decreasing pressure shown on Figure 6-9, the pressurizer level
{ was increasing as a result of the stuck open PORV, as shown on i~
Figure 6-12.. The RETRAN level turned around a little earlier 4
than the data and consequently the RETRAN prsssurizer filled up 4
earlier than the data. At about 280 seconds a pressurizer high coolant . level alarm caused the operator to stop one of the I
_ - . _ _ _ _ _._ __. . . . _ _ . . . ~ _. . _. _.
.. l TR-078' Rev. 0 l Page 38 l
makeup pumps and to throttle the high pressure injection valves.
The throttled HPI flow is not known; a small HPI-flow (less than l one pump) was assumed in RETRAN. Sensitivity studies have . '!
determined that the pressure 'and temperatures response- during :
this period is not sensitive to the value of the HPI flow. The f i
combined effects of throttled HPI flow and unavailability of the ;
secondary ' side heat sink caused the primary system to heat up between about 280 seconds:and 500 seconds as shown on Figures !
6-10, and 6-11. The RETRAN temperatures seen on these figures !
rose at a slightly faster rate than data. - T'he RETRAN pressure rise during this period as seen on Figure 6-9 rose to a slightly -,
higher value than the data. One of the factors contributing to i this higher rise is due to the RETRAN model going solid slightly.
earlier than the data. Another unknown during this time period is that just before 300 seconds letdown flow was off scale high and this value is not known.
During this time, the steam generator secondary side.had little or no water and was unavailable as a heat sink. The main steam pressure response is shown on Figure 6-13. The decreasing SG pressure during this period (200-500 seconds) is due to steam bleed flow leaving the dry generators' for auxiliary purposes such as turbine seals and steam line supports. At about 500 seconds into the accident, AFW was introduced to the OTSG's by manual opening of block valves. This reestablished primary to i l
secondary heat transfer, causing the primary coolant l l
TR-078 Rev. 0 ,
Page 39 -
temperatures and pressure to decrease. The SG secondary side repressurized and was controlled at the turbine bypass system setpoint as seen on Figure 6-13. The RETRAN predictions showed l close comparison to data for this parameter.
l Subsequent to 500 seconds with the initiation of AFW flow, the primary system pressures and temperatures decreased gradually.
The RETRAN predictions closely followed data from about 500 to l 1500 seconds at which time the RETRAN simulation was terminated.
The RETRAN predictions for the TMI-2 accident show good ,
comparisons to data for most of the parameters. The trends of the predictions are similar to measured data with some differences in the magnitude of the predictions. These are due l to uncertainties in the available data and due to differences in the TMI-1 model and the THI-2 plant. In view of these uncertainties no acceptance criteria was established, and these comparisons were determined to be adequate for model I verification.
i l
6.2 Verification Against Licensing Analysis As discussed in the introduction, GPUN system transient analysis is intended for both best estimate and licensing applications. The results of a system thermal hydraulics analysis are used either for direct comparison to accident analysis acceptance criteria (e.g.,
1 system pressure limits) or as a boundary condition for more detailed 1
. - - __ A
TR-078 Rev. O Page 40 core thermal hydraulic analyses.
I Since core reloads are the most common and expectpd reason for l
accident reanalysis, GPUNs ability to perform these analysis for TMI-I is demonstrated in this section. The purpose here is to show that for ' !
a representative series of transients, the modelling of licensing, I conservatisms is understood and implemented, and the TMI-l model coupled with the RETRAN code provides expected system response.
1 F
I The nuclear fuel vendor has performed reload safety analysis for the .
TMI-I initial core, and these are contained in the safety analysis a
section of the FSAR. The FSAR is the reference analysis for TMI-1 and determines the values of key reload parameters. In subsequent i reloads, it is necessary to ensure that those key parameters which !
influence accident response are maintained within the bounds or
. " limits" established in the reference analysis. Thus reanalysis of an ,
accident which is part of the licensing basis occurs only under certain conditions.
For cases where a parameter falls outside these previously. defined limits, an evaluation of the impact of the change on the.results for I the appropriate transients must be made. This evaluation may be based j on known sensitivities to changes in the various parameters in cases .
where a parameter change is small or the influence on the accident results is weak. For cases where larger parameter variations occur, or for parameters which have a strong influence on accident results, ;
i
, - - . , .---,o--,.-,.-.,.w.,.,---_-c.e- -,v
T 1
TR-078 Rev. O Page 41 i
explicit reanalysis of the affected transients is required and
, performed.
i If required, a reanalysis is performed and the results are compared to l .
j the appropriate analysis acceptance criteria. The reload evaluation 1
l j process is complete if the acceptance criteria are met. If the analysis acceptance criteria are not met, more detailed analysis J
methods and/or Technical Specifications changes may be required to l meet the acceptance criteria.
1 4
4 i
The analysis performed in this section address most of the analysis in j the FSAR and demonstrates that the TMI-I RETRAN model will provide correct and conservative system response for a wide range of initiating events.
While a comparison to FSAR predictions has been provided, it should be recognized that a simplistic comparison is not necessarily meaningful.
l The vendor predictions in the FSAR were performed for the initial core
^
\
over twenty years ago, and the computer programs used at that time were not as sophisticated as the currently available programs.
i Consequently, in many instances, simplifying assumptions /models were 4
used by the vendor. In view of all the differences, the overall comparisons are still quite reasonable.
i j
i 8
l I
TR-078 Rev. 0 Page 42 l
l 6.2.1 Complete Loss of Coolant Flow Accident i
} A four pump coastdown loss of coolant flow (LOCF) event is the complete loss of forced flow in the RCS. With the reactor at l power, the loss of forced flow through the RCS results in an increase in reactor coolant temperature and a reduction in heat
- removal capability of the reactor coolant. This could result in departure from nucleate boiling (DNB) in the core. Reactor 1 i
protection is provided by the power / pump monitors trip function '
of the RPS.
i l
l The FSAR results of this analysis show that the reactor can l I
sustain a LOCF accident without damage to the fuel. This is
.i
- done by comparing the cinimum DNBR for the coastdown with the l criterion value. The comparison presented in this calculation is limited to the flow coastdown prediction after loss of all pump power. Figure 6-14 shows the RETRAN results compared to the FSAR, with RETRAN predicting a slightly faster flow decay than the FSAR. The RETRAN predictions are conservative with respect to the FSAR and the difference is less than 3%.
6.2.2 TMI-2 Locked Rotor i i
The locked rotor transient is the postulated instantaneous seizure of a reactor coolant pump rotor. Flow through the affected reactor coolant loop is rapidly reduced thus leading to
TR-078 Rev. O Page 43 a reactor trip on a flux / flow imbalance signal. The reduced flow causes a reduction in heat transfer to the secondary side of the steam generators. This combined with the heat being transferred to the reactor coolant from the fuel rods causes a rapid expansion of coolant leading to an insurge into the '
- pressurizer.
l The TMI-2 FSAR results were chosen for this comparison as the system response for flow, power and pressure were provided, while the TMI-l FSAR only contained a DNBR curve. The geometric differences between TMI-2 and TMI-I are small, and the TMI-l model was used for this benchmark with changes for initial
. conditions and other known differences, i
Figures 6-15 to 6-17 show the comparison of results for flow I coastdown, neutron power and pressure, between the RETRAN model predictions and the FSAR. Figure 6-15 shows RETRAN predicting a slightly faster flow coastdown than the FSAR. The flow coastdown results in an increase in neutron power as shown in j l
Figure 6-16 until the reactor is tripped by the flux / flow trip. '
The system pressure is shown in Figure 6-17 with RETRAN predicting a more rapid rise and higher pressure peak than the FSAR. This pressure response is consistent with the faster flow coastdown predicted by RETRAN. The pressure response is also
- affected by the assumptions for the Doppler and moderator l
coefficients and the measurement errors in the flux / flow trip.
l l
l
. l TR-078 Rev. O Page 44 .
i Differences in these values could account for the differences in !
predictions. Code and model differences and differences in some modelling assumptions make a simplistic comparison not necessarily meaningful. The RETRAN predictions were l
conservatively computed and this comparison is adequate as part of the effort for model verification. .l 6.2.3 Startup Accident i
The objective of a normal startup is to bring a subcritical l l
reactor to the critical or slightly supercritical condition, and i
then to increase power in a controlled manner until the desired power level and system operating temperature are obtained. 1 During a startup, an uncontrolled reactivity addition could I
cause a nuclear excursion. The uncontrolled reactivity addition, through rod withdrawal from zero power, is a startup accident. As positive reactivity increases, the power level ,
1 increases, and the reactor coolant and fuel rod temperatures increase. This excursion is terminated by the strong negative Doppler effect if no other protective action operates.
The FSAR shows the results for the following two startup accident cases:
- 1) maximum worth control rod group at maximum rod speed l
- 2) all control rod assemblies at maximum speed.
l l
1 TR-078 Rev. O i Page 45 The RETRAN comparisons and a discussion of the results are j presented below. Since the FSAR results did not contain a time j i scale, the FSAR results were overlaid on RETRAN results so as to j
] I provide the best agreement on event times. l l
- 6.2.3.1 Maximum Worth Control Rod Group at Maximum Rod Speed i
This rod velocity and worth (1.09 x 104 Ak/k/sec) result in
] the maximum reactivity addition rate. This reactivity addition-rate results in an increase in power level. In the absence of j secondary heat removal, the increased power level increases the !
j reactor coolant and fuel rod temperatures. The Doppler effect j begins to slow the neutron power rise, but the heat input to .
l the reactor coolant increases the pressure past the high )
pressure trip setpoint. The overpressure transient is )
) terminated following lift of the pressurizer safety valves.
I I
1
! l Figures 6-18 through 6-22 present the RETRAN results, compared '
i j to the FSAR for neutron power, thermal power, system pressure, i
fuel temperature change and average core moderator temperature !
change. The results for neutron power and thermal power are shown on Figures 6-18 and 6-19 and the FSAR shows a slightly
, higher value than RETRAN for these parameters.
i The system pressure predicted by RETRAN in Figure 6-20, shows j the same trend as the FSAR until about the time the reactor )
- k I
l
- -, -. - .. . - . - . ~
l - l TR-078 Rev. 0 ,
Page 46 :
i trips. After reactor trip, the core power decreases, reducing f
the fluid expansion (heatup) rate. The mass flow rate into the l pressurizer should reduce significantly, resulting in a j i
decrease in the pressurization rate. RETRAN properly predicted :
l this sequence of events. i l
i In the FSAR prediction, the rate of primary system j pressurization remained unchanged following reactor trip and l the pressure transient was terminated by safety valve lift. l Subsequent to reaching the safety valve setpoint, RETRAN l controlled system pressure between the open and close setpoints
- as is to be expected. .
Figure 6-21 shows the average fuel temperature change l comparisons with RETRAN showing the same trend as the FSAR but l
having a slightly lower peak temperature. Similarly, Figure 6-22 shows the average core moderator temperature change which ,
also shows similar behavior to the FSAR but with a slightly l higher peak.
6.2.3.2 All Control Rod Assemblies at Maximum speed This reactivity addition rate (7.25 x 10 4 Ak/k/sec) is much higher than the previous case. Consequently, the power rise is more severe than the previous case. The power rise is turned around by the negative Doppler effect. The high neutron flux trip takes effect after the peak power is reached and terminates
4 TR-078 Rev. O Page 47 the transient. Because the event is terminated quickly, the pressure excursion is much less severe than that for the single rod group withdrawal from hot zero power.
)
)
Figures 6-23 through 6-27 present the RETRAN results, compared to the FSAR for neutron power, thermal power, system pressure, l l
fuel temperature change and average core moderator temperature !
i change. The results for neutron power and thermal power are >
shown on Figures 6-23 and 6-24 and RETRAN shows a much higher j s
neutron power peak than the FSAR. The thermal power response in ;
Figure 6-24 shows close comparison between RETRAN and the FSAR.
l
)
The system pressure predicted by RETRAN in Figure 6-25, matches the FSAR except for the initial rise. The RETRAN response appears to be more realistic than the FSAR, given the rapid rise in neutron and thermal powers.
1 Figure 6-26 shows the average fuel temperature change !
comparisons where the trends between RETRAN and the FSAR are almost identical with RETRAN having a slightly higher peak.
Figure 6-27 shows the average core moderator temperature change comparison with about a 3*F higher prediction by RETRAN.
l The RETRAN predictions for the startup accident show the same trends as vendor predictions with small differences in magnitude for some parameters. Code and model differences and some
TR-078 l Rev. O l Page 48 I
different modelling assumptions make simplistic comparison not '
necessarily meaningful. The RETRAN predictions in this i
comparison are adequate to be part of the effort for model !
l verification. The acceptance criteria are met with considerable l margin.
i l
6.2.4 Uncontrolled Control Rod Assembly liithdrawal at Power l i
A rod withdrawal accident (RWA) is the accidental withdrawal of i a control rod group while the reactor is at rated power. This uncontrolled withdrawal, through operator error or equipment failure, results in positive reactivity addition. As positive 1 reactivity increases, the power level increases, the reactor coolant and fuel rod temperatures increase, and, if the withdrawal is not terminated by the operator or reactor l protection system, core damage would eventually occur.
The FSAR presents the results of this transient for several initial power levels, reactivity insertion rates, trip delay l times and moderator and Doppler temperature coefficients.
However, a full range of system parameter transient results is i
presented only for one analyr.is from an initial power level of 100%.
This case was a one rod group reactivity addition rate of 1.09 x 104 (Ak/k)/sec and results in a reactor scram on high l
TR-078 Rev. 0 :
Page 49 flux. Figures 6-28 through 6-32 present the RETRAN results:
compared to the FSAR for neutron power, thermal power, system pressure, fuel temperature change and average moderator temperature change. The results for neutron power and thermal power are shown on Figures 6-28 and 6-29. and the increase in j these parameters is almost identical between RETRAN and the FSAR. A slight difference in neutron flux predictions between RETRAN and the FSAR results in an earlier FSAR reactor trip.
Due to the slow nature of this transient a very small difference !
in neutron flux corresponds to a much larger difference in trip time.
l The system pressure predicted by RETRAN in Figure 6-30, shows the same trend as the FSAR until reactor trip. Subsequent to '
reactor trip, the FSAR analysis models the steam generator heat removal as a heat demand versus time table. This non-linear decay of heat demand resembled the post reactor trip feedwater run-back and turbine trip. In RETRAN, post reactor trip is modelled dynamically as closure of the turbine stop valves with feedwater assumed to runback to zero in 15 seconds. This results in a pressure increase in the secondary which is relieved by opening of the turbine bypass and atmospheric dump valves. However, this secondary pressure increase is transmitted to the primary as shown in the figure and results in a conservative and accentuated overpressurization aspect of the event.
- .- . . ~ > - - ..
i 6
TR-078 Rev. O Page 50 Figure 6-31 shows the fuel temperature change comparisons, with RETRAN showing the same trend as the FSAR. The middle of the RETRAN three core nodes average metal temperature was used, and this predicts a larger fuel temperature change than the FSAR.
Figure 6-32 shows the average core moderator temperature change
- which also shows the same trend as the FSAR.
The overall RETRAN predictions show close comparison with vendor l predictions. The RETRAN predictions were conservatively calculated, and this comparison is adequate as part of the l effot; for model verification.
l 6.2.5 Cold Water Accident i
The classic cold water accident is initiated with startup of an idle reactor coolant loop which has been isolated by primary
. system isolation valves. Since the B&W plant design does not include primary system isolation valves, the classic cold water accident cannot occur. However, when the reactor is operated with one or more idle pumps and these pumps are then started, the increased flow rate causes the average core temperature (Tavg) to decrease. If the moderator temperature coefficient is negative, positive reactivity is introduced into the core and a power rise occurs.
Figures 6-33 through 6-35 present the RETRAN results compared to ;
1 n , - , - . .
4 4
TR-078 Rev. O Page 51 the FSAR for neutron power, thermal power and average moderator temperature change. The three parameters are closely coupled.
The RETRAN predictions show the expected response of the system as the pumps are started. The pump speed was assumed to reach
~
full speed of 1190 RPM in 9 seconds. The initial reverse flow through the loops with the idle pumps becomes positive as the pump speed increases, and results in the moderator temperature decrease r, .;hown on Figure 6-35. Using the assumption of a negative moderator temperature coefficient, this results in a power rise as.shown on Figure 6-33. The thermal power rises in response to the neutron power rise as shown on Figure 6-34.
The FSAR predictions show moderator temperature decreasing more !
rapidly than RETRAN resulting in an earlier neutron power rise I
than RETRAN. It is believed that the FSAR used an independently calculated mixed moderator temperature coefficient as a function i of time and imposed it as input to the compui.ar model.
The FSAR assumptions imply a more instantaneous mixing than RETRAN, as when an RCP is turned on, it takes time for the initial reverse flow through that loop to become positive. The loop flow and its effect on moderator temperature are dynamically calculated in the RETRAN model. The different modelling assumptions make a simplistic comparison not necessarily meaningful; however, the RETRAN results are conservative and adequate as part of the effort for model verification.
i l
I
. j l
TR-078 Rev. O Page 52 6.2.6 Steam Line Break Accident A steam line break is the result of a-break between the steam !
generator and the turbine. The FSAR analysis assumed a double-ended rupture of one steam line between the steam '
i generator and the turbine. All systems and components were - )
~
assumed operable.
The loss of the secondary coolant due to the break causes a decrease in steam pressure and the increased steam flow across the generator tubes lowers the average reactor coolant temperature. With a negative moderator temperature coefficient, the resulting positive reactivity insertion to the core will cause a power increase. A reactor trip occurs due to low reactor coolant pressure or high neutron flux. Following reactor trip, the turbine trips, and the turbine stop valve and feedwater control valves close. Low steam line pressure initiates automatic feedwater isolation, which causes the steam generator associated with the rupture to blow dry. Continued RCS cooldown and decay heat removal are achieved by emergency feedwater flow to the unaffected generator with steam flow through the turbine bypass valve.
i Although not accounted for in this analysis, the high pressure '
injection system would be actuated during the cooldown period l following a large area steam line, break. This system supplies 1
1 J
d TR-078 Rev. O Page 53 borated water to the RCS to increase the shutdown margin. Boron addition to the reactor coolant, during the controlled cooling
- to atmospheric pressure, will prevent criticality _ at lower temperatures. The return to critical and subsequent' return to -
5 .
power is a concern for this transient, and the FSAR analysis j which was bounding for TMI-1 concluded that the core would not i return to critical. Several acceptance criteria are associated 1
with the steam line break accident as outlined in the FSAR, but this calculation is limited to a comparison of the system response. ,
i Figures 6-36 through 6-40 show the RETRAN results compared to the FSAR for neutron power, thermal power, total reactivity, .
average reactor coolant temperature and temperature of isolated !
Initially, both steam generators blow down causing a drop in -
average reactor coolant temperature as shown on Figure 6-39.
The negative moderator temperature coefficient causes' an increase in neutron power as shown on Figure 6-36, until a high ;
I flux reactor trip occurs. The neutron power then_ decreases to j decay levels. The thermal flux follows the same profile as the j neutron flux, as shown on Figure 6-37. ;
The RETRAN model scrams the reactor shortly after the high flux' '
setpoint is reached, (114% assumed in the FSAR) while the FSAR I
t- - - -
a
i i
l .. TR-078 -
Rev. O j Page 54 j j power prediction continues to increase t'o a peak of greater than l 160%. This appears to be excessive, especially in light of the ,
t total reactivity shown on Figure 6-38 which indicates a positive l
{ reactivity of only 0.2% Ak/k. The vendor agrees- that the :
~
prediction for_ neutron power appears excessive; however, since i l this is conservative, it satisfies the purpose of licensing. !;
4 l analyses. Also noticeable in these and other figures is a I timing difference with the FSAR lagging RETRAN by a few seconds.
i This can be related to an FSAR simplifying assumption, as shown:
! on Figure 6-39, where the average reactor coolant temperature t
- stays almost constant for seven seconds before decreasing..
i.
Figure 6-40 shows the reactor coolant temperature leaving the i unaffected steam generator increases after the turbine stop i
- valve closes as a result of pressure recovery and a reduction of feedwater flow. The coolant temperature leaving the affected
! steam generator decreases until it has blown dry, at which time l it approaches the inlet temperature. Since the unaffected steam i
4 generator turbine stop valves are closed, and the steam j
generator with the rupture is dry, core decay -heat can be
{
! removed through the unaffected steam generator by venting steam i
through the turbine bypass valve.
Also, the operators can lower the RCS temperature by this means; however this analysis does not take credit for operator action.
Figure 6-40 shows the isolated steam generator maintaining a
, ,- . . _ . _ _ _ , , - , , . _ . . _ - , _ _ . , - , , , , . _ , - , , , , , , , . ~ , . . _ . ,
i l .
~
TR-078 ;
Rev. 0 l Page 55 i constant temperature corresponding to the (saturation) pressure I at which the turbine bypass valve is set to open. Eventually, ;
a thermal equilibrium state is reached with heat removal through ;
steam flow through the turbine bypass valve being equal to the i heat from core decay. ;
Figure 6-38 shows that after the reactor trip the core remains subcritical. The differences between RETRAN and FSAR after 60 ;
seconds relate to the difference in coolant temperature as seen on Figure 6-39. The RETRAN predictions were conservatively l calculated, and this comparison is adequate as part of the l effort for model verification.
i l
l l
l
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- Figure 6.2 Two Pump Coastdown Benchmark-Percent Normalized Flow -
_ _ _ _ _ _ _ _ _ _ _ . . .
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m l
i . .
l TM1-1 FOUR PUMP COASTDOWN - FSAR BENCHMARK i i i i i i 4
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TMI-2 LOCKED ROTOR - FSAR BENCHMARK e_ e
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TR-078 Rev. O Page 96 7.0 MODEL JUSTIFICATION This section provides justification of the accuracy of the TMI-1 RETRAN model and its adequacy for licens,ing applications. The validation efforts have been demonstrated by good comparison to transient data. The adequacy for licensing applications is demonstrated by performing a representative series of FSAR analyses and comparing the results to vendor predictions.
The THI-1 RETRAN model results provide the same trend as the vendor and conservative responses for a wide range of initiating events. The conservative response of RETRAN is sufficient justification of the model for licensing applications; however, sensitivity studies presented in this section further justify the nodalization, model options, phenomenological models, etc.
7.1 Built-in Models The TMI-I RETRAN model utilizes the following RETRAN built-in rodels:
- 1. Pump model for reactor coolant pumps.
- 2. Nonequilibrium pressurizer model for pressurizer.
- 3. Algebraic slip model.
7.1.1 Pump Model for Reactor Coolant Pumps The THI-l pumps are Westinghouse pumps and ' homologous curves were developed for input into RETRAN based on the pump characteristics. The RETRAN pump model calculates pump behavior
TR-078 Rev. O Page 97 through the use of these empirically developed curves such that head and torque response of the pump are uniauely defined as functions of volumetric flow and speed.
The _ adequacy of this model which is used in the single phase, nonnal quadrant is demonstrated by comparison of model predictions to measured plant data for four pump, two pump and one pump coastdown tests in Section 6. In all instances the RETRAN predictions showed a slightly more rapid flow coastdown than the data. Since this is conservative and the differences to data are small (less than 5%) the model is adequate for use in licensing applications.
7.1.2 Nonequilibrium Pressurizer Model for Pressurizer The nonequilibrium pressurizer model is used to predict the nonequilibrium effects (different temperatures) between the vapor and liquid in the steam-water interface region of the pressurizer. This is important during rapid pressurization events when the vapor region superheats. The use of a bubble rise velocity of 3.0 ft/sec., a density gradient of 0.8, and a rainout velocity of 3.0 ft/second are all industry standards.
The accuracy of the use of the model has been demonstrated in Section 6 by comparison of model predictions to measured plant data for the TMI-2 accident and for a reactor turbine trip. For the TMI-2 accid:nt, the model was extended beyond the range of I
L __ _-___ ________
TR-078 Rev. O Page 98 validity as specified in the RETRAN SER; howevar, the close comparison to data demonstrates the model adequacy for the case when the pressurizer fills up. For the reload transients, the pressurizer never fills up or empties and the pressurizer spray is conservatively assumed not to be available.
Previous comparisons have demonstrated non-equilibrium to be more conservative than equilibrium and the close comparison to data justify its application.
7.1.3 Algebraic Slip Model s
The algebraic slip model is used to determine the phase velocity differences for all applications. This is the more conventional approach and used more extensively in the industry. The turbine trip / reactor trip transient was analyzed with both algebraic and dynamic slip options. The differences were small as shown on Figures 7-1 through 7-5 with algebraic slip being more stable and having a faster computer run time.
7.2 Computer Code Uncertainties The evaluation of uncertainties in the computer code (e.g.,
correlations, numerical algorithms) is beyond the scope of this report, and has been done as part of the RETRAN code development in separate effects analyses and systems effects analyses of the correlations and code models. The models were concluded to be
~
TR-078 Rev. O Page 99 acceptable by the code developer. These correlations, algorithms, etc. were reviewed by the NRC as part of the generic review of the RETRAN code and found to be acceptable.
7.3 Nodalization 7.3.1 RCS and Steam Line Nodalization The nodalization scheme is described in detail in Section 5.
The nodalization is based on GPUN in-house experience over the years. The RCS piping uses a single node representation. This is typical modelling practice as demonstrated by models used in the industry. One node representation of the piping is sufficient as more detailed nodalization would only be required to follow a front such as cold water flowing down the pipe.
This phenomena can be modelled by use of the transport delay model in lieu of more nodes. No reload transient is expected to be affected by this phenomena. The TMI-l RETRAN model uses 10 mesh points to simulate movement of temperature fronts. This was found to be adequate by performing a sensitivity study on the steam line break accident, with no difference being observed in the results when using 0 and 20 mesh points. This lack of sensitivity is to be expected as the reactor coolant pumps are not tripped for this event.
The core is represented by three nodes and is also typical of models used in the industry. Since most PWR FSAR transients
TR-078 Rev. O Page 100 are adequately addressed using point kinetics more detailed core nodalization is not required. Similarly, for all reload transients a one node downcomer, bypass, lower and upper plenum has been determined to be adequate. The upper head is modeled as a single volume connected to the annulus and the upper plenum with two normal junctions representing flow paths through the upper guide structure and upper annulus.
This representation is adequate for all the reload transients as upper head voiding is not expected, and the small flows through this region have an insignificant effect on system response.
(The vendor did not model the upper head region in the FSAR analysis.)
The discussion above applies to and justifies the use of three nodes to represent the steam lines. For most reload transients, simplifying assumptions such as constant steam flow do not require modelling of the steam lines. The nodalization is equivalent or better than steam line nodalizations licensed by other utilities. For the Steam Line Break accident, a more detailed steam line nodalization was developed.
7.3.2 SG Nodalization The TMI-1 RETRAN model nodalizes the steam generator secondary tube bundle region with a vertical stack of 12 homogeneous
TR-078 Rev. O Page 101 volumes. The steam generator downtomer is modeled as a separated bubble rise volume. The validation of the steam generator modeling, including secondary heat transfer, consists of transient comparison to data. At steady state, the secondary mass inventory predicted by RETRAN is close to other code-predicted reference values. The amount of superheat at the exit matches plant data.
The TMI-I SG nodalization is similar to that used by Duke Power Company (9), which was extensively validated by comparison to a wide spectrum of Oconee plant transient data. Such an extensive and well documented data base does not exist for TMI-1. No benchmarks to scaled integral test facility data or to separate effects tests have been performed. Most of this data is of a proprietary nature and scaled facility test data has typically focused on LOCA phenomena, for which the RETRAN model is not intended. Another limitation is that the impact of scaling on the validity of data must be addressed. Separate effects tests are mainly useful in the code development process and have been used to qualify particular aspects of the code.
7.3.3 Core Conductor Geometry The base model contains three fuel meshes and three clad meshes.
These values were determined to be adequate for most reload transients by performing sensitivity studies. The startup and
TR-078 Rev. O Page 102 rod withdrawal accidents which result in large increases in fuel temperature were re-analyzed to determine the sensitivity of these parameters, discussed below.
7.3.3.1 Clad Mesh Sensitivity Increasing the fuel clad mesh from three to six resulted in an insignificant change in fuel temperature and three clad meshes were determined to be adequate.
7.3.3.2 Fuel Mesh Sensitivity Increasing the number of fuel meshes from three to six resulted in an increase in initial fuel temperature of eight degrees.
. A further increase from six to twelve changed the temperature by eight degrees, and going from twelve to twenty changed the temperature by only one degree. Thus for those reload transients which result in a large fuel temperature change, twelve mesh points will be used to maximize this change.
7.4 Time Step convergence The iterative solution technique was used for all transients. The very nature of this scheme is aimed at minimizing the computational effort while achieving an acceptable degree of accuracy. Several predictive and accuracy control algorithms select the appropriate
TR-078 Rev. O Page 103 time-step size. If anomalous behavior of the initial solution is perceived, the time advancement is reevaluated. With this scheme the user need only specify the maximum time step size. However, caution must be exercised as too large a time step size could result in code instabilities / errors or in missing important transient phenomena due to obtaining edits at too large a frequency.
Sensitivity studies were performed to obtain an optimum time-step size so as to achieve a balance between accuracy and computer run time.
For long running transients such as the turbine trip / reactor trip and the TMI-2 accident, the maximum time step size was relaxed after the initial phase of the transient during which phenomena such as opening and closing of valves was observed. Thus, typically, a maximum time step size of 0.001 seconds would be used for the first second of a transient, and then relaxed to 0.005 seconds. This scheme was found adequate for the reload transients due to their short duration.
However, a further relaxation may be specified later in the transient for the longer running transients, so as to maintain reasonable computer run time.
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8.0 CONCLUSION
S This report presents the methods for performing system transient analysis for Three Mile Island Unit 1 Nuclear Power Station. A description of the RETRAN model and general code features have been discussed. The accuracy of the model has been demonstrated by comparisons with plant data. The accident analysis methodology and adequacy of the model for licensing applications has been demonstrated by comparison of representative analytical results to vendor calculations. Sensitivity studies further justify the overall model and code options, and assure that these are selected so as to predict a conservative response. This demonstrates GPUN's ability to perform operational and licensing analysis of TMI-1, using RETRAN-02 MODS.I.
TR-078 Rev. O Page 110
9.0 REFERENCES
- 1. TR-045-A, BWR-2 Transient Analysis model using the RETRAN Code, November 1988.
- 2. RETRAN-02-A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems - EPRI NP-1850-CCMA. .
- 3. C. O. Thomas, Chief, Division of Licensing, NRC, to Dr. T. W. Schnatz, Chairman, Utility Group for Regulatory Application, NRC Safety Evaluation Report of RETRAN, September 2,1984.
- 4. Jim Harrison, Technical Director, WOTEC, Inc.
- 5. TMI-1 Final Safety Analysis Report.
- 6. Martin J. Virgilio, Acting Director, Division of System Safety and Analysis, NRC to C. R. Lehmann, Chairman, RETRAN Maintenance Group
" Acceptance for Referencing of the RETRAN-02 MOD 5.1 Code," April 12, 1994.
- 7. "WREM: Water Reactor Evaluation Model (Revision 1)" NUREG 75/056, May 1975.
- 8. Inconel alloy 600, Huntington Alloys Reference Book,1978 edition.
- 9. Qualification of the Oconee RETRAN model by comparison with plant transient data, Nuclear Technology, Volume 83, December 1988.
_ _ _ _ - _ _ - - _ _ _ - _ _ _ . .