ML20210S170
| ML20210S170 | |
| Person / Time | |
|---|---|
| Issue date: | 05/05/1986 |
| From: | NRC |
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| References | |
| NUDOCS 8605200454 | |
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Text
l TOPICAL REPORT EVALUATION OF PTS-PWR2 Report Nos, and
Title:
XN-NF-74-5, Rev. 2, " Description of the Exxon Nuclear Plant Transient Simulation Model for Pressurized Water Reactors (PTS-PWR)" October 1983.
Originating Organization:
Exxon Nuclear Company, Inc.
l 8605200454 860505 PDR TOPRP ENVEXXN C
O I.
BACKGROUND On May 15, 1975, the Exxon Nuclear Company submitted topical report XN-74-05 Rev. 1 for staff review and approval.
XN-74-05 Rev. 1, " Description of Exxon Nuclear Plant Transient for Pressurized Water Reactors" (PTSPWR), documents the analytical equations and assumptions used in the PTSPWR computer program. The PTSPWR computer program was designed to analyze FSAR Chapter 15 (as defined in NUREG-0800, "The Standard Review Plan") non-loss of coolant accident (LOCA) events.
Since the initial submittal of XN-74-5 Rev. 1, the PTSPWR computer program has undergone extensive modification and validation.
This is documented in XN-74-5(P) Rev. 2, Supplements 1 through 6 of Rev. 2 (see References 1 through 6) and in XN-NF-CC-38 Supplement 1 (Ref. 7).
As a consequence of the modifications to the PTSPWR program, the code was renamed PTSPWR2.
1 This SER addresses the PTSPWR2 computer program and its application to FSAR Chapter 15 non-LOCA transient analysis.
The staff's review has concluded that selected licensing calculations of Chapter 15 non-LOCA transient events are acceptable if analyzed using PTSPWR2 with appropriately conservative input /
1 methodology assumptions.
II.
SUMMARY
OF PTSPWR2 AND THE TOPICAL REPORTS Topical report XN-NF-74-5(P) Rev. 2 describes the PTSPWR2 code and its applica-bility to calculate transient and accident events for both Westinghouse and Combustion Engineering (CE) designed nuclear steam supply systems (NSSS).
The report documents the governing equations and their numerical solution for the various models.
Report XN-74-5(P) Rev. 2, Supp. 1 details the model modifica-tions required to simulate Combustion Engineering 2 x 4 plants.
Application of the code to Chapter 15 transients and accidents is discussed in XN-74-5(P) Rev. 2 Supplement 2.
Supplement 3 provides responses to NRC questions and provides corrections to the original topical report. Additional updates are provided in Supplement 4.
Benchmark comparisons performed by Exxon to qualify PTSPWR2 for Chapter 15 transients and accidants are detailed in 1
Supplement 5.
An update to the steam generator heat transfer model and an additional benchmark calculation are contained in Supplement 6.
Revisions and benchme.:s to the pressurizer model are described in XN-NF-CC-38 Sup. 1.
PTSPWR2 is a fixed nodalization code with two primary loops, each modeling a U-tube steam generator.
The primary system components modeled include the ves-sel plena, the reactor core, the primary coolant loops, the pressurizer, the steam generators, and the reactor coolant pumps.
The secondary side components include the riser section of the steam generator, the downcomer, the steam separators, the steam lines with their associated systems, and the feed water system.
The code also models plant control and protection systems.
Point kinetics is used for modeling the neutronics.
The thermal hydraulics model as-sumes one-dimensional HEM (homogeneous equilibrium model) flow with the addi-tional assumptions of incompressibility and quasistatics in specific volumes.
Mixing models and transport delay models are used in the plena and the loops.
Water properties are taken from the 1967 ASME steam tables.
Control system models are provided to describe plant protection and control systems.
Boron transport through the coolant loops is also described.
Component models include a two region non-equilibrium pressurizer, reactor coolant pumps, and a fuel rod model.
An automatic steady state initialization procedure is available.
PTSPWR2 is not utilized to predict localized heat transfer phenomena such as hot channel DNBR.
This is done by inputting core flows and inlet thermodynamic conditions from PTSPWR2 into a detailed core heat transfer model.
The XCOBRA III B code (Ref. 8) has been approved for this purpose.
III.
SUMMARY
OF REGULATORY EVALUATION A.
Review Approach This evaluation assesses the acceptability of PTSPWR2 to perform FSAR Chapter 15 licensing analyses for the following events found in the Standard Review Plan for both Westinghouse and CE plants.
2
SRP Section 15.1.1.
Decrease in feedwater temperature 15.1.2.
Increase in feedwater flow 15.1.3.
Increase in main steam flow 15.1.4 Inadvertent Opening of Steam Generator Relief / Safety Valve 15.2.1.
Loss of external load 15.2.2.
Turbine trip 15.2.3.
Loss of condenser vacuum 15.4.4.
Closure of Main Steam Isolation Valve 15.4.5 Steam Regulator Failure 15.2.6.
Loss of normal AC power 15.3.1.
Total and partial loss of forced reactor coolant flow 15.3.3.
Reactor Coolant Pump Rotor Seizure 15.3.4 Reactor Coolant Pump Shaft Break 15.4.1 Uncontrolled control rod assembly withdrawal from sub-critical or low power condition 15.4.2.
Uncontrolled control rod assembly withdrawal at power 15.4.3.
Loop Simulation for control rod misoperation 15.4.4 Startup of an inactive loop 15.4.5 Flow controller malfunction 15.4.6.
Baron dilution at power 15.4.8.
Loop simulation for control rod ejection 15.5.1 Inadvertent ECCS Operation 15.5.2 Excessive charging flow 15.6.1 Inadvertent Pressurizer Relief Valve Opening To support our evaluation, we have applied the acceptance criteria presented in the respective sections of the NRC Standard Review Plan (NUREG-0800, July 1981 Revisions). We have reviewed the applicant's supporting derivations and experimental data and have made audit calculations using the RELAP5 computer code. Our conclusions regarding the use of the PTSPWR2 code are stated in the Staff Position section of this report.
B.
Review of Analytical Models 3
~
The fixed nodalization two loop thermal hydraulic node / flow path network used by PTSPWR2 is shown in Fig.'1.
The reduced noding arrangement of PTSPWR2 causes some approximations.
On the primary side, the pressurizer is connected directly to the upper plenum and no separate volume is provided for the relatively stagnant upper head region.
In addition, no distinct volumes are provided for the downcomer and core bypass regions. While Fig.-1 depicts no pressurizer surge line, a model for pressure loss across. the surge line was added during the course of this review.
Sensible heat for primary and secondary system metal is not described.
These simplifications limit the range of application of the code.
i Other aspects of the model provide separate volumes for the core, upper plenum, lower plenum, inlet and outlet steam generator plena, U-tubes, and the cold and hot legs.
The pressurizer is a special non-equilibrium volume. Additions are made to the nodalization for Combustion Engineering (CE) 2 x 4 plants to allow for dual cold legs per loop.
On the secondary side, the steam generators are represented by volumes for the downcomer, tube region, upper plenum, and steam dome.
Each steam line is represented by two volumes connecting the common header volume.
Core Neutronics Model The point k etics equations of the neutronics are used with six delayed neutron precursor groups.
The prompt jump approximation is used for small reactivity changes.
The default set of delayed neutron parameters programmed into the code corresponds to those utilized by the staff.. Reactivity feedback from fuel temperature (Doppler) and moderator temperature and pressure changes are modeled. Reactivity contributions from control rod movement, as an input function of time and boron concentration can also be modeled.
As a consequence of modeling the core as one axial node, the reactivity feedbacks from the moderator and fuel are computed from average temperatures with no spatial weighting.
An attempt is made at refining spatial shapes by assuming that the moderator temperature profile follows the integrated power axial shape, thereby calculating an average temperature from that profile.
Exxonfresented 4
calculations which demonstrate that the moderator reactivities from this model differed by no more than a percent for both quasistatic and fast transients from those of a multiaxial node model using flux squared weighting.
In the case of the Doppler feedback, the reactivity can be off by as much as five percent for quasistatic transients and as much as thirty percent for fast transients, depending upon the range of axial peaking factors used.
The reactivity weighting procedures are sensitive to peaking in the respective temperature profiles.
However, it should be noted that uncertainties in neutronics data are of the order of 20% and that the Exxon calculations show that effective Doppler temperature changes are underestimated by the PTSPWR2 model which produces higher power on heat ups.
PWRPTS2 will not be used to evaluate rapid cooldown events (steam line breaks).
The staff finds the point kinetics model in PTSPWR2 satisfactory with respect to the equations and reactivity feedback model.
The use of the prompt jump ap-proximation should be restricted to step reactivity increases of less than
$0.20, for prompt neutron lifetimes greater than 10-3 seconds and less than
$0.40 for lifetimes greater than 10 4 seconds.
Decay heat is calculated using the model in Branch Position APCSB 9-2 of the NRC standard review plan.
The decay constants are fits to the ANS 1971 standard data with additional margin to account for uncertainties.
The contribution from heavy element decay is included.
This model is acceptable to the staff.
Thermal Hydraulics Only single phase flow is permitted within the core and the coolant loops, which places limits on the range of code application.
The primary coolant loops are further assumed to be incompressible.
This assumption maximizes the surge flow into the pressurizer from thermal expansion of the coolant. Within the coolant loops gravity heads, hydraulic losses and inertia are accounted for in the momentum equations which are solved to determine loop flow.
5
Coolant system pressure loss factors are initialized as constants and remain constant during the calculation.
This implies that in the case of flow coastdowns or decreasing flows the effect of increasing friction factors will not be seen. Much of the reactor system pressure loss results from form losses which are little affected by changes in flow rate, however.
The flow in each loop is solved separately.
Pump heads are computed using the homologous pump model.
The loop pressure is determined by the pressurizer model.
During this review Exxon modified the surge line model to include the line friction drop with a constant friction coefficient and a gravity head but no inertia term.
- However, no iteration is performed between this model and the pressure in the primary loop.
The surge flow therefore does not see the line resistance and is based only upon the unconstrained expansion rate in the primary system over a time step.
This approximation maximizes the pressure response to thermal expansion and contraction within the coolant loops.
The thermal energy balances are treated differently depending upon the section of the primary system being considered.
In the legs a transport delay model is used.
The transport delay model divides the hot leg into six fixed segments and the cold leg into 30 fixed segments.
A front is tracked through the incom-pressible segments as the velocities change.
Separate effects calculations were presented by Exxon which demonstrated that trends in energy and mass propagation are properly modeled.
In the steam generator and vessel plena, homogenized single node energy balances are used.
The one time constant mixing models assume a predeninant flow direction in order to determine the plenum mixing time constant.
The core coolant thermal behavior is treated quasi statically.
The coolant enthalpy distribution for the ten axial nodes is i
computed on the assumption that the distribution follows the integrated power shape up the channel.
For the core pin to coolant heat transfer, the single phase Dittus Boelter correlation is used normalized to an input initial heat transfer coefficient.
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Core heat conduction is calculated by a one dimensional radial heat transfer calculation using the Fourier conduction equation with a uniform power profile and temperature dependent thermal properties in the pellet but constant proper-ties in the cladding.
The pin is divided into four equal volume fuel radial nodes and one cladding node with the appropriate node center to node center conductances. The fuel / clad gap is treated by a constant gap conductance as all dimensions are assumed fixed during a transient.
The radially averaged fuel temperature is used for the Doppler feedback.
The fuel rod model is found acceptable for applications when core DNBR is not directly calculated by PTSPWR2.
The steam generator model was modified during the review.
The present model assumes that the primary side of the tubes is always single phase liquid and the secondary side is always two phase and saturated.
The heat transfer coef-ficients on the secondary side remain constant at the steady state values throughout the transient.
This assumption is acceptable as long as the steam generator tubes rema'ined covered.
The heat transfer on the inner tube walls within the primary system is calculated using the flow dependent Dittus-Boelter correlation. This flow dependence allows PTSPWR2 to be used for analysis of flow coastdown transients as will be discussed further in the section on code audits.
Steam flow out of the steam generators into the steam line volumes is assumed to be saturated steam.
Complete separation of steam is assumed so that only steam enters the steam lines and any liquid water remains in the steam genera-tors.
The steam line momentum balances include the momentum fluxes, the fric-tional drops and the inertia terms.
Steam flow out of the various valves is calculated using the valve model discussed in the following section.
The code utilizes saturated thermodynamic properties and derivatives in the pressure range 100-3100 psi which are taken directly from the 1967 ASME steam tables.
In the subcooled region the pressure dependence of the thermodynamic properties is neglected within the coolant loops.
Below the pressure of 2200 psi, the 2200 psi subcooled properties are used; in the range from 2200-2300 psi the 2300 psi properties are used while from 2300 psi to 3100 psi the 7
i
3100 psi properties are used.
Absolute errors by making such an approximation j
are of the order of ~ 1% which is acceptable.
For the pressurizer model the thermodynamic properties of water are utilized in comprehensive form with no approximations assumed.
Both the superheated and subcooled region are treated in the form of two dimensional tables.
Components (a) Pressurizer Reactor system pressure is determined by the pressurizer model.
A two region (vapor and liquid) non-equilibrium model was developed with each region being single phase and homogenized.
The inter-region mass and energy transfer terms model the condensation and bulk flashing processes.
Convective heat transfer is user determined between the two regions. The model provides a reasonable approximation for FSAR transients.
Boundary conditions include a heater model and a pressurizer spray model which is user controlled.
The fraction of spray deposited in either region is input.
Heat loss to the pressurizer boundary is not assumed which should produce pres-sure overestimates.
Based on the technically acceptable analytical equations and qualification work discussed later in this report the staff approves the use of :he PTSPWR2 pressurizer model with a zero inter-region heat transfer coefficient which is conservative.
(b) Reactor Coolant Pumps The reactor coolant pump model utilizes four quadrant homologous torque / head curves in single phase.
The pump heat is input and is added to the core heat.
The pump speed equation incorporates electrical torque (set to zero for coast-downs), hydraulic torque (using the homologous torque curves) and friction losses defined as a linear polynomial in the speed. Thus a coastdown including 8
i l
assumed sheared shaft can be simulated.
An option is also available for modelling the pump as a geometric loss (locked rotor).
(c) Valves Both safety / relief valves and isolation valves are modelled allowing for hysteresis, different opening and closing rates and partially open positions.
The Exxon critical flow correlation for steam flow out of relief and safety valves compares wall with the Murdock-Bauman correlation typically used for pure steam flow.
Liquid critical flow is not modeled.
Special Purpose Models (a) Control and Safety Systems The plant protection trip options include both Westinghouse and Combustion Engineering design logic.
For Westinghouse plants the trip signals include high nuclear power, overtemperature AT and overpower AT.
Additions made for Combustion Engineering plants include the asymmetric steam generator protection trip, the variable high power trip, the thermal margin / low pressure trip and the local power density trip.
The coding has been modified so that each trip signal utilizes a unique scram delay time.
The remaining control system models include pressurizer level and pressure controls, the feedwater controller, tur-bine pressure regulator, steam dump system, and a rod speed controller.
(b) Boron Transport Transport from the point of injection to the lower plenum is determined by input as the model assumes injection in the lower plenum with an input time 1
delay.
A table of head versus flow is used to simulate the injection pump.
Transport in the various plena and legs is the same as for the loop thermal hydraulics.
As the concentration is tracked in terms of ppm, errors in coolant density do not translate into inaccuracies in the boron concentration.
The model is therefore approved.
9
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1 Initialization i
Automatic initialization of the secondary system is carried out by adjusting feedwater flow rates and bundle exit qualities to be compatible with input data for T,y, dome pressures, steam generator water levels and feedwater enthalpies.
Steam generator heat transfer coefficients and primary side temperatures are extracted from the design conditions of loop flow and power.
Adjustments are made to coolant pump speed to be compatible with the loop flow resistances.
The steam flow is initialized with the adjusted feedwater flow and steam line friction.
The pressurizer is initialized with the input pressure.
The out-j surge is set equal to the adjusted spray flow rate.
Numerical Technique i
The numerical techniques employed are based upon first order explicit schemes and are not unconditionally stable. Modifications were made to overcome in-l stabilities during benchmark comparisons.
These modifications restrict tem-
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perature changes during a single time step.
C.
Code Qualification 4
Exxon,has presented qualification work in the form of three separate cate~gories:
(1) audit calculations using RELAPS, (2) comparisons with system i
effects experiments and (3) a benchmark against real plant data.
While each category has certain limitations, the overall combination of audit calculations, benchmarking against systems effects experiments and comparisons with real plant data provides justification for the overall acceptability of f
the various code models as applied to specific transients.
The audit l
calculations were performed for a flow coastdown event (three pump coastdown) 1 using an H. B. Robinson plant model (three loop Westinghouse). The comparisons-l against data took the form of analyses of three LOFT tests; L6-1 (loss of load)
L6-2 (pump coastdown) and L6-3 (excess load).
These comparisons therefore cover both heatup and cooldown et ets in addition to testing the models for the 1
pump coastdown against measured du.a The benchmark calculation of the Prairie 1
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Island steam generator tube rupture incident provides a measure of scaling effects in going to real plant scale.
Audit Calculations Results from a detailed RELAP5 model of the H. B. Robinson plant were presented by Exxon for a three pump coastdown event.
Differences in initial conditions appear to be of the order of a percent or less whereas differences in boundary conditions are minimal.
The audit calculation shows good agreement for the core power behavior and reasonable agreement for the loop flow coastdown.
The differences in coolant temperature are only a few degrees (see Figures 2 and 3).
It can be seen from Fig. 4 that the PTSPWR2 pressurizer model is over-reactive.
Both insurge and outsurge behavior are off by ~ 100 psi.
This behavior can be attributed to the forcing of the inter-region heat transfer to zero, the lack of compressibility modeling of the subcooled water in the coolant loops and the lack of a primary system metal heat flow model in PTSPWR2.
It should be con-cluded that the pressurizer model will systematically overpredict the extremes in the pressure behavior both during insurges and outsurges.
LOFT Experiments (a) LOFT L6-1 The L6-1 loss of load test is a heatup event.
During the first loop cycle time the steam flow rate and feedwater flow rate are well matched in the comparisons against data. While the initial conditions are offset slightly the hot leg flow is in reasonable agreement.
The comparison for core power is within a couple of percent. During this initial period of the first loop cycle time both the hot leg and cold leg temperature are overpredicted.
This is due in part to the overprediction of the steam generator pressure.
Conversely on the downswing during the reopening of the steam valve the rate of decrease in cold leg and hot leg temperatures is overpredicted (see Figure 5).
The pressurizer t
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model is overreactive; however, in this case the outsurge is markedly underpre-dicted while the insurge is reasonably well matched.
The excessive decrease in pressurizer pressure and reactor system temperature probably results from the lack of a primary system metal heat transfer model in PTSPWR2 whereas the LOFT test facility contains an untypically large metal mass for the amount of reactor coolant.
(b) LOFT L6-2 This system transient test includes a pump coastdown similar to the RELAP5 audit calculation.
Both power and primary flow are reasonably predicted by PTSPWR2 until pump coastdown is complete.
I Complications then arise which in part may be attributed to limitations in LOFT I
instrumentation or to deficiencies in the simulation, both in modelling and in specific input used.
The boundary conditions of feedwater flow and steam flow are reasonably matched.
For the first loop cycle time the loop hydraulics ap-pear to be well simulated by the PTSPWR2 pump model and hydraulics model.
Fig.
6 shows the cold leg temperature which is predicted to within 5 F of the test data.
As with the RELAPS H. B. Robinson audit calculation the pressurizer model demonstrates an oversensitivity both on insurges and outsurges.
(c) LOFT L6-3 This excess load test is a sample of the cooldown class of events and tests the code models in a range different from L6-1 and L6-2.
It was initiated at LOFT by ramping the main steam flow control valve MFECV to full open.
The valve was closed at 37 seconds.
Reactor trip occurred at 15.6 seconds. Good matches are obtained for the boundary conditions of feedwater flow and steam line flow.
There is also reasonable agreement on the power and the primary side hot leg flow. The steam generator dome pressure is well matched in the period when the MSFCV is open (see Figure 7). When the valve is closed PWR-PTS 2 which does not predict heat transfer from the steam generator shell to the secondary coolant 12
under predicts steam pressure.
Both the cold and hot leg temperature predictions are low.
This difference again probably results from the lack of metal heat transfer in the model and the relatively large metal mass in the test. The LOFT L6-3 benchmark also shows over-reactive behavior both on upswings, and down-swings for the pressurizer model.
The predictions of PTS-PWR2 are acceptable in the time of interest for DNBR calculations.
PWR Comparison To resolve questions regarding the effect of scaling, comparisons against real plant data are helpful.
Exxon has submitted such a comparison:
a benchmark against the Prairie Island reactor trip of October 2, 1979.
The reactor trip resulted from low reactor system pressure following a steam generator tube rupture.
The loss of inventory resulted in depressurization, load runback and finally reactor scram. Turbine trip on low pressurizer pressure produced a loss of load transient and a challenge to the validity of the PTSPWR2 models.
As is common for real plant experience the data are limited and provide some inconsistencies.
There are specific deviations but the trend in data appears to be in qualitative agreement.
Quantitative conclusions are difficult to draw.
There are definite discrepancies in cold leg temperature comparisons.
The loop T the loop AT and the hot leg temperature, appear to be in good 3y, agreement (see Figure 8).
D.
Limitations on PTSPWR2 for Transient and Accident Analysis (a) The PTSPWR2 steam generator model does not account for the effect of steam generator tube uncovery on secondary system heat transfer.
The code therefore cannot be used for analyses of large steam or feedwater line break events or loss of feedwater events for which the steam generator tubes might become uncovered without further justification and staff review.
(b) The relief and safety valve flow model is not designed to calculate flow rates when the inlet is liquid or two phase.
The code therefore cannot be used when the steam lines or pressurizer become flooded to the valve l
13
inlets without further justification and staff review.
Further justifica-tion should also be provided for the adequacy of steam generator or pres-surizer pressure calculations if flooded conditions are calculated.
(c) Because PTSPWR2 requires the primary loop (except for the pressurizer) to be subcooled, use of the code is not acceptable whenever the core outlet temperature exceeds saturation.
Since coolant loop saturation is likely to occur if the pressurizer becomes drained of liquid, the code should not be used if a zero pressurizer level is calculated without further justification and staff review.
(d) Staff review did not include the ability of PTSPWR2 to implicitly calculate heat transfer in the core hot channels; therefore, for all events in which minimum DNBR is an issue, an acceptable hot channel thermal hydraulic code such as COBRA III B must be combined with PTSPWR2 for calculating DNBR.
(e) Analyses utilizing PTSPWR2 should be made conservative by selecting con-l servative input.
Exxon methology for selecting conservative input is de-l scribed in XN-NF-84-73(P) Revision 2 " Exxon Nuclear Methodol'ogy for Pressurized Water Reactors:
Analysis of Chapter 15 Events," March 1986.
This document is currently undergoing staff review for application to Westinghouse 3 and 4 loop PWRs.
Approval is expected without>significant modification.
PTSPWR2 should be utilized for licensing actions only as described in XN-NF-84-73(P) Revision 2 or the latest appfeved version of that document.
(f) Since PTSPWR2 does not contain a space dependent core'neutronics model, local power peaking must be calculated separately for events producing high local peaking.
Acceptable methologysis described in XN-CC-28 for control rod misoperation and XN-NF-78-44 forsc'ontrol rod ejection.
l (g) The staff reviewed the adequacy of the PTSPWR2 for analysis of the tran-sients and accidents listed in Section III.A for Westinghouse and Combus-tion Engineering pressurized water reactors with inverted U-tube steam
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generators.
Applications of the code for other reactor types will require additional justification and staff review.
IV. QUALITY ASSURANCE AUDIT The NRC staff conducted audits of Exxon quality assurance procedures for com-puter code development on April 3-5, 1984 and again on July 8-12, 1985. Two nonconformances related to the PTSPWR2 code were identified.
One item concerned the lack of proper sign off on a calculation.
This was resolved and accepted by the NRC staff.
The other item concerned engineering training.
Exxon has instituted a training program which deals with PWR safety analysis.
The staff finds the quality assurance procedures applied in the development of PTSPWR 2 acceptable.
V.
STAFF POSITION We have reviewed the methods and assumptions described in XN-NF-74-5 Rev. 2 in-cluding the 6 supplements and have concluded, subject to the limitations stated in Section III.D of this report that the PTSPWR2 code contains acceptable mod-els and methods for transient analysis of the specific events listed in Section III.A.
The adequacy of the results obtained with PTSPWR2 will be highly dependent on the adequacy of the plant parameters and input into the code.
For this reason,the staff requires that results obtained with the PTSPWR2 code be accompanied with appropriate descriptions and justifications for the code input. We require that topical report XN-NF-74-5(P) be reissued to include all revisions and corrections as described in the supplements or otherwise identified by Exxon during the review.
1
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l 15 J
s REFERENCES e-(1) XN-74-5(P), Revision 2, Supplement 1, " Description of the Exxon Nuclear Plant Transient Simulation Model for Pressurized Water Reactors (PTS-PWR):
PTSPWR2 Modifications for St. Lucie Unit 1," Exxon Nuclear Co.pany, Inc.,
Richland, WA, October 1983.
4 (2) XN-74-5(P), Revision 2, Supplement 2,'" Description of the Exxoa Nuclear 7
Plant Transient Simulation Model for Pr,essurized Water Reactors (PTS-PWR):
Methodology and Application,'1 Exxon Nuclear Company, Inc., Richland, WA, January 1984' J
(3) XN-74-5(P), Revision 2, Supplement 3, " Description of the Exxon Nuclear Plant Transient Simulation Model for Pressurized Water Reactors (PTS-PWR):
Responses to NRC Question's," Exxon Nuclear Company, Inc., Richland, WA, September 1984.
(4) XN-74-5(P), Revision 2, Supplement 4, " Description of the Exxon Nuclear Plant Transient Simulation Model for Pressurized Water Reactors (PTS-PWR):
Code Updates
" Exxon Nuclear Company, Inc., Richland, WA, August 1984.
(5) XN-74-5(P), Revision 2, Supplement 5, " Description'of the Exxon Nuclear Plant Transient Simulation Model for Pressurized Water Reactors _(PTS-PWR):
Code Benchmarks," Exxon Nuclear Company, Inc., Richland, WAn August 1984.
(6) XN-74-5(P), Revision 2, Supplement 6, " Description of the Exxon ~ Nuclear Plant Transient Simulation Model for Pressurized Water Reactors (PTS-PWR):
Revised PTSPWR2/RELAPS Benchmark Analysis," Exxon Nuclear Company, Inc.,
Richland, WA, November 1984.
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(7) XN-NF-CC-38, Supplement 1, " Users Manual for PTSPWR2 a FORTRAN Program for Simulation of Pressurized Water Reactor' Plant Transients," Exxon Nuclear
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Company, Inc., Richland, WA, October 1979.
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~ XN-NF-82-21(P)(A), " Application of Exxon Nuclear Company PWR Thermal
- (8)
Margin Methodology to Mixed Core Configurations," Exxon Nuclear Company, Inc., Richland, WA, December 1984.
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LP 4%I# Ptit#Vl8 (14 D 114.
R.41f13 IEJittlT4
=ATI*
I
@ tuf 6dt Pt Ault tumuu sh a thfAAf10r T14 Ditr.v h it0W P!ilift. art FIGURE 1 PWR MODEL BLOCK DIAGRAM
l 1
m PTS /RELRPS COMPARISON icono 1
NME-
--- Original Pi5PWR2 Senchark Results O
(RN.74 5(P), Rev. 2. Supp. 5)
R(LAP 5 Results Rev,e.,,5P.R, en..,,
3, (Dittus Scelter HTC) b!
E E
,ooo.
d I
[ 3o00 -
l y
o o
ao to Go to too iao ato ato ago zoo T!rtC (SEC)
Figure 2:
Flow Coastdown PTS /RELRP5 COMPRRISON 35 l
C $@-
U 8_
~
_ e.t.
Sto -
--- original Pt5PWR? BencMark Results b
(Rh.74 5(P), Rev. 2, Supp. 5)
RitAPS Results g
Revised P15PWR2 Benc M ark H
(Dettus-80elter MIC) g, bsoo o
ao to so so ico nao sto iso iso zoo Tit 1E (SEC)
Figure 3: Cold Leg Temperature
PTS /RELRP5 COMPRRISON m
h 2000 -
Q M-5
-6 A
2500 -
y g,,..,.,......................
--- Origina) PISFWR2 Benchmark Results E
(14 74-$(P), Rev. 2. Supp. 5) mus Rewits M
u.is..a rispWR7 s,nce
.,6 g 3 00 (inito.uneiter nic)
L 3500 0
30 to e0 e0 00 na0 sto ne0 se0 300 Tine (SEC)
Figure 4:
Pressurizer Pressure LOFT TEST L6-1 BENCHMRRK 600 N
'E
.s 5"~
LEGEND
'TSPWR2
?.
.. M.I.PMO...
a e ~.
8 560 a
~.
g
.. m.,..... :...:'
,;...c,<.,
c.,...,.. w.v ----- -
c E sto-y Iu 6
G t.j s20 -
Uz sm 0
20 to 60 e0 100 120 360 160 te0 200 TinEISEC)
Figure 5: LOFT L6-1 Hot Leg Temperature w
\\
LOFT TEST L6-2 BENCHMARK s00 se0 -
LEGEND JP SPWR2
..L.uT.I.pg10..
g o8 sso -
J U
5 sto -
b b.. - ---
w ---......................._
s e
S 550 -
Q oo 500 0
20 to 60 00 100 120 tto 160 180 200 TinctSCC)
Figure 6: LS-2 Cold Leg Temperature LOFT TEST L6-3 BENCHHRRK 1 00 1000 -
LEGCND
- TSPWR2
.. M.L.Pf110...
c g s00 -
c y,,.
p ay 700 -
/
u 8
l a
cc 600 -
M m
500 0
20 to 60 00 100 120 110 160 100 200 TinEtSCC)
Figure 7: L6-3 Steam Pressure
9
~
PTS - PRRIRIE ISLRND BENCHMRRK soo Sa -
LEGEND
'"S2WR2 L.
O
)ATA N
w w & uae u uuv uO Q
E-sta -
- t2:=re e
8a s20-500 0
50 100 150 200 250 300 350 too Tine (SEC)
Figure 8:
Prairie Island LOOP Tavg
_ - _ - _ _ _ _ _ _