ML20046D158
| ML20046D158 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 02/28/1993 |
| From: | VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | |
| Shared Package | |
| ML20046D156 | List: |
| References | |
| NUDOCS 9308160223 | |
| Download: ML20046D158 (100) | |
Text
{{#Wiki_filter:VIRGINIA POWER SYSTEM TRANSIENT ANALYSIS USING VERSION 1 OF THE NORTH ANNA RETRAN MODEL NUCLEAR ANALYSIS AND FUEL NUCLEAR ENGINEERING SERVICES VIRGINIA POWER February 1993 93081602?3 930810 ,5 PDR ADOCK 0500033G J P PDR _
t TABLE OF CONTENTS , Title Page Table of. Contents .......................................... 2-List of Tables ............................................. 3 1.0 Introduction .......................................... 4 2.0 Model Description ..................................... 7 3.0 Model Qualification ................................... 13 , 3.1 FSAR Transient Comparisons ...................... 15 { 3.1.1 Rod Withdrawal at Power .................. 16 3.1.2 Rod Withdrawal from Subcritical .......... 17
- 3.1.3 Flow Coastdown ........................... 18 .
3.1.4 Loss of External Electrical Load ......... 20 3.1.5 Loss of Normal Feedwater ................. 20 - 3.1.6 Locked Reactor Coolant Pump Rotor ........ 23 3.2 Comparison to Plant Data - Steam Generator Tube Rupture Event ................................... 25 i 3.3 Miscellaneous Test - Turbine Runback ............ 27 4.0 Conclusions ........................................... 28 5.0 References ............................................ 29 Appendix - Figures ......................................... 30 r P i t 2 i
F t 6 i r LIST OF TABLES Table Title Page ; 1 Control Volume Summary ............................... -10 2 Momentum Junction Summary ............................ 11 . I t 6 I s 3
i 5 i
'I 1.O INTRODUCTION '!
i RETRAN is a transient thermal-hydraulics computer code developed [ for utility use by the Electric Power Research Institute-(EPRI). Virginia Power uses the RETRAN computer code in the performance of t non-LOCA system transient analyses for the North Anna. and Surry ! Nuclear Power Stations. The NRC approved the Virginia Power i topical report VEP-FRD-41-Alfor use of the RETRAN code on April 11, f 19852 . The analyses presented in VEP-FRD-41-A were performed using [ version RETRAN01 of the code. Since then Virginia Power has also ! performed transient analyses using version RETRANO2. A transmittal ! justifying the use of RETRAN02 through comparison with RETRAN01 was 5 subsequently provided to the NRC on November 19, 1985 via letter Serial No. 85-7533, t Virginia Power has recently updated the RETRAN modeling of'the .} Nuclear Steam Supply System (NSSS) for the' North Anna Power ~ Station. This update, hereafter referred to as Version 1 of.the' ! North Anna RETRAN model, is a revision to-the RETRAN modeling of [ North Anna as described in topical reportVEP-FRD-41-A. A similar ! revision to the RETRAN NSSS model for Virginia Power's Surry Power. ! Station is in progress. ; f The purpose of this report is three fold: 1) to. describe the i s updated North Anna RETRAN model, 2) to summarize .the major. j differences between it and the earlier North Anna RETRAN model,-and
- 3) to summarize the qualification of the updated North Anna RETRAN j
j model. Major differences between Version 1 and the earlier North _ Anna RETRAN NSSS model are: J [ s
- 1. Version 1 is available in four different geometric '
configurations including a multi-node modeling of the~ secondary. side of-the steam generators.
~
The geometric configurations-~are: ,)
- a. One -loop RCS, single node steam generator secondary.
- b. .Or.e-loop RCS, multi-node steam generator secondary j
- c. ihree-loop RCS, single node steam generatorzsecondary !
- d. Three-loop RCS, multi-node steam generator secondary _
4 i i q
i The earlier model is available in a one-loop and two-loop RCS (reactor- coolant system) geometry with a single node steam generator secondary. No multi-node steam generator secondary i geometry was developed for the earlier model.
- 2. In Version 1, the plant secondary side is modeled out to the condenser steam dumps. The earlier model includes only the ,
secondary side of the steam generators and main steam safety and ; relief valves.
- 3. All essential RPS (reactor protection system) and ESF f (engineered safety features) functions are included in Version 1. I The status of the RPS functions can be monitored even though the corresponding reactor trip may be disabled for a particular ;
application.
- 4. Version 1 contains an improved Doppler feedback model.
S. The replacement steam generator design scheduled for installation in North Anna Unit 1 is modeled by Version 1. Sensitivity studies have shown that the differences in the modeling of the steam generator design (original versus replacement) have no significant impact on the results of safety analyses for the plant. Therefore, the updated RETRAN model as designed -is equally applicable to both North Anna units.
- 6. Version 1 includes more realistic modeling of the pressurizer and main steam line PORVs (pressure operated relief valves) and '
safety valves. The individual main steam safety valves are now t represented explicitly instead of being modeled as a single valve for each steam line as in the earlier version. !
- 7. Version 1 RCS loop temperature input for control systems <
reflects that of the median Tavg instrumentation being implemented ' at both North Anna units upon removal of the RTD bypass lines.
- 8. All Version 1 model input has been calculated based on the
~
latest engineering drawings and design specifications to accurately ! reflect existing or planned unit designs. These changes provide for increased flexibility in performing the ; full range of non-LOCA accident analyses for which the RETRAN code P 5 t
is used, and in supporting plant operational issues. All calculations employing the Version'l model use EPRI version i RETRANO2 Mod 5 of the RETRAN code. The implementation of Version 1 of the North Anna RETRAN model does not entail any change to Virginia Power's core reload design methodology as described in Virginia Power's topical . report' VEP-FRD-424, including the accident basis assumptions for North Anna UFSAR Chapter 15 transient analyses. All figures referred to in the text are provided in the Appendix. -
,I s
1 6 9 i i i 6-l i
]
2.0 MODEL DESCRIPTION The model is available in four geometric configurations:
- a. One-loop RCS, single node steam generator secondary l
- b. One-loop RCS, multi-node steam generator secondary
- c. Three-loop RCS, single node steam generator secondary
- d. Three-loop RCS, multi-node steam generator secondary l The primary side noding of both the single node and multi-node steam generator configurations are identical; i.e., both have ten steam generator tube volumes and ten heat' conductors per steam generator. The earlier RETRAN model contains four steam generator tube volumes and heat conductors per steam generator. The secondary side of the single node steam generator configuration has a single RETRAN volume per steam generator, numbered 150, 250 or 350.
Figures 1 through 3 provide nodali;e' ion diagrams of the three-loop, multi-node steam generator secondary configuration. Control I volume, junction and heat conductor region numbers starting with an X refer to three-loop geometry regions where X can have the value of 1, 2 or 3. (For the one-loop configuration X has a value equal to 1.) Control volume region numbers are underlined whereas junction and heat conductor region numbers are not. Junctions are denoted by arrows pointed in the direction associated with positive flow. The region number for an unlabeled junction is equal to the region number of the downstream control volume for the junction. The modeling of the core and core bypass volumes and reactor trip reactivity insertion assumed a fuel loading of 100% Westinghouse Vantage SH fuel to more accurately reflect the present reload fuel i designs employed at the North Anna Power Station. Unlike the earlier North Anna one-loop geometry, the reactor vessel region above the core is more realistically divided into two volumes,'an upper plenum region and an upper head region. ; 1 The earlier North Anna NSSS RETRAli model included a separate steam - ) generator inlet volume. In the Version 1 model, the hot leg volume l now runs from the' reactor vessel outlet nozzle to the top'of the steam generator hot side tubesheet. The RCP' (reactor coolant pump) J suction leg runs from the top of the steam generator cold side j l 7 ! I l
)
i l
. . - -.- - = . - . .. ,
t i 1 tubesheet to the-RCP intake. The steam generator portions of the hot leg and RCP suction volumes reflect the dimensions of the North
' Anna replacement steam generators.
{
; Version 1 includes modeling of the RPS and ESF systems, pressurizer level instrumentation, steam generator level. control, main ! .feedwater and auxiliary feedwater systems, the turbine EHC. system -4 and automatic turbine runback, and high-head safety injection.- [
n The following reactivity components are modeled: j i
- a. Doppler feedback *
- b. Moderator feedback
- c. Control rod withdrawal ;
- d. Automatic rod control ;
- e. Reactor trip i In addition the model is designed to allow.for changes in soluble a boron reactivity to be incorporated when required for a particular i transient analysis. ;
-]
The Doppler reactivity feedback is calculated by a Virginia Power
- derived correlation of Doppler reactivity. as . a . function. of core -
average fuel temperature and core burnup. For atreanalysis of a '! FSAR' transient, the Doppler feedback algorithm is~ capable of being ; adjusted to a target Doppler temperature coefficient' or Doppler. l power defect by the application of a suitable weighting factor. '! Moderator reactivity feedback can be computed either- using ' a .! moderator temperature coefficient', or a reactivity _ function' based 'l on moderator density for a transient involving significant core: :; voiding. The decay heat.is modeled with sufficient conservatism to' ensure' bounding the decay heat predicted by the 1979. ANS . Decay ; Heat, 5 Standard with a two standard deviation uncertainty applied to the; . latter. l Tables 1 and 2 provide' summaries of the control volume and momentum
. junction nodalization specified in the Version 1 model. The tables i reflect a .three-loop, multi-node steam generator ' secondary- l geometry. Abbreviations are "SG" for steam generator, "MS" -for ;
l 8
.i
main steam, and "PORV" for pressure operated relief valve. All control volumes are standard HEM (homogeneous equilibrium mixture) l volumes except volume 17 which is the nonequilibrium pressurizer. ' All junctions use .the Baroczy two-phase multiplier with Fanning wall friction and have specified single-stream compressible flow except junction 21 (surge line to cold leg) . Except where mandated i by the differences in nodalization, the control volume and momentum junction options specified in the Version 1 model are identical to those of the earlier model. I I 9
?
9
]
3
p Table 1 CONTROL VOLUME
SUMMARY
Bubble Tmp Trp Volume Descriotion Volume # Index Delav Rx vessel upper plenum 1 0 No , Rx vessel upper head 10 x No Rx vessel downcomer 11 0 No-Rx vessel lower plenum 12 0 No Core bypass 13 0 Yes > Lower core section 14 0 No Mid core section 15 0 No Upper core section 16 0 No Hot leg piping X01 O Yes , Pump suction piping X13 0 Yes Reactor coolant pump X14 0 No Cold leg piping X15 0 Yes Pressurizer 17 x No I Surge line 18 0 No SG tubes XO3-X12 0 No '! SG downcomer
- X39 0 Yes ;
SG tube bundles
- X40-X48 0 No SG separator
- X49 0 No SG steam dome 'X50 x No l
Steam lines X60, X61 0 No Main steam header 400 0 No , i Notes ! Bubble index = 0 indicates volume is treated as homogeneous, l
= x indicates a bubble index applied.
Tmp Trp Delay = temperature transport delay option. i
- Present only in' ' multi-node steam generator: secondary-geometry configurations. ;
10
Table 2 MOMENTUM JUNCTION
SUMMARY
Junction Description Jct # Tvoe Valve Chok Ira Upper head - upper plenum 10 Norn 0 No No Downcomer - lower plenum 11 Norm 0 No No Lower plenum - bypass 12 Norm 0 No No ; Lower plenum - core 1 13 Norm 0 No Yes Core 1 - core 2 14 Norm 0 No Yes Core 2 - core 3 15 Norm 0 No Yes Core 3 - upper plenum .16 Norm 0 No Yes Bypass - upper plenum 17 Norm 0 No No RX vessel outlet nozzle X01 Norn 0 No No Hot leg - SG tubes XO3 Norm 0 No Yes SG - RCP suction X13 Norm 0 No Yes Rx vessel inlet nozzle X16' Norm 0 No No Pump suction X14 Norm 0 No No Pump discharge X15 Norm 0 No No Pressurizer - surge line 20 Norm 0 No No Surge line "C" cold leg 21 Norm 0 No No Pressurizer spray intake 18 Fill 0 No No Pressurizer spray 19 Fill O No No Pressurizer PORV No. 1 24 Norm x Yes No . Pressurizer PORV No. 2 25 Norm' x Yes No Pressurizer safety valves 27 Norm x Yes No SG tubes XO4-X12 Norm 0 No Yes 11
Table 2 (cont.) i Junction Description Jct # Tvoe Valve Chok Trp SG dome - downcomer X39 Norm 0 No No SG downcomer - hot riser X40 Norm 0 No Yes SG hot side tube bundles X41-X44 Norm 0 No Yes SG downcomer - cold riser X45 Norm 0 No Yes > SG cold side tube bundles X46-X49 Norm 0 No Yes SG tube bundle - separator X50 Norm 0 No Yes SG separator - dome X51 Norm 0 No No SG outlet X60 Norm 0 No No MS isolation valves X61 Norm x No No MS header inlet X62 Norm 0 No No , i Condenser steam dump 401 Norm x Yes No MS PORV (relief valve) X26 Norm x Yes No SG safety valves X27-X31 Norm x Yes No Feedwater inlet X35 Fill 0 No No Turbine steam flow 400 Fill 0 No No Notes Type = junction type, i.e., normal or fill. Valve = 0 indicates no valve model is specified. Chok = choking option. Trp = enthalpy transport option. l l 1 l 12 1 1
\
i
3.0 MODEL QUALIFICATION P Qualification of the Version 1 model was performed by three types of calculations:
- 1. Comparison to earlier RETRAN model predictions for selected FSAR transients.
- 2. Comparison with plant data for selected station events.
- 3. Miscellaneous calculations to test model functions not tested in the previous two types of comparisons.
The objectives of the qualification study were: 1) to test the new model to ensure that it met its design objectives and was correctly constructed, and 2) to assess the impact on North Anna safety analyses using the new model by comparison to predictions obtained with the earlier version of the RETRAN model. The RETRAN code models themselves (i.e. the correlations, numerical solution schemes, etc.) have been licensed generically by the. ' utilities. A USNRC Safety Evaluation Report (SER) was issued on Version 2 of the Code (RETRAN02)6,7 Individual plant capabilities and models for using RETRAN are typically documented to the NRC via utility topical reports. Virginia Power's RETRAN topical. report 1 was submitted to the NRC in 1981 and a SER was received in 19852, Reference 2 emphasized that the NRC viewed the primary objecthre of the report was-to demonstrate Virginia Power's general capability for performing non-LOCA accident analyses: ,
"The VEPCO topical report VEP-FRD-41, " Reactor System Transient' Analysis Using the RETRAN Computer Code" was '
submitted to demonstrate the capability which VEPCO has developed for performing transient analysis using the RETRAN l 01/ MOD 03 computer code."
"The staff has reviewed the... VEPCO model descriptions and finds them acceptable for demonstrating understanding of the !
RETRAN code."
" Based on the VEPCO RETRAN model and the qualification comparisons ..., the staff concludes that _VEPCO has 13 ,
1 l b
demonstrated their capability to analyze non-LOCA initiated transients and accidents using the RETRAN computer code."2 The NRC further acknowledged in the SER that future updates to the models would be done by Virginia Power as part of a maintenance program:
"The staff requires that all future modifications of VEPCO RETRAN model and the error reporting and change control models should be placed under full quality assurance procedures."2 The updated North Anna RETRAN model has been developed and qualified in accordance with these requirements. The model consists of several thousand individual input parameters which are descriptive of the plant. The basis _ for each parameter is >
documented in a detailed model input notebook, including reference sources, auxiliary calculations performed to develop the input parameters, and cautions and limitations associated with the application of each aspect of the modele The notebook is subjected to the same peer review and quality assurance requirements as a Virginia Power Engineering Calculation, thus producing a comprehensivo data base _ and reference volume which will be available to all users of the model. The Version 1 model has been qualified via a process which is equivalent, in scope and rigor, to the qualification of the earlier ' RETRAN models documented in Reference 1. This section summarizes some key results from the Version 1 model qua1ification. I l 4 l 14 I l i
3.1 FSAR TRANSIENT COMPARISONS FSAR transient comparisons were performed to evaluate the impact of the modeling differences between the earlier and updated RETRAN model on acceptance criteria or other important parameters for selected Chapter 15 FSAR transients. To the extent feasible, the earlier and updated model transient qualification cases were initialized to have identical transient initial conditions values, (I.C.s) disabled and analysis assumptions (e.g., setpoint functions). These qualification analyses were not designed to be reevaluations of the margin to safety for the selected transients as the assumptions and accident conditions used do not necessarily reflect the current licensing analysis basis. The basis for selection of the qualification transients was: 1) transients determined to have been limiting from prior analyses, 2) transients in each of the major categories of initiating events such as reactivity change, variations in primary coolant flow rate, and changes in primary to secondary heat transfer rates, 3) both symmetric and asymmetric transients with respect to the response of the RCS loops, and 4) transients spanning the full range of initial conditions (i.e., nominal full power, deterministic full power, part power, and hot zero power). Chapter 15 FSAR transients used for the qualification were:
- a. Rod Withdrawal at Power
- b. Rod Withdrawal from Subcritical
- c. Flow Coastdown 4
- d. Loss of External Electrical Load
- e. Loss of Normal Feedwater
- f. Locked Reactor Coolant Pump Rotor Unless otherwise noted, all Version 1 model results were calculated using a single node steam generator secondary geometry. All earlier version model results also employed a single node steam l generator secondary geometry since 1) the multi-node option is not available in the earlier model and 2) the present Virginia Power analysis basis for FSAR transients performed with the RETRAN model assumes a single node steam generator secondary geometry. :
i s l 15 l l J
i 3.1.1 Rod Withdrawal at Power Comparisons for the rod withdrawal at power transient were made at nominal initial conditions for various r.i.r's (rod insertion rates) and initial power conditions and included both minimum and maximum reactivity feedback cases. This transient is typically evaluated for the ANS Condition II DNB (departure from nucleate ; boiling) acceptance criterion and to ensure that the pressurizer does not go solid. A one-loop geometry is adequate due to the symmetry of the thermal-hydraulics response of the RCS loops. The qualification cases took no credit for:
- a. Reactor trip on positive or negative neutron flux rate ,
- b. Condenser steam dump
- c. Automatic turbine runback
- d. Automatic rod control system ,
Figures 4 through 10 provide a comparison for a relatively slow r.i.r of 0.1 pcm/sec with minimum reactivity feedback and initial full power conditions. A positive MTC (moderator temperature coef ficient) of +6 pcm/ *F was used. Both models produced a reactor trip on OTDT (overtemperature delta-T) with the earlier model case tripping sooner. Figure 10 demonstrates a difference in the calculation of Doppler feedback between the two models. The i earlier model calculates Doppler feedback as a function of the core heat flux whereas the Version 1 model uses a more physically ! realistic fuel temperature correlation. Figures 11 through 17 provide the same comparisons but this time the Version 1 Doppler reactivity correlation was used in the earlier model to calculate Doppler feedback. The steam lines were removed from the Version 1 model for this calculation, but this change had negligible impact on the predictions. Comparison of the two sets of figures demonstrates that the difference in the Doppler i feedback modeling is significant for a slow reactivity insertion ! rod withdrawal event in the timing of the reactor trip. However, ' the overall difference in the predicted RCS conditions at the time of trip is negligible. This is as expected, since the reactor protection system generates a trip signal at essentially the same conditions in either case. i 16
Figures 18 through 24 ree M e/A e full power muimum reactivity feedback transient wf E a faster r.i.r (3G pcm/sec). Both models again trip on '4M. The faster r.i.c reduces the moderator and Doppler icedback diffarences between the models. The predictions match fairly well except for the pressurizer pressure (Figure 18) which demonstrates the impact of differences in the modeling of the pressurizer PORVs. The most significant difference is in the modeling of the opening and closing characteristic of the PORVs. The earlier modM PORVs open very quickly and effectively control the pressure at the lift setpoint. The updated model has delays for Loth air charging and opening time. This leads to the overshoot and undershoot in the response. The updated modeling is based on surveillance test results from the plant. The dif ferences are not particularly significant to the rod withdrawal at power transient. The final rod withdrawal at power comparison is for a fast r.i.r (40 pcm/sec) with maximum reactivity feedback at a reduced power (60% nominal). Figures 25 through 31 show the predictions to be reasonably close except that the slightly faster increase in vessel Tavg for the earlier model leads to an earlier reactor trip (both models trip on OTDT) and the pressurizer pressure shows more fluctuation with the updated model due to the differences in the modeling of the PORVs discussed above. Overall the comparisons appear acceptably close. Most of the deviation is due to differences in the modeling of the Doppler reactivity and the pressurizer PORVs. In that the Version 1 Doppler feedback correlation is more physically realistic and well qualified, its use is preferred. Likewise, the updated model's representations of the pressurizer PORVs and safety valves are also more realistic than that of the earlier model. 3.1.2 Rod Withdrawal from Subcritical Comparisons for the rod withdrawal from subcritical transient are provided in Figures 32 through 36. The analysis was performed at hot zero power (HZP) initial conditions for a r.i.r of 100 pcm/sec. The transient is typically evaluated for the Condition II DNB criterion by showing that the maximum fuel average temperature, maximum clad average temperature and maximum core average heat flux 17
are less than the steady state full power values. A one-loop RCS geometry is adequate due to the symmetry of the thermal-hydraulics response of the RCS loops. The qualification cases took no credit for:
- a. Reactor trip on positive or negative neutron flux rate ,
- b. Condenser steam dump i
- c. Automatic turbine runback
- d. Automatic rod control system The earlier model calculated Doppler feedback using a Doppler defect as a function of fuel temperature. The Version 1 model applied a neighting factor to provide an equivalent Doppler defect over a HZP to nominal full power core average fuel temperature range.
Both cases had a reactor trip on the low and intermediate range high neutron flux signal at 8.8 sec. Inspection of Figurtas 32 through 36 shows excellent agreement between the two models. 3.1.3 Flow Coastdown The complete loss of flow event is a potentially limiting core DNB transient. The qualification analysis assumed nominal initial conditions and was performed with the one-loop RCS geometry. Two complete loss of flow cases are typically evaluated: a decrease in bus voltage and a decrease in bus frequency.- Since these cases are similar except for the trip time, the decrease in bus voltage case was analyzed. The Version 1 model RCP (reactor coolant pump) input was adjusted to provide predictions which conservatively bound the isothermal , flow coastdown behavior indicated by plant test data. - This changed the shape of the coastdown somewhat from that used in the earlier-model. The moderator reactivity was set to O pcm/*F for the qualification - to eliminate the positive' interaction which exaggerates 18
differences between the two model's Doppler feedback calculation that was observed in the rod withdrawal at power cases (Section 3.1.1). The pressurizer PORV modeling in the Version 1 case was modified to that of the earlier model, i.e., to open quickly and ride the setpoint. This modification will not impact the DNB acceptance criteria since the change to the updated model's PORVs impacts the pressure only after the time of minimum DNBR. Figures 37 through 45 provide comparison plots for the transient. Figure 37 plots the normalized core flow rate which is one of the key inputs to the DNB analysis. The updated model predicted flow rate is slightly lower than that of the earlier model for the first two seconds. Thereafter, the updated model predicts a larger flow. These differences are the result of the adjustments made to bound the North Anna plant data that were described above. The predicted normalized core flow rate and normalized heat flux are key inputs to the detailed core thermal hydraulics analysis. The slight increase in the normalized core power of the earlier model during the pre-trip period (Figure 38) is due to the positive Doppler feedback that results from the decrease in heat flux as the flow begins to coastdown. The updated model, with the fuel temperature based Doppler feedback correlation, shows no such behavior. The normalized heat flux presented in Figure 39 shows essentially no deviation between the two models. The steam and feedwater flow rates (Figures 40 and 41) are ; initially nearly identical. At later times the operation of the secondary PORV's causes a noticeable difference. The secondary PORV flow is not included in the earlier feedwater model. The steam generator pressure (Figure 42) agrees very well and the dif ferences are consistent with the earlier model not including the - steam lines. The remaining parameters (Figures 43 through 45) do not suggest any inappropriate differences between the models. Overall, the differences in the results between the two cases are small and consistent with the modeling differences. 3.1.4 Loss of External Electrical Load Comparisons for the loss of load transient were made at beginning-of-cycle, nominal full power initial conditions. The transient is 19
typically evaluated for both DNB and overpressure acceptance , criteria at both beginning-of-cycle and end-of-cycle conditions. The transient is limiting for the overpressure criterion but not the DNB criterion, so a case employing the overpreEsure analysis basis assumptions was used for the qualification. A one-loop : geometry is adequate due to the symmetry of the thermal-hydraulics response of the RCS loops. The qualification analysis took no credit for:
- a. Direct reactor trip on turbine trip
- b. Steam dump system (condenser and atmospheric steam relief valves disabled)
- c. Pressurizer sprays
- d. Pressurizer PORVs [
i
- e. Automatic rod control system Figures 46 and 47 compare the pressurizer pressure and steam generator secondary pressure for the two models. The slight deviation in secondary pressure of Figure 47 is due to the presence of the steam lines in the Version 1 model. This is confirmed by Figure 48 where the steam lines have been effectively eliminated in the updated model case.
The loss of load cases described above used a single node steam generator secondary geometry. Figures 49 through 54 show the , impact of a multi-node steam generator secondary geometry (both , cases employ the Version 1 model). The deviation between predictions is mostly due to the steam generator secondary parameters which vary as would be expected due to the multi-node steam generator configuration's more accurate modeling of the liquid shrinkage in the steam generator secondary (Figure-52). 3.1.5 Loss of Normal Feedwater Comparisons for the loss of normal - feedwater transient were performed with a three-loop, single node steam generator secondary geometry. In the past, this transient has been used as the limiting event for establishing minimum required AFW (auxiliary feedwater) pump flow. The three-loop geometry is used to model initially unbalanced steam generator levels. 20 I j i e
The qualification analysis took no credit for:
- a. Reactor trip on low-low steam generator level in steam generators "B" and "C"
- b. Reactor trip on high pressurizer pressure
- c. Reactor trip on high pressurizer level ,
- d. Reactor trip on OTDT or OPDT
- e. Reactor trip on low steam generator water level coincident with steam flow /FW flow mismatch '
- f. Reactor trip on low loop coolant flow
- g. Pressurizer heaters
- h. Pressurizer sprays
- i. Pressurizer PORVs
- j. Steam dump system (condenser and atmospheric steam relief valves are disabled)
- k. Automatic rod control system
- 1. Safety injection on low pressurizer pressure The steam generator with the highest initial level (steam generator loop 1) is aligned to the turbine-driven AFW pump which is "A",
assumed to fail. The event is initiated by turning off main feedwater to all steam generators at ten seconds. The first ten seconds of steady state operation provide a check on proper code initialization. AFW is delivered to steam generators "B" and "C" by the two motor-driven AFW pumps with reduced volumetric flow rate of 200 gpm each. These pumps start 60 seconds after the reactor trip. Off site power is maintained so the reactor coolant pumps remain on for the entire transient. The moderator and Doppler reactivity effects were eliminated. Feedback alters the rate that. energy is'added to the system after the main feedwater pumps have tripped. Qualification of the updated model's reactivity algorithms is provided primarily by the rod withdrawal from suberitical and rod withdrawal at power , comparisons described in Sections 3.1.1 and 3.1.2 above. Both models had a reactor trip on low-low steam generator level, the earlier model tripping at 31.1 sec and the updated model at 21
e 30.8 sec. At 36 see the main steam safety valve initial setpoint (1100 paia) was reached for both cases. AFW initiation occurred at 91.1 sec for the earlier model and 90.8 sec for the updated model. The predictions are provided in Figures 55 through 64. Pressurizer average pressure, pressurizer liquid volume and reactor coolant temperature all behave similarly. These three parameters start off at steady state and begin to increase after the main feedwater pumps have been turned off. Upon reactor trip, the parameters decrease rapidly until the steam generator levels drop to a point where the secondary loop is no longer able to remove the heat generated on the primary side. Two successive peaks follow as the heat generation rate and the heat removal rate switch dominance. Thus, a transient with three separate, distinct peaks is observed. (See Figures 55 through 60.) , The magnitude of the first peak is related to the amount of energy put into the primary system between the main feedwater trip and the reactor trip. After the reactor trip, the pressurizer average-pressure, pressurizer liquid volume and RCS temperature drop. The severity of this drop is less in the updated model's pressurizer. This difference is due to the coolant density change in the RCS volume. When a coolant density change occurs in the RCS, the volume change may be seen in the pressurizer. The models drop to equivalent temperatures, but the pressurizer level drop is not equal due to the differences in the RCS' volumes and nodalization between the two models, especially in the upper portion of the reactor vessel where the Version 1 model has separate volumes for , the upper plenum and upper head while the earlier model has a ; single volume. Thus the effective active fluid volume of the reactor vessel which communicates with the rest of the RCS is smaller in the Version 1 model since the upper head is stagnant. . The long term behavior of the loss of normal feedwater transient greatly depends on the secondary side's ability to remove heat. Comparison of the secondary side steam generator liquid volume and pressure is provided in Figures 61 through 64. Figure 61 shows- , that steam generator "A" drys out, changing its primary-to-secondary mode of heat transfer (identically for both models). The secondary side results compare favorably well except for the noticeable dip in the loop "B" steam genvrator liquid volume at
-1 22 I l
i
about 2000 sec for the Version 1 model (Figure 62) . This is due to a change in the heat transfer mode of steam generator "B" (and "C") from the nucleate boiling Thom correlation to a forced convection r vaporization Schrock-Grossman correlation at 1900 sec. At 2500 sec ; the heat transfer mode returns to nucleate boiling and remains [ there for the remainder of the transient. This change in the primary-to-secondary heat transfer mode 'is very sensitive to mass. . Additional calculations have confirmed that a slightly greater : secondary mass in the updated model case would have prevented the I heat transfer mode change. , Further investigation showed that differences in the modeling of the main steam safety valves was the most significant contributor to the difference in model predictions. The earlier model groups { the five safety valves per steam line into a single valve function whereas in Version 1 each valve is modeled individually. The earlier model allows more steam relief through the safeties so that the steam generator level drops more rapidly. The results provide acceptable qualification of the updated model , for modeling the loss of normal feedwater transient and related [ thermal-hydraulic behavior. The comparison differences are due primarily to the modeling changes which make the updated model more realistic. 3.1.6 Locked Reactor Coolant Pump Rotor The locked reactor coolant pump rotor event is evaluated for primary and secondary overpressure and core DNB acceptance criteria. The qualification analysis was performed with assumptions based on an evaluation of the overpressure acceptance criteria. This is adequate for model qualification since it is expected that the case analyzed will provide similar trends in terms of the DNB sensitive parameters of core heat flux, core inlet temperature, core mass flow rate and RCS pressure. ., Since only one of the RCP rotors is locked, the asymmetric nature > of the transient requires a two-loop or three-loop RCS geometry. .The earlier model analysis was performed with two-loop geometry. The earlier model reflected the original steam generator design whereas Version 1 models the replacement design. 23 l
.t
1 The analysis took no credit for:
- a. Steam dump system (condenser and atmospheric steam relief ;
valves are disabled)
- b. Automatic rod control system
- c. Pressurizer heaters
- d. Pressurizer sprays and PORVs
- e. Reactor trip on high range, high level flux ;
i
- f. Reactor trip on negative flux rate
- g. Reactor trip on high pressurizer pressure
- h. Reactor trip on high pressurizer level *
- i. Reactor trip on OTDT The earlier model calculated Doppler reactivity from a table of '
Doppler defect as a function of core heat flux. Hence, the Doppler reactivity. predicted by the earlier model will vary with the rate of core heat flux change. Both models had a reactor trip on low RCS flow, the earlier model trip being at 1.04 sec and the updated model at 1.10 sec. Figures 66 and 67 show Version 1 to predict greater reverse flow through ' the loop with the locked rotor. Figures 66 and 67 show a divergence in - the RCS flow predictions , between the two models. This divergence was found to be due mostly to differences in the models' reactor vessel and RCS loop loss coefficients and inertias. Those for. the Version 1 model were ! based on detailed predictions of vendor design codes not available
)
when the earlier model was constructed. Version 1 flow predictions compared well with North Anna plant data for two reactor coolant pumps running with reverse flow through the third' loop. , Figure 69 shows the Doppler reactivity decreasing more rapidly in the earlier model. As noted in Section 3.1.1 above, this is caused i by the earlier model's functional dependence.upon core heat flux. As before, this causes a noticeable divergence only after reactor trip. Overall, the two models show excellent' agreement, with the new model being slightly less conservative for the overpressure case, and more conservative with respect to-flow coastdown. l 24 ] i l l
3.2 COMPARISON TO PLANT DATA - STEAM GENERATOR TUBE RUPTURE EVENT A steam generator tube rupture occurred on July 1987 in the C steam generator at North Anna Unit 1. An analysis of the event was completed immediately thereafter with the earlier RETRAN model. The qualification analysis performed with the Version 1 model used the boundary conditions developed for this previous work. In other words, the purpose of the effort was not to independently study the event, but to evaluate the behavior of the updated model. The analysis was performed with a three-loop, multi-node steam generator secondary configuration. Results compared to the plant data are illustrated in Figures 71 through 76. The overall magnitude and character of the predicted response agree very well with the data. The general agreement with the data is as good as or better than the previous work with differences between data and , prediction occurring at similar transient times. The pressurizer pressure (Figure 71) agrees very well with a couple of explainable small variations. When the pressurizer finally empties at approximately 1250 seconds (Figure 72), the pressure data show a more substantial drop in pressure than the prediction, suggesting that the head probably flashed too much mass in the model. This could be a result of the initial upper head temperature. The manner in which the head would be modeled for any particular licensing analysis will be dictated by transient specific considerations, therefore this best estimate difference is of no significance. The timing of the start of pressurizer refill and the rate of refill look quite good. As the pressurizer level reaches the 80% span, the model predicts a sharp drop in the level and a corresponding drop in pressurizer pressure. This is a result of the safety injection fluid cooling and depressurizing the reactor vessel upper head, and thus sucking the fluid back out of the pressurizer. In the model this behavior is exaggerated by the equilibrium upper head model. , Figures 73 and 74 illustrate the steam generator secondary pressure [ for the A and B loops and C loop respectively. The A and B pressure response is generally good with the most notable difference from about 600 to 1200 seconds. This is a period when, according to the plant, the condenser dumps were closed. The gradual decay in steam pressure may have been due to other leakage 25
pathways not represented in the model. At 1180 seconds, the conderner ' dumps were again enabled and the rate of pressure decrease is fairly consistent with the data. The C loop i differences are similar for the period from 600 seconds, when the ; condenser dumps became inactive, to 960 seconds when the main steam I isolation valves were closed. Complete closure of the condenser dumps valves during this period appears unlikely.- Figures 75 and 76 provide predictions of the cold leg temperatures. The differencas in the A and B loop predictions are consistent with the differences in the pressure response. Since the reactor l coolant pumps are running, the C loop cold leg temperature looks very much like that of the A and B loops. i The agreement between the multi-node steam generator case predictions and the data is similar to that of the previous analysis. The differences could be diminished by further study, but the agreement is certainly close enough to demonstrate the qualification of the model. .This is particularly true given the fact that no special features were added to enhance the match with the data. i t 9 f a 4 26 4
3.3 MISCELLANEOUS TEST - TURBINE RUNBACK This test examined the overall response of the condenser steam dump and control rod control systems to a transient that does not result in a reactor trip or the actuation of safety systems. The load rejection is controlled by the condenser dump and rod control systems maneuvering the plant to a steady condition at the new power level. A standard turbine runback from 100% to 60% power was examined for runback rates of 10% per minute and 200% per minute. Cases for each runback rate were run at beginning-of-cycle and end- . of-cycle conditions with a one-loop, multi-node steam generator secondary geometry. A comparison of multi-node versus single node : steam generator secondary predictions was also performed for the beginning-of-cycle case with 200%/ min runback rate and are provided in Figures 77 through 89. The predictions are consistent with expectations for the different runback rates. The perturbation begins with the turbine flow. The steam flow rate drops off sharply and the condenser dumps open to control Tavg. The control rods respond to the power and temperature error and drive the rods into the core. The remaining primary and secondary parameters respond as expected. The RCS parameters, Tavg, reactor power,- pressurizer level, and pressure, all compare very well. The steam generator pressure response shows some differences between the two models. The steam pressure is lower with the single node geometry (Figure 81) but only on the order of 10 psi. The steam generator level response shows the obvious difference (Figure 82). The single node geometry cannot predict the shrink and swell behavior during the course of ' the transient. The steam generator level controller will drag the level back to setpoint for both geometries, but the dynamics will be different. 1 27 r i
4.0 CONCLUSION
S Version 1 of the North Anna RETRAN model is an updated version of the earlier RETRAN model Virginia Power has used to perform safety analyses for its North Anna Nuclear Power Station. Version 1 includes improvements over the earlier model in that several of the system functions are more realistically modeled, additional secondary side noding is included, and the model design allows for more flexibility in its use, especially in the support of plant t operational concerns. The Version 1 model was qualified through 1) comparison of predictions of selected FSAR transients with those of the earlier RETRAN model, 2) comparison with plant data for selected station events, and 3) miscellaneous calculations to test model functions not tested in the prior two types of qualification. From the comparison to earlier model predictions for the FSAR transients, it was concluded that the most visible changes are due to the more realistic modeling.of Doppler reactivity feedback and PORV and safety valve behavior. These changes, however, do not represent changes in the methods or assumptions used in performing licensing analysis and their incorporation still ensures that adequate conservatisms will be applied to such analyses. It is concluded that Version 1 has been acceptably qualified for use in the safety analysis of the North Anna Power Station in accordance with the requirements of Reference 2 and Virginia Power quality assurance procedures. 4 28 i
F 5.0 REFERENCEB
- 1. Virginia Power Topical Report VEP-FRD-41-A, " Virginia Power Reactor System Transient Analyses Using the RETRAN Computer Code,"
May, 1985.
- 2. Letter from C. O. Thomas (NRC) to W. L. Stewart (Virginia Power) , Acceptance for Referencing of Licensing Topical Report VEP-FRD-41, " Virginia Power Reactor System Transient Analyses Using the RETRAN Computer Code," April 11, 1985.
- 3. Letter from W. L. Stewart (Virginia Power) to H. %. Denton (NRC), " Virginia Power Surry and North Anna Power Stations Reactor System Transient Analyses," Serial No. 85-753, November 19, 1985.
- 4. Virginia Power Topical Report VEP-FRD-42, Rev. 1-A, " Reload Nuclear Design Methodology," September, 1986.
- 5. ANSI /ANS-5.1-1979, "American National Standard for Decay Heat Power in Light Water Reactors," approved August 29, 19 ~, 3 .
- 6. Letter from C. O. Thomas (NRC) to T. W. Schnatz (UGRA),
Acceptance for Referencing of Licensing Topical Reports EPRI CCM-5, "RETRAN-A Program for One Dimensional Transient Thermal Hydraulic Analysis of complex Fluid Flow Systems," and EPRI NP-1850-CCM, "RETRAN-02-A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems," 2:eptember 4, 1984.
- 7. Letter from A. C. Thadani (NRC) to R. Turia (GPJ), " Acceptance for Referencing Topical Report EPRI-NP-1850 CCM-A, Revisions 2 and 3 Regarding RETRAN02/ModOO3 and ModOO4," October 19, 1988.
29 g-
V R-4 APPENDIZ - FIGURES l t 7 a e I k l 30
3 is r-
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. FIGURE 1 ;{
NAPS RETRAN MODEL PRIMARY' NODALIZATION. !
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[25 2, 27 Spray l
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]
(Loop 3) _ g XO3 10, :; RCP A -, O / t Upper - I I Zli . Plenum -> XD1 : Y Hot Leg c@e NM :xis17 1 Leg- g A 416 l
~ -
l A X17 g HHSI h g f15- !f ^ [
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li- :1! ;
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' nn Control Volume
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jet is
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-:::::::, t v::::,,_
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*- rO b
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\! ; 1.ft1 > Steam '
16q > /_ 161
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FIGURE 18; I ROD WITHDRAWAL ATEPOWER . 30 PCM/SEC -100% POWER - MAX. FEEDBACK-PRESSURIZER PRESSURE .i 8 8 Tu- _ J
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i i l a tu. g i tu
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.I
'2/10/93 19.45.51 n, -
_. }
1 i s FIGURE 26 ROD WITHDRAWAL AT POWER 40 PCM/SEC - 60% POWER - MAX. FEEDBACK PRESSURIZER L10UID VOLUME 8 b 8 o _- O / g , t mN. ' I j M ' i F- g
' \ "O /
O \ UJ '
, \
2 0. / \ 3 / O '
\ > f s O , s 38" ,
O ,
- s ,
CY. O Nf, o_ c 8 s
*0.00 2'O.00 4'O.00 6'O.00 8'O.00 1'0.00 0 120.00 TIME (SEC)
SOLID LINE = EARLIER MODEL DASHED LINE = VERSION 1 2/10/93 19.45.51
r i FIGURE 27 ROD WITHDRAWAL AT POWER 40 PCM/SEC - 60% POWER - MAX. FEEDBACK NUCLEAR POWER 8 , h, 8 6
?- ,
,- n
' s g i 86 i
/
% 1
$8 '
od O. o. i > 8 e
< i Wo I .
dR I DC- s z , , E i g- i s w,,,-----. , 8 .
%.00 2'o.co 4'o.oo - 6'o.oo e'o.oo -t'o.oo o iao.co-TIME (SEC)
SOLID LINE = EARLIER MODEL DASHED LINE = VERSION 1 ,
.2/10/93- 19.45.51
F1GURE 28 ROD WITHDRAWAL AT POWER 40 PCM/SEC - 60% POWER - MAX. FEEDBACK STEAM GENERATOR STEAM PRESSURE 8 . d ' 8. 8 6 , ls N s 2-
/
28
~
mo '
/
LJJo / To / Db / w / a < Q. o / g
<8 i 3- /
m Oo mo , 8. cn 8 8
%.00 2'0.00 4'o.00 6'O.00 8'0.00 1'0.00 0 1'20.00 -
TIME (SEC) SOLID LINE = EARLIER t10 DEL I DASHED LINE = VERSION 1 l i 2"0/93 19.45.51 ,
F1GURE 29 ROD WITHDRAWAL AT POWER 40 PCM/SEC - 60% POWER - MAX. FEEDBACK VESSEL AVERAGE TEMPERATURE . 8 C , c'
's h, t '
O I 1 O l t O / \ - D / \ [ / > 0 8
,' 1 tDef \
Og- i t
/ \
a '
\
>o <o !
\
s YN ' 8- ,' \ s
/ *s i
/ \
g /
/ \ . \
C / \ S- / N .
/
/
l 8
%.00 2'O.00 4'O.00 6'O.00 8'O.00 l'00.00 l'20.00 Ji TIME (SEC) ,
SOLID LINE = EARLIER MODEL : DASHED LINE = VERSION 1 i 2/10/93 19.45.51-
i FIGURE 30 ROD WITHDRAWAL AT POWER 40 PCM/SEC - 60% POWER - MAX. FEEDBACK MODERATOR REACTIVITY 8 > d 8 d-O /
-8 . s i s
8.. s /
- s i Eo 00
\ /
g g h \ l S o.. I ' I K o I x I 08 W, I C w 8 I > OO. o , . r8 i
?~ g i
8 '
\ I f ,
i
$ 1 . ,
' O'. 00 2'O.00 4'O.00 6'O.00 8'O.00- l'00.00- l'20.00 TIME (SEC) .
SOLID LINE = EARLIER MODEL ! DASHED LINE = VERSION 1 I I 2/10/93 19.45.51 . t
FIGURE 31 ROD WITHDRAWAL AT POWER 40 PCM/SEC - 60% POWER - MAX. FEEDBACK DOPPLER REACTIVITY 8 5 9-8
'o -8 E
og ,_
!bs_ ,'
* /
O , I I l E8 i Us- I
- o. .1 Q. s l Co 's Oo % 'I h_
~~ l 8 -
5
'O.00 2'O.00 4'O.00 6'O.00 8'O.00 1'0.00 0 l'20'.00 TIME (SEC)
SOLID LINE =" EARLIER MODEL DASHED LINE = VERSION 1 :; l 2/10/93 19.45.51 u
i FIGURE 32 ROD WITHDRAWAL FROM SUBCR-ITICAL l AVERAGE FUEL TEMPERATURE- j : 1 o
- O 2
O. _t t
.c o t O '!
o - J.- a
.L o ,,
no gg. " =;
~
e w " 98 JC. d " wN <6 g .. ; tu Ho '6
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e8.
>N
., i o -'
o d 46
. @- A, l 8 .. .
o . . . . . ._a
%.00 -4'.00 8'.00 l'2.00 ~ I'6.00 2'O.00 2'4.00 !
TIME (SEC) . SYMBOLS ~-= EARLIER MODEL
-SOLID LINE =. VERSION.1 ,
'2/10/93 19.46.31 ,
1
'b
f i FIGURE 33 . ROD WITHDRAWAL FROM SUBCRITICAL -l AVERAGE CLAD TEMPERATURE !
-l o 1 o .i
'8.
e ! a o 9 i o : g- i i o I o ;
-d j u.g. o e -
w- ; 98 " o8. 4 !
<c o "
"8 . ,
c,.58. . . o o . d g. a.- j
, a 6 6 6 6 6 - .
o k.oo L4'.oo e'.oo t'2.oo is.oo 2'o.oo 2'4.00-TIM TSEC1
.- i :
SYMBOLS =' EARLIER MODEL i i SOLID LINE.= VEPON 1 ! 2/10/93 19.46.31 1
FIGURE 34 l ROD WITHDRAWAL'FROM SUBCRITICAL .: CORE-AVERAGE HEAT FLUX l
,i 2 i i
g .;
-i o a weo -
N *-
,O -
_a . lc cc OS z. wo . 6 x . , o - A
.J u.o " -
r d- "
. 1
< 1 w ;
I a ; d b & d- * ' en '. o a
-R . . . . . . . # 1
.o5 - ~ 4'.05 - ~ e!oo- 1'2.00- f6.oo 2'o.oo - 2'4.00-j iTIME .. (SECT r ..
SYMBOLS:.:=~ EARLIER MODEL- 1 SOLID LINE = VERSION 1 .I I 2/10/93 19.46.31 i
FIGURE 35 ROD WITHDRAWAL FR011 SUBCRITICAL- -l-NUCLEAR POWER ; l O , O ! 5.
}
O 2 m o -
- 00. ,
w- ,; F-5 u_ o o oe- : Z t o e- : 00 ' 40 d e g-b '! e wo 30 - o g-n. t x WoJO C O&- D z -! 8 . . .. . . . . __ _ 1 - - - -. N.06 - 4'.06 ~ ~ 8500 l'2.00' 1'6.00 '2'O.00 ~2'4.00; ". TIME (SEC) .
-l 1
SYMBOLS = EARLIER'MODEL~ l SOLID LINE = VERSION 1 l 3
*i 3/23/93 9.46.27 l h
l FIGURE 36 ] ; ROD WITHDRAWAL'FROM SUBCRITICAL ' i
^
CORE AVERAGE MODERATOR TEMPERATURE ,! 8 _ d
$- ]
8 i 8- . i
^8 u.
c
@S-e , :
o8 , Oe i
~
E c. '6 ro c 38 a
<d c. . 6 ,
ujo E o a- .: Oo o & l d' A A, 'i f g- o 8 . _ _ _ _ _ _ _ .J 1 5
%.00 4'.00 -8'.00 1'2.00. I'6.00' 2'O.00' 2'4'.00' ,
H TIME (SECT 1 SYMBOLS = EARLIER MODEL- ~ < l SOLID:LINE = VERSION 1- .t a
-t 2/10/93 -.19.46.31 >-
, , , - -y . - .,
FIGURE 37 FLOW C0ASTDOWN NORMALIZED TOTAL RCS MASS FLOW RATE 8
- g. .
W W < i ES x f-O x J s k N se N E f- \
\
Q w \
\
Nm N -o _J < f-
's s r s N
O zn
's s c
d-
's
's s-M %
'%.00' 2'.00 4'.00 6'.00 8'.00 1'O.00 1'2.00 TIME (SEC)
SOLID LINE = EARLIER MODEL DASHED LINE = VERSION 1 l 2/10/93- 20,00.14 i
FIGURE 38 FLOW COASTDOWN NORMALIZED NUCLEAR POWER 2 8 -
\
N
\
O \ mk \ yo- r x w 50 O_ o* e w Dd-z 8 g-8 k.oo 2'. 00 - 4'.00 6'.00 8'.00 t'o.oo ia.oo-TIME (SEC) , SOLID LINE ' = EARLIER t10 DEL DASHED LINE = VERSION 1 2/10/93 20.00.14
FIGURE 39 FLOW COASTDOWN e CORE AVERAGE HEAT FLUX , d 8 J-X o .J '8 q:- w I wo tco of- . o O w 28 .J < f-N r Cr o Zo*
- o. ,
8 k.00 2'.00 4'.00 6'. 00 . 8'.00 1'O.00 l'2.00 TI!1E (SEC) SOLID LINE = EARLIER MODEL DASHED LINE = VERSION 1 - I 2/10/93- 20.00,14
F b FIGURE 40 FLOW COASTDOWN ; TOTAL STEAM FLOW RATE 8 t d U-8 . d b l 8 id I I ko 19 i t O. . we i O. l >
'8 26.
oc _J LL mo 09
< o.
rv , E I
< \ '
Lao Ho l Dd. 1 i ; I O I R .,
%.00 2.00 4'.00 6'.00 0'.00 1'O.00 1'2.00 TIME (SEC)
SOLID LINE = EARLIER t10 DEL
-i DASHED LINE = VERSION 1 i
~}
2/10/93 20.00.14
FIGURE 41-1 FLOW COASTDOWN TOTAL FEEDWATER FLOW RATE 8 d b" 8 . 88 a I w H I < s "o I a.9 t 08- , 5 I ro 09 Jo. u.c m 4 m , <o r9 i g @" l y I , & \ < I 3o OO I W d. I w ru l.i I i ___ . i / . ,- g ,? l
\
Cb.oo 2.00 4'.00 6'.00 e'.00 1'o.00 1'2.00 TIME -(SEC) : SOLID LINE = EARLIER MODEL DASHED LINE = VERSION 1 i l l
'1 1
.2/10/93 20.00,14 i
FIGURE 42 FLOW COASTDOWN STEAM GENERATOR STEAM PRESSURE 8 b 9 8 6 ' 9 -
~
s'
/
20 f -f Uhn / %9- ,'
/
Wo 3 9 / m - $g9- e CC ' a / EO /
/
<f Wo. / W@ / , O /
/
/
O / O. p j
/
8 5
%.00 2'.00 4'.00 6'.00 8'.00 l'O.00 1'2,00 TIME (SEC) .
SOLID LINE = EARLIER MODEL DASHED LINE = VERSION 1 1 2/10/93- 20.00.14 1
FIGURE 43 FLOW-COASTDOWN VESSEL AVERAGE TEMPERATURE 8 4 ' 8-8 8- ! 8 -
- s s
O /
$- / \
s
/ \
o / \ 00 ws #
/ \
90' f s ,
/ g.
O '
/ g go . _ __ g
-s N
$" \
\
\
\
O N O J \
\
C.
@ \
\
\
8 5
%.00 2'.00 4'.00 . 6'.00 . 8'.00 1'O.00- 1'2.00-TIME (SEC) :
SOLID LINE = EARLIER MODEL DASHED LINE = VERSION.1 ( 2/10/93 20.00.14
1 FIGURE 44 FLOW COASTDOWN ; PRESSURIZER LIOUID VOLUME - 8 . d a- . G g , ~~,
. / N 8.
@ ' s
, ~ ,
, s i f
\
"O \
M r O. \- , bb. O
\
\
w \ r \ 08
._1
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s . g O@-
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\
Ch. o s
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ce \ N g + a_g g
- d. \
\
\
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"b.oo 2'.00 4'.00 6'.00 . 8'.00 1'o.co ia.co. i TIME (SEC)
SOLID LINE = EARLIER MODEL DASHED LINE = VERSION 1 i 2/10/93 20.00.14.- a
4 FIGURE 45 FLOW COASTDOWN PRESSURIZER PRESSURE . 8 5
?.
m 8 6
~.
o / d ' i
. N
-m /
\
@ /
\ [
/
wo /
\
ee ' -) U)y-
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U) \ wo s % \ Q. o \ e ey-N s , o_ s 8 's ' 6 s N 8 5 k.00 2'.00 4'.00 6'.00 8'.00 ib.00 i2.00. TIME (SEC) , 1 SOLID LINE = EARLIER r10 DEL l DASHED LINE = VERSION 1 j 1 2/10/93 20.00.I4
1 FIGURE 46 LOSS OF LOAD PRESSURIZER PRESSURE t 8 5 b" 8 d
~
o ' 78 d h" / m I t
< 1 I -o #
(f)O ' 1 CLd \
" v ., /
N \ w . t x \ Do t , (Do \ ' (Of WN. I xN \ Q. Eo No s s afo. 's, ' N ,'
~~___ ,
8 d
~0.00 10.00. 20.00 30.00- 40.00 50.00 60.00 TIME (SEC)
SOLID LINE = EARLIER MODEL , DASHED LINE = VERSION 1 2/10/93 20.01.43
r FIGURE 47 ; LOSS OF LOAD ; SG SECONDARY STEAM PRESSURE - i 8 d . 8 '! 8 fu. no i g ,
.m8 !
O. ~ . I w-j w0
& I a- i m? , .
' U)o. -
w- I
& l' +
- a. i r8 J
wS. l wn m :! 00
'mo ,
'O i 8- l 8 ;
.00- io.oo. :a'o;oo... 3'o.oo 4'o.00 5'o'. 00 - 6'o.004
- TIME (SEC) . : i SOLID LINE-.= EARLIERLMODEL !
' DASHED'LINE =' VERSION.1.
'2/10/93 20.01.43 ..
Li I FIGURE 48 ! LOSS OF LOAD : SG SECONDARY STEAM PRESSURE " 8 4 8
- m. t 8 >
8 2.'- o CON . g- .I
~!
o; Wo r x3 o. .. co? ' (0 0. W- : e ~; Q. i r8 ! W8. rm (f) .; o8 en
~8.
e J
?
8 - 3
.00 t o' .co ~ 2'o.00- 3'o.oo ~ 4'o.oo s'o .co 6'o;oo TIME (SECT SOLID;LINE = EARLIER MODEL ;
' DASHED LIN_E =-VERSION 1 - NO: STEAM LINES . $
i j 1]
-2/10/93- 20.01.45'
.]
d
FIGURE 49 LOSS OF LOAD PRESSURIZER PRESSURE 8-8. n 8 5 78 o
\ '
E8, G- o \ k- \ w \ x \ S8 ' s m5 ..
$$- \ '
a. To No
- a. -
8-o T .- o.co t'o.co 2'o.co 3'o oo 4'o.co s'o.co 60.00 TIME (SEC) ' SOLID LINE = VERSION 1 SINGLE NODE SG' DASHED LINE = VERSION 1 MULTI-NODE SG. l l l
'2/10/93 .20.02.32
FIGURE 50 LOSS OF LOAD NEW MODEL SG GEOMETRIES I 8 5 6. 8 i o s
- g. s
\
\
mo NO
\
g n- \ u.8-w e \
\
W \ 58 \ ds s > 8- \ 3
\
08 o.
\
O \
.Jk- \
\
E \ N s 18 N d s 3- 's
's % s 8 ;
5 "b.00 l'O.00 2'O.00 3'O.00 4'O.00 5'O.00 6'O.00 ; TIME (SEC) SOLID LINE = VERSION 1-SINGLE NODE SG DASHED LINE = VERSION 1 MULTI-NODE'SG , 2/10/93 20.02.32
t FIGURE 51 LOSS OF LOAD ; SG SECONDARY STEAM PRESSURE 8 6 8. 8 5 8.
' ~- ~~
l a ~8 t.0 I > $~ I l WO # MO I O I m, t.O O. I L1) ~ l & I + 1 I'8 i <d we. , wm to e 00 Wo
'8.
e r 8
.00- t'o.oo 2'o.co 3'o.oo 4'o.co s'o.co 6'o.oo ,
TIME (SEC) SOLID.LINE = VERSION l' SINGLE NODE SG , DASHED LINE = VERSION 1 MULTI-NODE SG i 2/10/93 -20.02.32
F1GURE 52 LOSS OF LOAD SG NARROW RANGE LEVEL t 8 d- , 8 i
- d. I T l I
i 8 ' m - t e i i _J o I wo t
- \
_J t I Zo o y> C-g
\
l i 8 i d- i - t i t 8 s---------------. --. k.00 l'O.00 2'O.00 3'O.00- IO.00 5'O.00 6'O.00' TIME (SEC) :
~
SOLID LINE = VERSION-1 SINGLE NODE SG ; DASHED LINE = VERSION 1-MULTI-NODE SG l i 2/10/93 20.02.32 ,
FIGURE 53 LOSS OF LOAD ; SG TOTAL SECONDARY MASS 8 i R-8 d
- o. s O
-8
\
m \ E \ coo d \ S. A s u)(u U) \
< s Eo s O g JN \
NS' s O s 08 's ma ',
$- 's
's s,
8 s
%.00 1'O.00 2'O.00 3'O.00 4'O.00 5'O.00 6'O.00 TI(1E (SEC)
SOLID LINE = VERSION 1 SINGLE NODE SG DASHED LINE = VERSION 1 MULTI-NODE SG 2/10/93 20.02.32
.2
. i FIGURE 54 !
LOSS OF LOAD , VESSEL AVERAGE TEMPERATURE 4 8 l g -i 3- !
.i!
8 1 5 8- .i i o : 9
$- s l m t j-Lt. \
O g
% \
-t O cn_
we s ,
\
-}.'
O \ M j
$8 '
s N s 4 8- s s Li
- t
\ -
s , 8
- s
\
1
- [
C g .i g- s j
.s
's i
's ' !
8 i
. 00 l'o.oo- 2'o.oo - 3'o.oo 4'o.oo s'o.oo - 6'o.oo d TIME- (SEC) :!
i
' SOLIDLINE = VERSION 1' SINGLE NODE SG. 'l
-t DASHED LINE = VERS 10N l 1 - t1ULTI-NODE SG :i
't l
~
. 2/10/93 20.02,32 -l 1
-4 -
f p, =
, F1GURE 55 l LOSS OF NORMAL FEEDWATER l PRESSURIZER PRESSURE -
i 8 5 o o , Y. ru O 70 '
& i .
b~ /
2
FIGURE 56 LOSS OF NORMAL FEEDWATER PRESSURIZER LIQUID VOLUME '
[
8, 8.
r 8
o$-
) i i
8 - i i R- 8 i :
>- 1 b
w.8 , E o. ges - _; ,'i . O L / i
> t i
o ' oo 1 i
~ o. t / \ 1 D50 i \ 1 o
ig t
/
f s ' 1 g
._J / g o <
f go g j l
' I ES- ~'
\
\
8 s o,
.......q
.....q . . .
....q.
TIME (SEC) SOLID LINE = EARLIER MODEL DASHED LINE
- f t 9S10N 1 2/10/93 20.05.52
\
F1GURE 57 LOSS OF NORMAL FEEDWATER LOOP A COLD LEG TEMPERATURE 8 id
- g. ,
8 5 O' g N I l 8 l 1
,'\ > n 10 .
' \
l i c l' \ f g S 't , 1 o8 w n , i i Sr?. l l I c r , s O l t I I \ OO i 09 I i ng* t
/
\
sn I s I 8 e f.
/
/
l ' O _ _ _ _ _ _ _ _ _. ' n m.
.......q o . . .
.....q , . . .
.....q . . _. .
.....q , ;
TIME (SEC)
. SOLID LINE = EARLIER MODEL t DASHED LINE = VERSION 1 2/10/93 20.05.52-
FIGURE 58 ' LOSS OF NORMAL FEEDWATER LOOP A HOT LEG TEMPERATURE 8 5 8-8 5 3-o 9 -------- -y, 8. e m u_ o Go j Wd e8- . H Oo Io 8- ,s / - l s t
, s -r 8 ,
s d 8-8 t i og. . . . . . . . q g, . . . . . . . . g . . . . . . . . g . . . . . . . 3 g, TINE CSEC) SOLID LINE = EARLIER MODEL DASHED LINE = VERSION 1 2/10/93 20.05,52
.o FIGURE 59 LOSS OF NORMAL-FEEDWATER LOOP B COLD LEG TEMPERATURE- ,
8 8. n 4 8 as d i V -i
$" #) I i .;
I I I I i i ' ' 8 ti s i 4 i i , i i- ' b- 1 1
,- i i
-, i LL
, 1
' .i :
! .g
\ g 08
' id -
l g I l
' l 93.ut t l-i .t I- I o l
.{
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/. .i-IF i I
, i.
, t :
y . 8 l O. I o i 1 i 8 i !
- i t g _ _ _ _ _ _ _ _ _ .
.....q ,.. . .
.....q . .. . ..... . . . . . . . . .;
TIl1E (SEC) 4 i SOLID L INE-= EARLIER ' t10 DEL l DASHED'LINE.= VERSION.1 j l I
=2/10/93 :20.0".52' ' , . IV .
p r FIGURE 60 LOSS OF NORMAL FEEDWATER LOOP B HOT LEG TEMPERATURE 8
'd 8-8 5
3-o o
----------f
,8.
C w8 ; Wd 98-r oo IO
&0 8- ,, i -
/%
r 8 \ d 8-8 5 a j . . .
.....q , .
. . ..... g . . . ..... g . . . ..... .
TIME (SEC) SOLID LINE = EARLIER MODEL DASHED LINE = VERSION 1 1 2/10/93 20.05.52
FIGURE 61 LOSS OF NORMAL FEEDWATER-LOOP A SG LIQUID-VOLUME 8 5 Tu-8
.b. ..
o
.g ,
d : q 9-H bo 9 o w$- r a \ I o '
>o s o9 .
s O s. s . O s .
~
s o 'g O \ OS- \
\
\
\
\
o . . .
.....q , . . . ..... . .....g . . . . . . . , ,
TIME (SEC) . SOLID LINE = EARLIER MODEL DASHED LINE = VERSION 1 t 2/t0/93 20.05.52 L. .. _ 1
FIGURE 62 LOSS-OF NORMAL FEEDWATER LOOP B SG LIQUID VOLUME 8 d 7v-8 5 8-O
*O 9
8. R~ 6-- bo 1
^
9 o i w$- r D I J O I >o ' x o90. s N s i D % y I .O 's 1 3, - _= g o . (l@- 8 6 . . . . . . . . ,.
.......g........qj........ .
TIME (SEC) SOLID LINE = EARLIER MODEL. - DASHED LINE = VERSION 1 r 2/10/93 '20.05.52
.t t,
e . FIGURE 63 LOSS: OF = NORMAL FEEDWATER LOOP A SG STEAM PRESSURE-8 8 8 is
' ~ ~
_s _
\
28 l-58
&;9-w8 x
D-u)? U)o. w-Q.- r8 w8. rm u) Oo u)o I
. t i
/
8 8 e . . .
.....q ,.. . . ..... .. . . ..... . . . . . . . .
TIME (SEC1 > v. SOLID 'LINE = EARLIER t10 DEL - DASHED LINE'=' VERSION 1 p 2/10/93' 20.05.52. L .. -
^^
m: . FIGURE 64 LOSS OF NORMAL FEEDWATER LOOP B SG STEAM' PRESSURE : O o 8 o fg
. s O v mO
<O -
.-8 m
69-WO To o-m? M O. w- ; CL O lt O w8. rm 00 WO i I O / h" /
/
O / 8 ej. . .
.....q , . . .
.....q ,
.......q . . . . . . . . ,
TIME CSEC) SOLID LINE = EARLIER MODEL DASHED LINE =. VERSION 1 ; 6 2/I0/93 20.05.52 L -_..
FIGURE 65 ; LOCKED ROTOR OVERPRESSURIZATION CASE LOOP 3 COLD LEG PRESSURE 8 5 Ts. N
- \
/ '
O O / \ f \ 8- , \
\
/
28 [ \ MS ' s Eb8- I i i
\ $8 -
i 1
\
\
a? W I \
\
$N-g I
\
\
Q. o 1
/
e s W8 s Jg- \ o \- J \ i Ok 9 ', N- \
\
O \ /* \ O \ f s' s,
%.00 2'.00 4'.00 6'.00 8'.00 1'O.00 1'2.00-TIME (SEC) 1 SCLID LINE - EARLIER MODEL DASHED LINE - VERSION 1 2/10/93 20.29,32
-l
'l 4 i
- i FIGURE'66 LOCKED ROTOR OVERPRESSURIZATION CASE NORl1ALIZED TOTAL RCS 11 ASS FLOW RATE 5
8 d' E 26-O J' U)o u)e ' E d' : i
. s i r s E ru \
os s z g- s N , s
\
E., ___ ! O
, t k.00 2'.00 4'.00 6'.00 8'.00 1'O.00 1'2.00 Tll1E (SEC)
SOLID LINE '- EARLIER t10 DEL DASHED LINE - VERSION 1 2/10/93 20 29.32 a1.
W: p , FIGURE 67- ! LOCKED ROTOR OVERPRESSURIZATION CASE- l LOOP 1 NORMALIZED MASS FLOW RATE i t 8 l 8 ~
.l 8 :
r d- [
.o \
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FIGURE 68 LOCKED ROTOR-OVERPRESSURIZATION CASE NORf1ALIZED CORE POWER 8 .
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i ~ FIGURE 69 . LOCKED ROTOR OVERPRESSURIZATION CASE , DOPPLER REACTIVITY 8 d'
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FIGURE 70- - LOCKED ROTOR OVERPRESSURIZATION. CASE-LOOP 3 COLD LEG-TEMPERATURE 8 . 5 8- ; 8 i d / Q- t
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t FIGURE 71 NAPS 1987 STEAL 1 GENERATOR TUBE RUPTURE EVENT-PRESSURIZER PRESSURE . 8 < d C. ^ a 8
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- l FIGURE 72 !
NAPS 1987 STEAM. GENERATOR. TUBE RUPTURE EVENT-i PRESSURIZER LIOUID LEVEL , l 8 ; d N-8 ; 8. 8 ;
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. I FIGURE 73 3
NAPS 1987 STEAM GENERATOR TUBE RUPTURE- EVENT: - l SG "A" AND "B" SECONDARY STEAM PRESSURE .i a 8 : 5
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g F1GURE 74 NAPS 1987 STEAM GENERATOR TUBE RUPTURE EVENT SG "C" SECONDARY STEAL 1 PRESSURE 8 5 o. 8 5 -
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p FIGURE 75 NAPS 1987 STEAM GENERATOR _ TUBE RUPTURE EVENT , LOOP "A" AND "B" COLD LEG TEMPERATURE-8 b
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FIGURE 76 NAPS 1987 STEAM GENERATOR TUBE RUPTURE EVENT , LOOP "C" COLD LEG TEMPERATURE 8 5 8-8 5 I 8-A
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FIGURE 77 TURBINE RUNBACK 100% TO 60% 200%/ MIN BOC COMPARISON o PRESSURIZER PRESSURE 9 8 N-8 6 M.
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'l/22/93 13.44.46
FIGURE 78 TURBINE RUNBACK 100% TO 60% 200%/ MIN BOC COMPARISON PRESSURIZER LIQUID LEVEL 8 8 8 56 m W \
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FIGURE 79 TURBINE RUNBACK 100% T0 60% 200%/ MIN BOC COMPARISON NEUTRON POWER 8 4
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'1/22/93 13.44.46
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L 7 FIGURE 81 U TURBINE RUNBACK 100% TO 60% 200%/ MIN BOC COMPARISON STEAM GENERATOR STEAM PRESSURE 8. m i 8 5 8-2oO md (b&- /
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FIGURE 82 TURBINE RUNBACK 100% TO 60% 200%/ MIN BOC COMPARISON STEAM GENERATOR NARROW RANGE LEVEL 8 5-8 5-r $8 o- d_ we w '
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FIGURE 83 TURBINE RUNBACK 100% TO 60% 200%/ MIN BOC COMPARISON . TURBINE MASS FLOW RATE 8 4 9-m e. O* w l-x8 b- b ' e - a w9 Q8- a x a 3k u.g. w z @8 ' ?$- 8 1 5 '
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i FIGURE 84 - TURBINE RUNBACK 100% TO 60% i 200%/ MIN BOC COMPARISON STEAM MASS FLOW RATE ! 3. o_ m O w H i "8d_ u_ a o k
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DASHED LINE = SINGLE NODE SG 1/22/93 14.26.30
FIGURE 85 : TURBINE RUNBACK-100% TO 60% ' 200%/ MIN BOC COMPARISON TOTAL FEEDWATER MASS FLOW RATE 8 6 - 8 Eh-w H "o - u.R ?'t 08-i t i d ! 'i . W8 ' < o. ec sjt : 2 o \
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FIGURE 86 ! TURBINE RUNBACK 100% TO 60% i 200%/ MIN BOC COMPARISON j CONDENSER DUMP MASS FLOW. RATE 8
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1/22/93 14.31.03 ,
F1GURE 87 TURBINE RUNBACK 100% TO 60% 200%/ MIN BOC COMPARISON DOPPLER FEEDBACK REACTIVITY 8 6
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t F1GURE 88 ! TURBINE RUNBACK 100% TO 60% 200%/ MIN BOC COMPARISON l MODERATOR FEEDBACK REACTIVITY ! 8 i 8. 8 i d S.'- ; y , O - 68 l o.
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i 1/22/93 13.44.46 I
T 1 l FIGURE 89 TURBINE RUNBACK 100% TO 60% 200%/ MIN BOC COMPARISON ROD CONTROL REACTIVITY 8 d' 8 95 E 08 (b ' . [0 E8. s' o "85 f. o' x Zo s 09 s ao N
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