ML20087J819
| ML20087J819 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 03/09/1984 |
| From: | Lindquist T, Nutt W SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML18051A825 | List: |
| References | |
| XN-NF-84-18, NUDOCS 8403230106 | |
| Download: ML20087J819 (150) | |
Text
~
r XN NF 8418
[
[
f PLANT TRANSIEN" ANALYSIS FOR PALISADES NUCLEAR POWER PLANT WITH I
50% STEAM GENERATOR PLUGGING l
l l
MARCH 1984 l
l RICHLAND, WA 99352 ERON NUCLEAR COMPANY,INC.
l l$RE88ER*ois88[tg
5 r'
5 XN-NF-84-18
[ , Issue Date: 3/9/84 l
l PLANT TRANSIENT ANALYSIS FOR PALISADES NUCLEAR POWER PLANT WITH 50% STEAM GENERATOR PLUGGING
{
Prepared by: , 3[7[64-T. R. Lindquiyt / /
PWR Safety Arfalysis Prepared by:
. T. Nutt I/ Mk
//-
f/W PWR Safety Analy s Concur: % Y8/P7 W. V. Kayser, Manager PWR Safety Analysis Concur: - 8 J. C. Chandler, Lead Engineer Reload Fuel Licensing Concur: M, f_ 3/.r/.h/
/ '
J. tg'1 organ, Manager Propasals & Customer Services Engineering
^
!I .}
Approve: -
[/M%?/y R. B. Stout, Manager Licensing & Safety Engineering Approve: 'Y/ .b e e. . , _
f,. I'/'/ V G. ' A. Sofer, Manager /
Fuel Engineering & Technical Services gf ERON NUCLEAR COMPANY,Inc.
NUCLEAR REGULATORY COMMIS$10N DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was derived through research and development programs sponsored by Exxon Nuclear Company, Inc. It is being sub-mitted by Exxon Nuclear to the USNRC as part of a technical contri-bucon to facilitate safety analyses by licensees of the USNRC which utilize Exxon Nuclear-fabricated reload fuel or other technical services provided by Exxon Nuclear for licht water power reactors and it is true and correct to the best of Exxon Nuclear's knowledge, information, and belief. The information contained herein may be used by the USNRC in its review of this report, and by licensees or applicants before the USNRC which are customers of Exxon Nuclear in their demonstration of compliarms with the USN RC's regulations.
Without derogating from the foregoing neither Exxon Nuclear nor any person acting nn its behaif:
A. Makes any warranty, express or implied, with respect to the accuracy, completeness, or usefulness of the infor-mation contained in this document, or that the use of any informatiort apparatus, method, or process disclosed in this document will not infringe privately owned rights; or B. Assumes any liabilities with respect to the use of, or for darrages resulting from the use of, any information, ap-paratus, method, or process disclosed in this document.
XN- NF- F00,766 s
L i XN-NF-84-18 L
i
( TABLE OF CONTENTS I Section Page
1.0 INTRODUCTION
AND
SUMMARY
.................................... 1 2.0 CALCULATIONAL METHODS AND INPUT PARAMETERS. . . . . . . . . . . . . . . . . . 5 l
2.1 CODE DESCRIPTION....................................... 5 2.2 MODELING UNCERTAINTIES................................. 7 2.3 DESIGN PARAMETERS...................................... 8
- 3. 0 TRAN S I E NT ANALYS I S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 3.1 ANTICIPATED OPERATIONAL OCCURRENCES REQUIRING ONLY RPS ACTI0N........................................ 19 3.1.1 Loss of Load Event.............................. 19 j 3.1.2 Excess Load Event............................... 20 3.1.3 RCS Depressurization Event...................... 21 3.2 ANTICIPATED OPERATIONAL OCCURRENCES REQUIRING RPS ACTION AND/0R OBSERVANCE OF THE LC0s................... 22 3.2.1 Loss-of-Coolant-Flow Event...................... 22 3.2.2 CEA-Withdrawal Event............................ 23 3.2.2.1 Rod Withdrawals From Full Power........ 24 3.2.2.2 Rod Withdrawals From Part-Power........
25 i
3.2.3 CEA Drop Event.................................. 27 s
( 3.3 POSTULATED ACCIDENTS................................... 29
( 3.3.1. Primary-Pump-Seizure Event...................... 29 3.3.2 Loss of Feedwater with a Loss of Offsite Power.. 30 -
4 . 0 D I S CU S S I O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 138
5.0 REFERENCES
.................................................. 139 L-
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h 11 XN-NF-84-18
(
(
( LIST OF TABLES
( Table Page 1.1 Fuel and Vessel Des ign L imits . . . . . . . . . . . . . . . . . . . . . 3 1.2 S um a ry o f Re s u l t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
( 2.1 Trip Setpoints for Operation of Palisades
( Reactor at 2125 MWt ............................... 9 2.2 Nominal Operating Parameters Used in the i l
Transient Analysis of Palisades at 2125 MWt ....... 10 l I
2.3 Palisades Fuel Design Parameters for !
Exxon Nuclear Fuel ................................ 11 2.4 Kinetics Parameters ............................... 12 3.1 Transient Events .................................. 33 g 3.2 Index of Symbols .................................. 34 1
3.3 Event Table for the Loss of Electric Load ......... 36 3.4 Event Table for the Excess Load ................... 37 3.5 XCOBRA-IIIC Input for Excess Load ................. 38
( 3.6 Event Table for the PORV Failure .................. 39 3.7 XCOBRA-IIIC Input for PORV Failure ................ 40 f 3.8 Event Table for the Four Pump Coastdown ........... 41
[ 3.9 XCOBRA-IIIC Input for Four Pump Coastdown ......... 42 L
3.10 Event Table for Fast Rod Withdrawal from 100% Power ........................................ 43 3.11 XCOBRA-IIIC Input for Fast Rod Withdrawal from 100% Power ................................... 44 3.12 Event Table for the Slow Rod Withdrawal from 100% Power ................................... 45
( '3.13 XCOBRA-IIIC Input for Slow Rod Withdrawal from 100% Power ................................... 46
iii XN-NF-84-18 LIST OF TABLES (Cont.) }
Table Page 3.14 Event Table for Fast Rod Withdrawal from 50% Power ..................................... 47 3.15 XCOBRA-IIIC Input for Fast Rod Withdrawal from 50% Power ..................................... 48 3.16 Event Table for Slow Rod Withdrawal from 50% Power ..................................... 49 3.17 XCOBRA-IIIC Input for Slow Rod Withdrawal 4 from 50% Power ..................................... 50 3.18 Event Table for CEA Drop ........................... 51 3.19 XCOBRA-IIIC Input for CEA Drop ..................... 52 3.20 Event Table for the Locked Rotor ................... 53 3.21 XCOBRA-IIIC Input for the Locked Rotor ............. 54 3.22 Event Table for Loss of Feedwater with Loss of Offsite Power .............................. 55 i
iv XN-NF-84-18
( LIST OF FIGURES 1
Figure Page 2.1 PTSPWR2 System Model ............................... 13 2.2 Limiting Condition for Operation Based on L inear Heat Generat ion Rate . . . . . . . . . . . . . . . . . . . . . . . . 14 2.3 Scram Curve for Palisades .......................... 15 2.4 Limiting Axial Shape for MDNBR Calculations at 100% Power ...................................... 16 2.5 Limiting Axial Shape for MDNCR Calculations at 50% Power ....................................... 17 3.1 Steam Generator Flows for Loss of Electric Load .... 56 4 3.2 Primary Loop Temperature for Loss of Electric Load ...................................... 57
[ 3.3 Pressurizer Flows for Loss of Electric Load ........ 58 3.4 Liquid Volume in Pressurizer for Loss
( of Electric Load ................................... 59 3.5 Pressurizer Pressure for Loss of Electric Load ..... 60
( 3.6 Reactivities for Loss of Electric Load ............. 61 3.7 Reactor Power for Loss of Electric Load . . .. . . . . . . . . 62
( 3.8 Core Heat Flux for Loss of Electric Load ........... 63 3.9 Steam Line Flows for Excess Load ................... 64 3.10 Primary Loop Temperatures for Excess Load .......... 65 3.11 Liquid Volume in Pressurizer for Excess Load ....... 66 3.12 Pressurizer Pressure for Excess Load . . . . . . . . . . . . . . . 67 3.13 Core Inlet Temperature for Excess Load ............. 68 3.14 Core Flow for Excess Load .......................... 69
y XN-NF-84-18 LIST OF FIGURES (Cont.)
Figure Page 3.15 Reactivities for Excess Load ....................... 70 3.16 Reactor Power for Excess Load ...................... 71 3.17 Core Heat Flux for Excess Load ..................... 72 3.18 Pressurizer Relief Valve Flow for the P ORV F a i l u re . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 73 3.19 Pressurizer Pressure for the PORV Failure . . ... .. . .. 74 3.20 Reactor Thermal Power for the PORV Failure ......... 75 3.21 Reactor Heat Flux for the PORV Failure . . . .. . . . . . . . . 76 3.22 Core Inlet Temperature for the PORV Failure ........ 77 3.23 Core Flow for the PORV Failure ..................... 78 3.24 Core Flow for the Four-Pump Coastdown .............. 79 3.25 Reactor Thermal Power for the Four-Pump Coastdown .......................................... 80 3.26 Core Heat Flux for the Four-Pump Coastdown ......... 81 <
3.27 Primary Loop Temperatures for the Four-Pump Coastdown ................................ 82 3.28 Liquid Volume in Pressurizer for the Four-Pump Coastdown ................................ 83 s
3.29 Pressurizer Pressure for the Four-Pump Coastdown ... 84 3.30 Core Inlet Temperature for the Four-Pump Coastdown ................................ 85 3.31 Reactivities for Fast Rod Withdrawal from 100% Power ..................................... 86 3.32 Reactor Thermal Power for Fast Rod Withdrawal from 100% Power ......................... 87 I.
1
b vi XN-NF-84-18 i
5
{ LIST OF FIGURES (Cont.)
( Figure Page 3.33 Core Heat Flux for Fast Rod Withdrawal
( from 100% Power .................................... 88 3.34 Primary Loop Temperatures for Fast Rod
[ Withdrawal from 100% Power ......................... 89 3.35 Pressurizer Pressure for Fast Rod Withdrawal from 100% Power ......................... 90 3.36 Core Flow for Fast Rod Withdrawal from 100% Power ......................................... 91 3.37 Core Inlet Temperature for Fast Rod Withdrawal from 100% Power ......................... 92 3.38 Reactivities for Slow Rod Withdrawal from 100% Power .................................... 93 3.39 Reactor Power for Slow Rod Withdrawal from 100% Power .................................... 94 3.40
( Core Heat Flux for Slow Rod Withdrawal from 100% Power .................................... 95 3.41 Primary Loop Temperatures for Slow Rod W ithdrawal f rom 100% Power . . . . . . . . . . . . . . . . . . . . . . . . . 96 3.42 Liquid Levels for Slow Rod Withdrawal
( from 100% Power .................................... 97 3.43 Pressurizer Pressure for Slow Rod Withdrawal from 100% Power ......................... 98 3.44 Core Flow for Slow Rod Withdrawal from 100% Power .. 99 3.45 Core Inlet Temperature for Slow Rod Withdrawal from 100% Power.......................... 100 3.46 Reactivities for Fast Rod Withdrawal
( from 50% Power ..................................... 101 i
3.47 Reactor Thermal Power for Fast Rod Withdrawal from 50% Power .......................... 102
vii XN-NF-84-18 LIST OF FIGURES (Cont.)
Figure Page 3.48 Core Heat Flux for Fast Rod Withdrawal from 50% Power ............................... ..... 103 3.49 Primary Loop Temperatures for Fast Rod W i thdrawal f rom 50% Power . . . . . . . . . . . . . . . . . . . . . . . . . . 104 3.50 Pressurizer Pressure for Fast Rod W ithdrawal f rom 50% Power . . . . . . . . . . . . . . . . . . . . . . . . . . 105 3.51 Core Flow for Fast Rod Withdrawal from 50% Power ... 106 3.59. Core Inlet Temperature for Fast Rod Withdrawal from 50% Power .......................... 107 3.53 Reactivities for Slow Rod Withdrawal .
from 50% Power ..................................... 108 3.54 Reactor Power for Slow Rod Withdrawal from 50% Power ..................................... 109 3.55 Reactor Heat Flux for Slow Rod Withdrawal l from 50% Power ..................................... 110 J 3.56 Steam Generator Pressure for Slow Rod W ithdrawal f rom 50% Power . . . . . . . . . . . . . . . . . . . . . . . . . . 111 3.57 Primary Loop Temperatures for Slow fiod W ithdrawal f rom 50% Powe r . . . . . . . . . . . . . . . . . . . . . . . . . . 112
]
3.58 Liquid Levels for Slow Rod Withdrawal from 50% Power ..................................... 113 '
3.59 Pressurizer Pressure for Slow Rod W ithdrawal f rom 50% Power . . . . . . . . . . . . . . . . . . . . . . . . . . 114 3.60 Core Flow for Slow Rod Withdrawal from 50% Power .......................................... 115 3.61 Core Inlet Temperature for Slow Rod W i thdrawal f rom 50% Power . . . . . . . . . . . . . . . . . . . . . . . . . . 116 3.62 Reactivities for CEA Drop .......................... 117 1
J 1
l
________-____-_a
viii XN-NF-84-18
( LIST OF FIGURES (Cont.)
( Figure Page
( 3.63 Reactor Power for CEA Drop ......................... 118 3.64 Reactor Heat Flux for CEA Drop ..................... 119
( 3.65 Turbine Flow for CEA Drop .......................... 120 3.66 Core Inlet Temperature for CEA Drop ................ 121 3.67 Volume of Water in Pressurizer for CEA Drop ........ 122 3.68 Pressurizer Pressure for CEA Drop . . . . . . . . . . . . . . . . . . 123 3.69 Core Flow for CEA Drop ............................. 124 3.70 Core Flow for the Locked Rotor ..................... 125 3.71 Reactor Power for the Locked Rotor ................. 126 3.72 Reactor Heat Flux for the Locked Rotor ............. 127 3.73 Core Inlet Temperature for the Locked Rotor ........ 128 3.74 Primary Loop Temperatures for the Locked Rotor ..... 129
( 3.75 Pressurizer Surge Flow for the Locked Rotor ........ 130 3.76 Pressurizer Water Volume for the Locked Rotor ...... 131 3.77 Pressurizer Pressure for the Locked Rotor ....>..... 132 3.78 Core Power for Loss of Normal Feedwater ............ 133 3.79 RCS Flow Rates for Loss of Normal Feedwater ........ 134 3.80 -Steam Generator Liquid Levels for Loss of Normal Feedwater ........................... 135 3.81 RCS Temperature for Loss-of Normal Feedwater ....... 136 3.82 Pressurizer Liquid Level for Loss of Normal Feedwater ................................ 137
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[ l XN-NF-84-18
1.0 INTRODUCTION
AND
SUMMARY
Recent analyses (1,2) have addressed the limiting DNBR and pressure transients for Palisades at 2530 MWt with plugging levels up to 21%. The plant transient analysis reported here was performed to support operation of the
(
Palisades Nuclear Power Plant for Cycle 6 at a power of 2125 MWt with 50% of the steam tubes plugged. The purpose of the analysis is to demonstrate that the plant protection system (PPS) protects the specified acceptable fuel design limits (SAFDLs) and vessel pressure design limits given in Table 1.1 for anticipated operational occurrences (A00s) and that the postulated accidents (pas) do not violate the criteria on fuel damage or vessel pressure I given in Table 1.1. Further, it is demonstrated that the reactor meets the appropriate criteria for a loss of feedwater due to rupture of a feedwater
{ line with loss of offsite power.
The present analysis provides a verification of the thermal margin using
{
PTSPWR2(3) to simulate the plant response and XCOBRA-IIIC(4) to calculate the ;
f local coolant conditions in the core based on the plant response. The MDNBR is obtained from the local core conditions and the Exxon Nuclear DNB correlation,XNB(5). The fuel failure criterion is verified by calculating the number of fuel rods expected to experience DNB. Asyrmietric plugging levels are accounted for by including a 50F inlet temperature penalty in the f DNBR calculation. The calculations support up to a 20% difference in plugging between the two steam generators.
The loss of Normal Feedwater transient, coincident with a station blackout and the loss of auxiliary feedwater to the least plugged steam generator, was simulated with SLOTRAX.(6) The intent of this simulation is to
2 XN-NF-84-18
~
assure adequate decay heat removal and to prevent the loss of primary coolant inventory through the pressurizer relief valves.
A description of the transient calculational methods is provided in Section 2.0. The transient events analyzed for operation at 2125 MWt are the most limiting events in terms of DNBR and pressure and comprise an adequate set of simulations to assure operation within the criteria of Table 1.1 for Cycle 6. The simulations of the limiting transients are discussed in Section 3.0. Section 4.0 provides a rationale for not analyzing the entire spectrum of transients in the Standard Review Plan (SRP). since they are bounded by prior analyses or by transients discussed in Section 3.0.
The key results of the analysis are summarized in Table 1.2 and confirm the criteria of Table 1.1 are met. The lowest value of MDNBR for any A00 is 1.372 for the control element assembly (CEA) drop event. The loss of ]
feedwater transient resulted in a more severe pressurizer transient. How-ever, the pressurizer did not fill with water and long term decay heat removal was established.
The analysis of the limiting transients has shown that the PPS settings provide the level of protection required by Table 1.1, that the transient allowances (7) in the thermal margin / low pressure (TM/LP) trip are appro-priate, and that the limiting conditions for operation (LCOs) are sufficient to provide protection for those transients requiring them. Thus, the analysis supports operation of Palisades at 2125 MWt in Cycle 6 with 50T of the steam generator tubes plugged.
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[ 3 XN-NF-84-18
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[ Table 1.1 Fuel and Vessel Design Limits l
[
Event Class Criteria
(
Anticipated Operational . Specified acceptable fuel Occurrences (A00s) design limits (SAFDLs)
. MDNBR, based on XNB, >1.17
. Local power density 21 kW/ft
. Pressure < 2750 psia
[ Postulated Accident (PA) . Fuel damage is limited to a small fraction of the fuel in the core
( . Pressure < 2750 psia
(
(
j 1
f
l l
l Table 1.2 Summary of Results Maximum Maximum Maximum Maximum Power Core Pressurizer Primary to Level Average Flux Pressure Secondary Transient (MWt) 2 (Btu /hr-ft ) (psia) AP (psi) MDNBR CEA Drop 2199 144,344 1950 1383 1.372 Four Pump Coastdown 2160 139,666 2008 1347 1.579 l
PORV Failure 2218 144,452 1950 1347 1.636 Excess Load 2504 148,709 1950 1534 1.782 Loss of Electric Load 2260 142,443 2500 1527 1.828 ,
Uncontrolled Rod Withdrawal Rod Withdrawal 9 6x10-4 2671 150,096 2008 1350 1.679 Ap/sec from 2125 MWt Rod Withdrawal 9 2.5x10-5 2388 152,907 2105 1436 1.674 Ap/sec from 2125 MWt Rod Withdrawal 0 6x10-4 1631 83,844 1997 1208 2.007 Ap/sec from 1062.5 MWt Rod Withdrawal 0 5x10-5 1286 82,584 2312 1308 1.664 Ap/sec from 1%2.5 MWt Locked Rotor 2198 139.667 2053 1347 1.523 5
l l
2
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5 XN-NF-84-18
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( 2.0 CALCULATIONAL METHODS AND INPUT PARAMETERS 2.1 CODE DESCRIPTION
[ The transient analysis for Palisades was performed using PTS-PWR2(3), the Exxon Nuclear Company plant transient simulation model for
( pressurized water reactors. The simulation code models the behavior of pressurized water reactors under both normal and abnormal conditions by solving the transient conservation equations for the primary and secondary f systems numerically. Core neutronics behavior is modeled using point k inet ics, and the transient conduction equation is solved for fuel tem-peratures and heat fluxes. State variables such as flow, pressure, temperature, mass inventory, steam quality, heat flux, reactor power and reactivity are calculated during the transient. Where appropriate the
[ reactor protection system (RPS) and control system are modeled to describe the
, transients.
The system model used by PTSPWR2, shown in Figure 2.1, models the reactor, both primary coolant loops, both steam generators and both steam lines. All major components (pressurized, coolant pumps, and all major
( valves) are also modeled.
The present calculations were performed using the NOV76A version of the PTSPWR2 code, along with appropriate updates. These updates include:
(1) A correction to the mass balance on the secondary side of the steam
( generator. 1 I
(2) An improved pressurizer model. '
(3) A modified set of trip functions to . describe - a Combustion Engineering plant.
6 XN-NF-84-18 (4) A dynamic flow coastdown model.
(5) Appropriate changes to the primary loop hydraulic behavior to describe the 2 hot leg - 4 cold leg configuration of Palisades.
Updates 1 and 2 were documented in Reference 3. Updates 3-5 were prepared specifically for this analysis.
For Palisades the calculated Thermal Margin / Low Pressure (TM/LP) trip is used in conjunction with the limiting conditions of operation (LCOs) to protect the specified acceptable fuel design limits (SAFDLs) based on departure from nucleate boiling (DNB). The DNB SAFDL l imit is further protected by several trips based on single state variables. These latter trip setpoints are listed in Table 2.1 along with the uncertainties and the trip .
time delays appropriate for each of the RPS trips. The TM/LP trip depends on the hot leg temperature, TH and the cold leg, TC . The form of the trip function is:
18.8269 TH - 1.2944TC - 8587.479 Power < 100% '
P VAR " 23.9615 TH - 6.1932TC - 8970.464 Power > 100%
Pressurizer pressure is the system variable which is compared to the trip setpoint, PVAR. The TM/LP trip protects the core from the onset of DNB with at least a 95% probability as long as the plant is operated within the limiting conditions of operation (LCOs), including the LC0 on peak linear heat generation rate (LHGR) shown in Figure 2.2.
The scram curve shown in Figure 2.3 was used in the plant transient simulations. The time in the figure is measured from the holding coil I
release.
1 a
I
[ 7 XN-NF-84-18
[
I The pump response to a loss of power was modeled by setting the shaf t
{
rotation speed derivative equal to the pumping torque, divided by the
( effective inertia. The flow in each of the four cold legs was calculated based on the pump head and the required pressure drop. The effective inertia was
[ then adjusted to provide a good fit to plant data for operation in prior cycles at lower plugging levels (l). The loop pressure drop was then adjusted to provide 99 M1b/hr vessel flow for 50% tube plugging. This caused the plant to
( balance at full pump speed with reduced flow. The pump coastdown transient behavior was then determined by the effective inertia previously determined.
DNBR calculations were performed using XCOBRA-IIIC and Exxon Nuclear's critical heat flux correlation, XNB. The boundary conditions, core flow, inlet temperature, heat flux, and pressure were taken from the PTSPWR2
( simulation at the time of MDNBR, as predicted by the hot channel model in PTSPWR2, and used as input to XCOBRA-IIIC runs. Figures 2.4 and 2.5 are axial shapes which satisfy the LC0 on LHGR, Figure 2.1, with maximum radial peaking for 100% power and 50% power, respectively. Part power transients were assumed to retain the peaking shown in Figure 2.5. The parameter uncer-tainties described in Section 2.2 were applied to the boundary conditions for
(
MDNBR calculation except in the case in which the trip occurred on the f parameter. As an example, for transients terminated by the high flux trip, the value of heat flux calculated by PTSPWR2 at the time of MDNBR was used
( directly since the power errors, 2% calorimetric +3.5% transient allowance, were already included in the heat flux value via the trip function.
2.2 MODELING UNCERTAINTIES The present plant transient analysis is a deterministic analysis.
Thus, steady state measurement and instrumentation errors were taken into
r J
8 XN-NF-84-18 account in an additive f ashion to ensure conservative calculations of MDNBR.
The plant uncertainties related to initial conditions in the MDNBR calcu-lations are:
Power + 2% for calorimetric error Inlet coolant temperature + 70F for deadband and measurement error and asymmetric steam generator plugging RCS pressure - 22 psi for steady-state measurement errors.
Combined with minimum design flow and peaking uncertainties, these parameter uncertainties conservatively bound the MDNBR. Uncertainties are accounted for in the trip functions and in the XCOBRA-IIIC analyses. Table 2.2 is a list of operating parameters used in this analysis.
The trip setpoints are summarized in Table 2.1. A verification of f
the TM/LP is presented in Reference 7. The TM/LP setpoint was modeled conser-vatively in the transient analysis to provide bounding simulations of the RPS response. This was done by including the effects of hot leg and cold leg temperature errors, thermal power calibration, asymmetric inlet temperature
]
allowance and trip pressure bias.
The pressurizer control system was modeled in such a f ashion that it could not mitigate the effects of transients. The spray system was operable during DNBR transients while the heaters were off, thus tending to minimize DNBR. For pressurization transients, e.g., loss-of-electric load, the spray system and pressurizer relief . valves were removed from the simulation.
2.3 DESIGN PARAMETERS The ENC fuel design parameters for Palisades are summarized in Table j
2.3. Table 2.4 lists 'the bounding values for neutronics parameters, for beginning of cycle (B0C) and end of cycle (E0C).
__j
- v - -
- _ .- em e .r _ v Table 2.1 Trip Setpoints For Operation of Palisades Reactor at 2125 Nt.
Setpoint Uncertainty used in Analysis Delay Time High Neutron Flux 106.5% 1 5.5% 112.0% 0.4 sec
' Low Reactor Coolant Flow 95% i 2.0% 93.0% 0.6 sec High Pressurizer Pressure 2255 psia + 22 psi 2277 psia 0.6 sec l Low Pressurizer Pressure 1750 psia 1 22 psi 1728 psia 0.6 sec Low Steam Generator Pressure 400 psia + 22 psi 78 psia 0.6 sec ,
Low Steam Generator Level
- 6 feet i 10 in 6 feet 10 in. 0.6 sec l
l
- Below operating level.
5
$i a:
10 XN-NF-84-18 _
Table 2.2 Nominal Operating Parameters Used In the Transient Analysis of Palisades at 2125 MWt Core Total Core Heat Output, MWt 2125 Total Core Heat Output, MBtu/hr 7252.7 Heat Generated in Fuel, % 97.5 System Pressure, psia 1950 Total Coolant Flow Rate, Mlbs/hr 99 Effective Core Flow Rate, Mlbs/hr 95.1 Core Inlet Coolant Temperature, OF 535 Average Core Coolant Temperature, OF 564 Hot Channel Factors s TotalPeakingFactor,F{ FA x Fr xFLxFE = 2.607 A
Radial Peaking Factor, F 1.500 Axial Peaking Factor, FZ 1.544 Local Interior Peaking Factor, FL 1.093 Engineering Factor, FE 1.030 Heat Transfer Core Average Heat Flux, Btu /hr-ft 2 139,671 Steam Generators Total Steam Flow, M1bs/hr 8.97 Secondary Steam Pressure, psia 600 Feedwater Temperature, OF 395 l Number of active Steam Generator Tubes, S.G.#1(40% Plugging) Sill S.G.#2 (60% Plugging) 3408
-- . - i ..i
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11 XN-NF-84-18
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b, u Table 2.3 Palisades Fuel Design Parameters for Exxon Nuclear Fuel
(
Fuel Radius 0.175 inches Inner Clad Diameter 0.357 inches Outer Clad Diameter 0.417 inches Active Length 131.8 inches Active Fuel Rods Per Bundle 208
(
\
s %
w h- .,
x ~~
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r ,
, l 12 XN-NF-84-18
]
Table 2.4 Kinetics Parameters.
]
Symbol Parameter Value
]
Beginning of End of Cycle Cycle og Moderator Coefficient (an/oF) x 104 + 0.50 - 3.50 aD Doppler Coefficient (Ap/0F) x 105 - 0.87 - 2.11 op Pressure Coefficient (ap/ psia) x 106 - 1.00 + 7.00 aB Boron Worth Coefficient (op/ ppm) x 104 - 0.80 - 1.00 -
Beff Delayed hutron Fraction, % .75 .45 aCRC Net
- Rod Worth (%op) - 2.90 - 2.90
- Total rod worth minus stuck rod worth
13 XN-NF-84-18 tunsin( f amof fLt tutt S-ir(nu NN s . _ s- a. ..
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w at:0 Davial' atg osetas _,,,,,
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Figure 2.5 Limiting Axial Shape for MDNBR Calculations at 50% Power c.__.__.._..._.._.._.... __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
18 XN-NF-84-18 3.0 TRANSIENT ANALYSIS The transients analyzed for Palisades are categorized as either Anti-
[-
cipated Operation Occurrences (A00s) or Postulated Accidents (pas). The A00s are further categorized as either requiring only the action of the reactor
{
protection system (RPS) to meet the Specified Acceptable Fuel Design Limits
[ (SAFDLs) or those requiring RPS action and/or observance of the Limiting Conditions of Operation (LCO).
Table 3.1 lists the transient events considered and summarizes the disposition of each transient. The boron dilution event was not analyzed since, as a reactivity insertion transient at power, it is bounded by the CEA withdrawal transient. The other transient not reanalyzed was the excess-feedwater flow transient since it produced a cooldown rate less severe than that produced by the excess-load transient. The loss of-A.C.-power event was not simulated as a DNBR transient since it is bounded by the four pump coastdown. The steam tube rupture was not reanalyzed since the TM/LP still protects against fuel damage and the radiation release is determined by the operating limits on primary and secondary activity levels.
The steam line rupture was not treated in this analysis since it is a secondary side-induced transient and would be no worse, from the point of view of a return to power, than the prior analysis. The two factors which would mitigate the event are: 1) reduced heat transfer area from primary to secondary which slows the cooldown, and 2) decreased primary loop flow which l slows the transient and increases the core inlet boron concentration.
(
19 XN-NF-84-18 3.1 ANTICIPATED OPERATIONAL OCCURRENCES REQUIRING ONLY RPS ACTION The transients analyzed which fall into this category are: the loss-of-load transient, the excess-load transient, and the RCS-depressurization transient.
3.1.1 Loss of Load Event This event was analyzed to simulate plant performance upon a turbine trip without a direct reactor trip. The abrupt loss-of-heat sink results in a rapid rise in the reactor coolant system (RCS) temperature and an expansion of the coolant which produces an insurge of water into the pressurizer and, ultimately, an increase in pressurizer pressure. The criterion employed is that the peak transient pressure must not exceed the ASME code limit of 110% of design pressure (i.e., 2750 psia). The SAFDLs were not approached in this transient since power was appreciably less than that required to reach 21 kW/ft and the MDNBR occurred at the start of the event.
The transient was simulated with bounding E0C kinetics. The pressurizer spray was turned off and the effects of the relief valves (PORVs) were also ignored in order to produce as high a pressure as possible during the simulated transient. The steam dump and bypass were also removed from the model for the same reason.
Figures 3.1 to 3.8 show the simulated plant response of the loss-of-load event. As the event was initiated, steam line flow dropped )
dramatically within the first few seconds (Figure 3.1). Shortly thereafter, primary temperatures began to rise rapidly due to the loss of the heat sink.
The rapid expansion of the primary loop inventory caused an insurge into the h pressurizer (Figure 3.3) and the subcooled. water volume in the pressurizer i
a
b
[ 20 XN-NF-84-18
( rose (Figure 3.4), causing a dramatic rise in RCS pressure (Figure 3.5). The pressure rise produced a reactor trip on high pressurizer pressure. The i safety relief valves opened (Figure 3.3) at about 10 seconds and controlled the pressure nearly at the setpoint. The maximum pressure reached was 2500.6 i
( psia.
( Figures 3.6 to 3.8 show the reactivity traces, the power and heat flux, respectively. Table 3.3 summarizes the events during the transient.
3.1.2 Excess Load Event Inadvertent opening of the turbine control valve, steam dump valves and/or the steam bypass valve would result in increased steam flow and increased heat extraction. The resultant cooldown of the RCS would produce a positive reactivity insertion at E0C conditions when a large, negative moderator feedback coefficient exists. Protection against core damage is provided by the high neutron flux trip (VHPT), the low steam generator _
pressure trip, and the TM/LP trip.
{
The pressurizer heaters were assumed to be inoperable to f provide a conservative MONBR calculation. Bounding E0C kinetics parameters were used in the simulation.
The l imiting excess-load transient is the simultaneous opening of steam dump and bypass valves. The plant response to this event was simulated by rapidly ramping in 2 seconds the steam flow to 179% of rated flow.
l Figures 3.9 to 3.17 show'the simulated plant response. As the steamline flow increased, the heat extraction from the primary loop increased and the steam j I
generator exit temperatures began to decrease (Figures 3.9 and 3.10). This !
I I
l
)
1 21 XN-NF-84-18 cooldown transient propagated to the' core and simultaneously caused a contraction of the coolant inventory. The net result is a reduction in the liquid volume in the pressurizer and the RCS pressure (Figures 3.11 and 3.12).
The positive feedback from the moderatcr cooldown, Figure 3.15, produced a slight power ramp, Figure 3.16, which resulted in an increase in core heat flux, Figure 3.17, and a small reversal of the inventory shrinkage (Figure 3.11) before the reactor tripped on the high neutron flux trip.
Table 3.4 is an event summary for this transient and Table 3.5 summarues the input for the XCOBRA-IllC calculation of MDNBR.
3.1.3 RCS Depressurization Event The event simulated was a f ailure of both pressurizer relief valves fully open. The kinetics parameters used in the simulation were bounding BOC values. The pressurizer heater capacity was set to zero to allow a more rapid depressurization.
Figure 3.18 to 3.23 suninarize the transient results for this event. Table 3.6 is an event table for the transient, and Table 3.7 is a listing of the boundary conditions input to the XCOBRA-IIIC calculation.
Figure 3.18 shows the steam flow through the relief valve following the inadvertent opening of the PORV. The RCS begins to depressurize (Figure 3.19) ;
and power increases slightly (Figure 3.20). Core flow drops slightly due to the slight increase in core inlet temperature (Figure 3.22). The reactor )
trips on the TM/LP with ample margin to DNB. It is concluded that the bias in the TM/LP is sufficient to protect the core during this event.
.m .. .
b 22 XN-NF-84-18
{
( 3.2 ANTICIPATED OPERATIONAL OCCURRENCES REQUIRING RPS ACTION AND/0R OB5ERVANCE OF THE LCOs
( The transients discussed in this subsection require observance of the LCOs for DNB and for linear heat rates in order to protect the SAFDLs, and consist of: the loss-of-coolant flow event, the CEA withdrawal event, and the CEA drop event, r 3.2.1 Loss-of-Coolant-Flow Event Flow reductions result in an increase in enthalpy rise across the core and a subsequent increase in coolant temperature in the hot leg of the RCS. The increased local enthalpy and decreased flow result in a reduction of margin to DNB in the core. The most severe transient, a loss of power to all four RCS pumps simultaneously, was evaluated by simulating a coastdown of all four RCS pumps in the PTSPWR2 model and observing the MDNBR for the transient.
Bounding BOC kinetics were used. The pump coastdown curve, Figure 3.24, is a best estimate curve. The flow trip setpoint is set 3% low to provide a conservative MDNBR.
The event sequence for the transient is summarized in Table
( 3.8. Table 3.9 lists the input to XCOBRA-IIIC. Figures 3.24 to 3.30 show the L
simulated plant responses to the four-pump coastdown. The reactor thermal f power increases slightly preceding the scram (Figure 3.25), although core heat flux falls due to the decreased heat transfer to the coolant (Figure 3.26). The core average temperature (TCA in Figure 3.27) shows the rise in average temperature which accompanies the reduced coolant flow in the core.
This increase in core average temperature causes an increase in RCS inventory
{ volume and an insurge to the pressurizer accompanied by a pressure increase
23 XN-NF-84-18 (Figures 3.28 and 3.29). The core inlet ~ temperature remains fairly constant, falling only af ter the cold leg temperature decrease, resulting from the flow decrease in the steam generator, reaches the core (Figure 3.30).
3.2.2 CEA-Withdrawal Event An inadvertent withdrawal of a bank of CEAs introduces positive reactivity which increases both core power and heat flux. Two potentlal initiators of this event are: 1) operator error; and 2) a malfunction of either the CEA drive mechanism or of the drive control system which results in an uncontrolled, continuous withdrawal of a CEA bank. Heat extraction through the steam generator lags behind the power increase and the increased power is converted to heat in the RCS. Protection against violation of some of the SAFDLs is provided by the VHPT, the TM/LP trip, or the high pressure trip.
A spectrum of uncontrolled rod withdrawls was simulated with PTSPWR2 by increasing the reactivity linearly at rates which could be achieved in the reactor.
Initial power for the transient was either 1062.6 MWt or 2125.2 MWt since the most limiting part-power transient was found to occur from 50f. power.(2) The purpose of the simulations was to demonstrate that the f ast rod withdrawals could not produce enough overshoot from scram delays to endanger either of the SAFDLs. Further, the simulation was to demonstrate
]
that the heat =up re'es used in setting the transient bias in the TM/LP were chosen such that the reactor would be tripped by a trip other than the TM/LP and such that the SAFDLs are not endangered during slow rod withdrawal transients.(8) l
\
[
24 XN-NF-84-18 J
(
l
( 3.2.2.1 Rod Withdrawals ~From Full Power The f astest reactivity insertion modeled was a linear ramp at 5x10-46p/sec. This value conservatively bounds the achievable rates.
BOC kinetics were used to produce the greatest overshoot. The results of the I simulation are displayed in Figures 3.31 to 3.37. The fast reactivity
( insertion (Figure 3.31) produced a rapid rise in reactor power (Figure 3.32) which was terminated by the high neutron flux trip. The peak heat flux in the core occurred between 3 and 4 seconds (Figure 3.33). The increased heat flux resulted in an increase in the loop temperatures (Figure 3.34), a rise in pressurizer pressure (Figure 3.35), and a decrease in coolant mass flow (Figure 3.36). Because of transport delays, the rapid increase in core inlet temperature, shown in Figure 3.37, occurred after the minimum DNBR.
Table 3.10 is an event table summarizing the tran-sient. The overshoot to 2670.5 MWt corresponds to a transient 16.1 kW/ft. The transient results are benign in terms of either DNB or lineai heat generation I
limits. Table 3.11 consists of the boundary conditions used in the XCOBRA-(
IIIC calculation.
f A sicw rod withdrawal transient was also run from 100%
power. This transient does not serve as part of the basis for the TM/LP(7) and no verification of transient allowance. is required. The transient was ;
4 simulated using bounding BOC kinetics and a reactivity insertion rate of 2.5x10-5 Ap/second. Figures 3.38 to 3.45 show the transient behavior of f several key system variables during the transient. As reactivity is inserted (Figure 3.38), the reactor power undergoes a nearly linear power ramp up to the high neutron flux trip (Figure 3.39). The care heat flux lags just l
l
25 XN-NF-84-18 slightly (Figure 3.40), and increasing primary loop temperatures (Figure 3.41) force more water into the pressurizer (Figure 3.42), causing an increase in pressurizer pressure (Figure 3.43). The core flow decreases as the density of the water in the cold leg decreases (Figure 3.44) and the core inlet temperature rises (Figure 3.45).
The transient is summarized as an event table in Table 3.12. The input for the DNBR calculation is described in Table 3.13.
As reported in Reference 2 , the MDNBR for BOC kinetics was found to be nearly invariable with reactivity insertion rate. It was further found to rise witn decreasing reactivity insertion rates only as the high pressure trip became active in terminating the transient.
3.2.2.2 Rod Withdrawals From Par,t-Power Rod withdrawal transients from part-power have been performed for operation at 2530 MWt(1,2). Two observations were made: First, the most limiting transients start from 50% power; and second, the worst DNBR results occur at the highest average heatup and power increase rates at which the TM/LP has to function. Since the TM/LP, the VHPT and the high pressurizer pressure trip serve to protect against DNB in this transient and, since the basis for the TM/LP is protecting slow power transients from part-power, it is only necessary to show that the VHPT or the high pressure trip intervene at the )
required heatup and power increase rates.
The fast rod withdrawal from part power is not expected to be a DNBR _ limiting transient since the VHPT will allow only an increase of 15% of rated power before the trip setpoint is reached. A reactivity insertion rate of 5x10-4 op /second coupled with bounding BOC j kinetics gives the fastest power ramp.
1
)
b
[
26 XN-NF-84-18
(
( Figures 3.46 to 3.49 sumarize the transient results for reactivities (Figure 3.46), power (Figure 3.47), core heat flux resulting from the power increase (Figure 3.48), and the primary loop temperatures (Figure 3.49) which increase in response to the increase core heat flux.
Figures 3.50 to 3.52 show the time traces for pressurizer pressure, core flow and core inlet temperature. Table 3.14 is the event table for this transient,
)
and Table 3.15 gives the XCOBRA-IIIC input for the MDNBR calculation.
[ Reference 2 reports a study of the spectrum of
\
l withdrawal rates from part power. While the BOC conditions were found to be more DNBR limiting for reactivity insertion rates greater than 2x10-4 ao/sec-ond, the action of the TM/LP was required in the mid-cycle transients and, had enough reactivity been available, at E0C conditions. Thus, it is necessary to verify that the average primary temperature heatup rate and power rates used in creating the TM/LP will cause the VHPT or high pressure trip function to scram the reactor with an acceptable MDNBR resulting.
A reactivity insertion rate of 5x10-5ao/second using bounding E0C kinetics results in a transient which has a power ramp rate and average temperature ramp rate less than or equal to the values used in creating the TM/LP biases. Figures 3.53 to 3.61 summarize the transient results. Tables 3.16 and 3.17 are event tables and XCOBRA-IIIC input tables, respectively.
[
The rate of net reactivity insertion is quite low for f this transient because of the large negative moderator feedback (Figure 3.53). The resulting power and heat flux rises (Figure 3.54 and 3.55) show a
{
27 XN-NF-84-18 distinctly nonlinear behavior because of this effect. While the actual power does not ramp upward significantly in this transient, the steam generator pressure (Figure 3.56) and the primary loop tcmparatures (Figure 3.57) are increasing nearly linearly thtoughout the transient. This produces a nearly constant insurge of liquid into the pressurizer (Figure 3.58) but, because of the pressurizer sprays, the pressure does not ramp up very rapidly until the gas volume becomes quite small and the power begins to rise because of the pressure effect on the moderator density. Core flow (Figure 3.60) falls throughout the transient as the cold leg heats up and core inlet temperature rises nearly linearly.
This transient trips on the high pressurizer pressure trip with temperature and power ramp rates which do not exceed those used to calculate the TM/LP trip and this simulation has the effect of validating the TM/LP trip basis.
3.2.3 CEA Drop Event A failure in the CEA drive mechanism can result in an inadvertent full-length insertion of a CEA during power operation. . Fixed demand from the turbine would cause a cool-off transient in the RCS and, for negative moderator feedback, a return to the original power with a sig-nificantly greater radial peaking in the core. Since the power initially decreases following the dropping of the CEA, no reactor' trip occurs and protection of the SAFDLs is provided solely by the LCOs.
This event was simulated by introducing a step decrease in total reactivity at steady-state and full power. Bounding E0C kinetics 1
J
. . . . .. . .. . .. .. I
28 XN-NF-84-18
( parameters were used and the reactivity insertion was selected to conserva- l tively bound that due to the most reactive CEA being inserted. A radial peaking factor of 116% was included during the return to power. During the cooldown transient, inlet temperature fell, mass flow rose and pressure,
{
which was not controlled in this transient, fell. The increased radial peaking and reduced pressure tended to decrease the DNBR while the decreased inlet temperature and increased flow tended to increase the DNBR.
( Table 3.18 summarizes the event sequence of the transient.
Table 3.19 is the XCOBRA-IIIC input used to calculate the MDNBR. This transient does not produce an MDNBR below the target value of 1.17 as would be expected since it was used to establish the LC0 on inlet temperature in Reference 7 In the basis, no credit was taken for the cooldown of the core
( inlet flow, hence significant margin to DNB can be expected for this transient.
( The initial decrease in reactivity due to the dropped CEA is offset by the doppler feedback initially as the fuel cools off (Figure 3.62).
As the reactor power and heat flux fall off (Figures 3.63 and 3.64), the
( turbine flow begins to increase (Figure 3.65) to maintain a constant heat extraction from the plant. As the turbine flow increases, the primary loop cools off (Figure 3.66) and the RCS coolant inventory becomes more dense.
This results in an increase moderator feedback and a reduction of the liquid
(
in the pressurizer (Figure 3.67). The pressrizer pressure decreases (Figure 3.68) and the core mass flow increases in response to the cooldown of the primary loop. After 90 seconds, the reactor power has nearly stabilized and the minimum DNBR has already occurred.
29 XN-NF-84-18 3.3 POSTULATED ACCIDENTS The events which fall in this category are assumed to occur infrequently and are not required to meet the SAFDLs. The ultimate criterion applied to these transients is a radiation exposure limit 10 CFR 100. In assessing the safety of operation in Cycle 6, a comparison of expected pin failure with prior cycles is used to judge the acceptability of the fuel performance. Fuel failure is conservatively assumed coincident with the occurrence of DNB. Hence, for the DNBR limiting accident analyzed in this subsection, the seized pump rotor, the expected number of fuel pins undergoing DNB was used as the evaluation criterion. In addition, the loss of normal feedwater transient was analyzed for long-term decay heat removal assuming the coincident failure of the au,tomatic feedwater valve to the least plugged steam generator. Because of the decreased heat transfer area, decay heat removal at natural circulation flows can lead to the pressurizer filling due to RCS heatup.
3.3.1 Primary-Pump-Seizure Event The instantaneous loss of pumping power caused by dis-integration of the pump impeller or a complete seizure of the pump shaft would result in a rapid flow decrease through the affected cold leg, and would cause a reactor trip due to low flow in that loop. The flow reduction rate would be more drastic than in a total loss of pumping power and would create a more rapid approach to DNB. Bounding B0C kinetics parameters were used to maximize the power excursion and delay the shutdown of the power following the trip.
J
30 XN-NF-84-18
[
The transient was simulated by stopping one of the four
{
pumps at full-power operation. Pressurizer pressure control was retained so
( that the spray would decrease the pressure transient. The results of the simulation are shown in Figures 3.70 to 3.77. The event sequence is
( sumarized in Table 3.20. XCOBRA-IIIC input is in Table 3.21.
The instantaneous loss of pumping power to a single cold leg caused an immediate flow reversal in that cold leg. As the flow rose in the
( intact cold legs, due to the reduction in core flow for the intact loop and to the flow reversal in the seized loop for the intact leg in the seized loop, a near steady-state flow was achieved within 1 second. The new flow was 78.2%
of the original flow and resulted from an 8% increase in flow in the intact loop, a reverse flow of about 30% in the seized locp, and an increased flow of about 153% in the intact leg of the seized loop. Figure 3.70 shows the core inlet flow transient.
Power rises slightly before the reactor trip (Figure 3.71);
however, heat flux never rises as high as the initial value (Figure 3.72).
The core inlet temperature is nearly constant for the first three seconds (Figure 3.73) before rising rapidly. Over the first 2-3 seconds, the only significant temperature changes are in the clad temperature and the core average temperature (Figure 3.74). The rise in core average temperature is reflected by the pressurizer surge flow (Figure 3.75). The pressurizer water volume continues to increase after the scram (Figure 3.76) as does the pressurizer pressure.
3.3.2 Loss of Feedwater with a Loss of Offsite Power
[ Operation of Palisades with 50% of the steam generator tubes plugged reduces the heat transfer area available to reject decay heat to the
(
31 XN-NF-84-18 steam generators. Should the primary coolant become sufficiently hot, volumetric swell of the RCS inventory can potentially fill the pressurizer and force coolant out the safety and relief valves. A significant amount of inventory loss in the RCS or an inability to protect the core and achieve a cooldown without pumping power remains a possibility for a reactor with' reduced heat transfer mechanisms. The reduced initial power tends to offset this effect by lowering the decay heat load on the system.
The analysis was performed for a limiting case, a loss of normal feedwater with loss of offsite power. This transient causes a loss of normal feedwater to both steam generators, a turbine trip, and a loss of offsite power. In addition, it was assumed that a valve failure results in auxiliary feedwater not being available for cooling to the least plugged steam generator. The simulation of the transient was performed using SLOTRAX with an asymmetric loop model which had 60/40 plugging in its steam generators.
The event was initiated from full power by ramping the normal feedwater to both steam generators to zero in one second. Table 3.22 is an event table for the transient. Simultaneous with the loss of normal feedwater, the reactor was tripped and the primary coolant pumps were allowed to coast down. Auxiliary feedwater was not introduced until 1 minute after )
main feedwater pumps stopped to allow for time necessary to start the motor.
on the auxiliary feedwater pump. Upon initiation of auxiliary feedwater, a valve failure results in the total output- from the motor-driven auxilicry feedwater pump being introduced into the 60% plugged steam generator. The plant was then allowed to recover passively without the benefit of letdown flows in the RCS. ,
32 XN-NF-84-18
[
Figures 3.78 to 3.82 sumarize the transient results. The
{
reactor thermal power (Figure 3.78) is predominantly decay heat af ter the
( first few seconds. The loop flow rates (Figure 3.79) show the asymmetric flow j i
behavior of the two loops with their different plugging levels. The steam
( generator liquid level for the intact steam generator reflects the fact that it is isolated and floods up to a level at which it is controlled by dumping steam either to the atmosphere or to the condenser, automatically. The
( affected steam generator dries out in 2975 seconds (Figure 3.80).
A key system variable is the average temperature of the RCS.
It should reach an early peak and decrease with time which provides decay heat removal via natural circulation. The peak value determines the amount of expansion of the primary loop coolant inventory and thus the change in liquid
( level of the pressurizer. Figure 3.81 shows the RCS temperature as a function of time. Figure 3.82 shows the liquid level in the pressurizer.
The results of the simulation (Figures 3.81 and 3.82) demonstrate that the pressurizer does not fill and that decay heat removal via natural circulation is established.
(
l
(
l I
33 XN-NF-84-18 J
Table 3.1 Transient Events )
)
Transient Disposition
]
A00s Requiring Only RPS Action Boron Dilution Not Analyzed Loss of Load Analyzed Loss of Feedwater Not Analyzed
~
Excess Load Analyzed Excess Feedwater Not Analyzed RCS Depressurization Analyzed A00s Requiring RPS Action and/or LC0 Loss of Coolant Flow Analyzed )
Loss of A.C. Power Not Analyzed CEA Withdrawal Analyzed CEA Drop Analyzed pas Seized Rotor Analyzed Steam Line Rupture .Not Analyzed Steam Generator Tube Rupture Not Analyzed
]
Loss of Feedwater with a loss of Offsite Power Analyzed -1 mme M
L 24 XN-NF-84-18
[
Table 3.2 Index of Symbols
{
Symbol Description Units CFWPR Volume of water in the pressurizer ft3 DK Net reactivity $
{
DKD0P Doppler feedback $
( DKMAN Manual reactivity inserted S DKMOD Moderater pressure feedback $
LEVPR Pressurizer liquid level ft LEVSG1 Downcomer liquid level in steam ft generator #1 LNB1 Subcooled level in steam generator #1 ft
( PL Reactor power MWt PPR Pressurizer pressure PSIA PSGI Pressure in steam generator #1 PSIA GDA Core heat flux Btu /hr-ft2 GPR Pressurizer heater power kWt GT Total power extracted from the steam Btu /sec generators TAVG1 Average temperature in Loop 1 0F TCA Core average temperature OF
( TCIO Core inlet temperature OF TCLAD Average clad temperature- 0F-(
TCL1 Cold leg temperature in Loop 1. OF
( THL1 Hot leg temperature in Loop 1- 0F TLPI Reactor vessel lower plenum inlet OF
[ ' temperature TSG1PI -Inlet temperature to steam generator #1 0F h i i
_ _ _ _ _ _ . J
35 XN-NF-84-18 )
J Table 3.2 Index of Symbols (Cont.)
]
Symbol -
Description Units TSGlP0 Outlet temperature from steam generator #1 of WOOSLT Flow from dome to steamline Ib/sec
]
WFWT Total feedwater flow lb/sec WPRRV Pressruizer relief valve flow lb/sec )
WPRSU Pressurizer safety valve flow lb/sec WRV1 Flow from the relief valve in steamline #1 lb/sec WSV1 Flow from the safety valve in steamline #1 lb/sec WTB Flow through turbine Ib/sec WUPPR Surge flow to pressurizer lb/sec )
]
1
)
)
1 i
E 36 XN-NF-84-18 L
Table 3.3 Event Table For The Loss of Electric Load
[
Time Event Value or Setpoint
[ 0 Turbine flow reduce to zero 4.62 Peak power 2259.8 MWt 5.87 Peak core heat flux 142443 BTU /hr ft3
[ 8.76 Reactor trip on high 2277 psia pressurizer pressure 10.34 Pressurizer safety valve 2500 psia
[ opened 10.77 Peak core average temperature 573.3 0F 10.90 Peak pressurizer pressure 2500.6 psia l
[
{
l l
(
{
l
37 XN-NF-84-18
)
)
Table 3.4 Event Table For Excess Load
]
Time EVENT Value or Setpoint ,
O Begin ramping turbine flow 2.0 Maximum turbine flow 179% of rated 8.49 High neutron flux trip 2380.2 MWt 9.05 Peak power level 2504 MWt
]
9.66 MDNBR 1.782 9.77 Peak core heat flux 148,709 8tu/hr-ft2
]
)
}
)
J
F 38 XN-NF-84-18
( Table 3.5 XCOBRA-IIIC Input For Excess Load VARIABLE VALUE UNITS Pressure 1918.4 psia Inlet temperature 536.38 0F
( Core flow 25,805 lb/sec Power 2259.3 MWt
[
[
[
{
l s
1
39 XN-NF-84-18 Table 3.6 Event Table For The PORV Failure )
TIME ,
EVENT VALUE OR SETPOINT (Seconds) ]
N 0 -Pressurizer relief va lve 1 opens J 31.63 Reactor trip on TM/LP 1722.6 psia
)
- s
, Peak power level ,
's 2218 MWt 32.01 Peafcorekeatflux
' 144,451 Btu /hr-ft2 1 s
j Peak core average 565 F 32.31 s temperature s
1.636
)
32.32 Minimum DNBR .
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L 40 XN-NF-84-18 ,
l L
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Table 3.7 XCOBRA-IIIC Input for PORV Failure VARIABLE VALUE UNITS Pressure 1733.6 psia Inlet temperature 541.7 0F Core flow 25.413 lbs/sec Power 2238.6 MWt i
(
)
s
41 XN-NF-84-18 Table 3.8 Event Table For 'The Four Pump Coastdown TIME (Seconds) EVENT VALUE OR SETPOINT 0 ,
Pump trip Peak core heat flux 139,666 Btu /hr-ft2 1.28 Reactor trip on low flow 93%
1.56 Peak Reactor Power 2160.3 MWt 2.81 Minimum DNBR 1.579 3.32 Peak core average 5690F temperature 5.07 Peak pressurizer pressure 2008 psia
[
{
l j
f
L 1
42 XN-NF-84-18
. Table 3.9 XCOBRA-IIIC Input For Four Pump Coastdown v
VARIABLE VALUE UNITS Pressure 19956.5 psia Inlet temperature 542.7 0F Core flow 20,547 lb/sec Power 2108 MWt I
i l
b
_ _ _ _ _ _ _ _ _ _ )
43 XN-NF-84-18 Table 3.10 Esent Table For Fast Rod Withdrawal From 100% Power TIME (Seconds) EVENT VALUE OR SETPOINT 0 Maximum rod withdrawal 5x10-4 ap/recond rate initiated 1.89 Reactor trip in high neutron 2380 MWt flux 2.50 Peak power level 2670.5 MWt 3.36 Minimum DNBR 1.679 3.42 Peak core heat flux 150,095 Btu /hr-ft2 3.58 Peak c. ore average 5650F temperature a
5.32 Peak pressurizer pressure 2008 psia 4
L b
44 XN-NF-84-18 L
( Table 3.11 XCOBRA-IIIC Input for Fast Rod Withdrawal From 100% Power v
~
VARIABLE VALUE UNITS Pressure 1960 psia TINLET 542.5 0F Core flow 25,390 lb/sec j power 2280 MWt
45 XN-NF-84-18 rable 3.12 Event Table For The Slow Rod Withdrawal At 100% Power o
TIME (Second) EVENT VALUE OR SETP0 INT 0 Reactivity insertion begins 2.5 x 10-5 AP/sec.
30.00 Reactor trip on high 2380 MWt neutron flux 30.53 Minimum DNBR 1.674 30.61 Peak core heat flux 152,906 BTU /hr-ft2 ,
31.27 Peak core average 568 0F temperature 33.27 Peak pressurizer pressure 2105 PSIA s
L 46 XN-NF-84-18 L
p Table 3.13 XCOBRA-IIIC Input For Slow Rod Withdrawal From 100% Power VARIABLE VALUE UNITS Pressure 2062 PSIA Inlet temperature 545.2 0F Core flow 25,329 lb/sec Power 2320 MWT i
(
A
J 47 XN-NF-84-18 Table 3.14 Event Table For Fast Rod Withdrawal From 50% Power TIME (Seconds) EVENT VALUE OR SETPOINT 0 Reactivity insertion begins 5 x 10-4 ap/sec 3.16 Variable high power trip 1392 MWt i
3.79 Peak power. level 1630.9 MWt 4.79 Minimum DNBR 2.007 4.80 Peak core heat flux 83,843 Btu /hr-ft2 4.97 Peak core average 556 0F temperature 5.83 Peak pressurizer pressure 199) PSIA ,
j
]
4 a'
L 48 s XN-NF-84-18 e
s I
Table 3.15 XCOBRA-IIIC Input For Fast Rod Withdrawal From 50% Power VARIABLE VALUE UNITS Pressure 1965.3 PSIA I
Inlet temperature 545.8 0F Core flow 25,387 lb/sec Pcwer 1276 MWt i
49 XN-NF-84-18 Table 3.16 Event Table for Slow. Rod Withdrawal from 50% Power Time Value or (seconds) Event Setpoint 0 Reactivity insertion begins 5x10-5 ap/sec. 216.27 Minimum DNBR 1.664 234.11 Reactor trip on high pressurizer 2277 psia pressure Peak power level 1285.6 MWt 234.68 Peak core heat flux 82,584 Btu /hr-ft2 235.31 Peak core average temperature 5870F 235.52 Steam line safety valves opened 1000 psia 235.55 Peak steam dome pressure 1002 psia 236.32 Peak pressurizer pressure 2312 psia i
i
/
50 XN-NF-84-18 f Table 3.17 XCOBRA-IIIC Input for Slow Rod Withdrawal from 50% Power Variable Value Units Pressure 2165.6 PSIA Inlet Temperature 580 0F Core Flow 25,354 lb/second Power 1395 MWt
( ,
51 XN-NF-84-18 -
Table 3.18 Event Table for CEA Drop Time Value or (Seconds) Event Setpoint 0 CEA Drop -2x10-3 Ap/sec.
2.13 Minimum Power 1796 MWt 56.19 Peak power level 2199 MWt 63.52 Peak core heat flux 144,344 Btu /hr-ft2 71.44 Minimum DNBR 1.372 I
)
J l
t
I 52 XN-NF-84-18 s
Table 3.19 XCOBRA-IIIQ Input for CEA Drop o
Variable Value Units Pressure 1929.7 . PSIA Inlet temperature 534.5 0F Core flow 25,504 lb/sec.
', Power 2,237 MWt Radial peaking
- 1.9595
(
l
)
k
- This peaking represents 116% of the Technical Specification limiting on L radial peaking for an interior channel.
l
. . ~ . . . . . . .
53 XN-NF-84-18 Table 3.20 Event Table for the Locked Rotor Time Value or (Seconds) Event Setpoint O Pump seizure -
0.48 Reactor trip on low flow 93%
0.80 Peak power level 2198 MWt 1.38 Minimum DNBR 1.523 1.77 Peak core temperature 5700F 5.02 Peak pressurizer pressure 2053 psia
]
br a _ _ .._ _ _
l 54 XN-NF-84-18 L
Table 3.21 XCOBRA-IIIC Input for the Locked Rotor Variable Value Units Pressure 1964.5 PSIA Inlet temperature 542.5 0F Core flow 20,263 lb/sec.
Power 2133 MWt v
4 J
A
i 55 XN-NF-84-18 4
Table 3.22 Event Table for loss of Feedwater with Loss of Offsite Power i
i l
Time Value or (Seconds) Event Setpoint 0 Reactor trip; primary and main feed- -
water pumps coastdown 25.0 Maximum steam generator pressure 1009.0 psia (intact) 1012.9 psia (isolated) <
60.0 Auxiliary feedwater initiated to 70.8 lbm/sec.
intact steam generator 2800.0 Maximum pressurizer pressure 1983.3 psia 2875.0 Maximum pressurizer water volume 1332.8 ft3 2975.0 Isolated steam generator dries out -
i 4
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( 138 XN-NF-84-18 4.0 DISCUSSION The ENC transient analysis performed for the Palisades nuclear power
( plant demonstrates adequate margin to fuel and vessel design limits for Cycle 6 under normal operation, anticipated transients, and postulated accidents.
[ The transients analyzed in Section 3 were selected because they were shown in the prior analyses (1,2) to have less margin than the transients not analyzed.
{
The loss-of-load event was analyzed as an overpressurization transient
( and, as such, bounds events such as the loss-of-feedwater or a loss-of-heat-sink in one steam generator. The action of-the pressurizer safety valve in controlling the overpressurization is sufficient to demonstrate the accept-ability of the plant for overpressurization transients. The loss-of-feedwater in conjunction with a loss of A.C. power was analyzed as a long term
( cooldown event because of the reduced heat transfer area.
The excess-load event was analyzed as the limiting cooldown A00. The action of the variable high power trip in terminating the transient without a significant degradation in DNBR was sufficient to bound the results of an
[
excess-feedwater transient.
( The RCS depressurization transient represents the most pressure tran-sient in the A00 category and was used to test the TM/LP bias. As a test of the TM/LP bias, it was found to be less limiting than the CEA-withdrawal event.
The loss-of-coolant flow event is a limiting A00 for flow reduction and bounds the loss of A.C. power as a DNBR transient. Further, it provided one
{
of the two transtents which was analyzed to set the LC0 for DNB.
4
b 139 XN-NF-84-18 i
(
5.0 REFERENCES
- 1. " Plant Transient Analysis of the Palisades Reactor for Operation at 2530 MWt", XN-NF-77-18, Exxon Nuclear Co., Inc., Richland, Washington, July
{ 1977.
- 2. " Rod Withdrawal Transient Reanalysis for the Palisades Reactor", XN-NF-
[ 83-57, Exxon Nuclear Co., Inc., Richland, Washington, February 1983.
- 3. " Description of the Exxon Nuclear Plant Transient Simulation Model for
( Pressurized Water Reactors (PTSPWR)", XN-74-5(P), Rev. 2 and Rev. 2 Supplement 1, Exxon Nuclear Co., Inc., Richland, Washington, October 1983.
- 4. "XCOBRA-IIIC: A Computer Code to Determine the Distribution of Coolant During Steady-State and Transient Core Operation", XN-NF-75-21(P), Rev.
2, Exxon Nuclear.Co., Inc., Richland, Washington, September 1982.
1, Exxon Nuclear Co., INC., Richland, Washington, April 1982. i
[ 6. "H.B. Robinson Loss of Feedwater Transient at 1955 MWt Model Description and Results", XN-NF-82-91(P), Exxon Nuclear Co., Inc., Richland, Washington, November 1982.
{
- 7. " Palisades Cycle 6 Setpoint Document", XN-NF-84-14, Exxon Nuclear Co.,
Inc., Richland, Washington, March 1984.
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L XN-NF-84-18 Issue Date: 3/9/84 PLANT TRANSIENT ANALYSIS FOR PALISADES NUCLEAR POWER PLANT WITH 50% STEAM GENERATOR PLUGGING
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Distribution
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F. T. Adams J. C. Chandler
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R. A. Copeland J. S. Holm W. V. Kayser T. R. Lindquist
( W. T. Nutt G. A. Sofer
( R. 8. Stout (Information Only)
G. N. Ward
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CPCo/H. G. Shaw (10)
Document Control (5)
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