ML20195H030
| ML20195H030 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 06/13/1988 |
| From: | Odell L SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML18053A418 | List: |
| References | |
| XN-NF-86-91(NP), NUDOCS 8806280206 | |
| Download: ML20195H030 (42) | |
Text
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XN-NF 91 ( N P )
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1 h l ADVANCEDNUCLEARFUELSCORPORATION
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LOW FLOW TRIP SETPOINT AND THERM AL M ARGIN AN ALYSIS FOR THREE PRIM ARY COOL ANT PUMP OPER ATION OF THE P ALIS ADES RE ACTOR l
i JUNE 1988 l
R A
K 55 1
i ADVANCEDNUCLEARFUELS CORPORATION XN-NF-86-91(NP)
Issue Date:
6/13/88 LOW FLOW TRIP SETP0 INT AND THERMAL MARGIN ANALYSIS FOR THREE PRIMARY C0OLANT PUMP OPERATION OF THE PALISADES REACTOR 4
Prepared by 7
/
/
I L. D.~0' Dell, Team Leader l
PWR Safety Analysis l
Licensing & Safety Engineering Fuel Engineering & Technical Services l
l l
June 1988 9f
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J NUCLEAR REGULATORY COMMISSION REPORT DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT i
PLEASE READ CAREFULLY This technical report was derived through research and development pro-grams sponsored by Advanced Nuclear Fuels Corporation it is being submit-ted by Advanced Nuclear Fuels Corporation to the U.S. Nuclear Regulatory Commission as part of a technical contribution to facilitate safety analyses by licensees of the U.S. Nuclear Regulatory Commission which utilize Ad-venced Nuclear Fuets Corporation fabricated reload fuel or other technical services provided by Advanced Nuclear Fuels Corporation for light water power reactors and it is true and correct to the best of Advanced Nuclear Fuels Corporation's knowledge, information, and belief. The information con-tained herein may be used by the U.S. Nuclear Regulatory Commission in its review of this report, and under the terms of the respective agreements, by licensees or applicants before the U.S. Nuclear Regulatory Commission which are customers of Advanced Nuclear Fuels Corporation in their demonstration of compilance with the U.S. Nuclear Regulatory Commission's regulations.
Advanced Nuclear Fuels Corporation's warrantles a..d representations con-coming the subject matter of this document are those set forth in the agree-ment between Advanced Nuclear Fuels Corporation and the customer to which this document is issued. Accordingly, except as otherwise expressly provided in such agreement, neither Advanced Nuclear Fuels Corporation not any person acting on its behalf; A. Makes any warranty, or representation, express or im.
plied, with respect to the accuracy, completeness, or usefulness of the information contained in this docu-ment, of that the use of any information, apparatus, method, or process disclosed in this document will not j
infringe privately owned rights, or 1
S. Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, ao-parstus, method. or process disclosed in this document.
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XN-NF-86-91(NP)
IABLE OF CONTENTS Section Eagg
1.0 INTRODUCTION
1 2.0
SUMMARY
2 I
3.0 LOSS OF COOLANT FLOW ANALYSIS 4
3.1 Event Initiators........................
4 3.2 Event Consequences.......................
4 3.3 Reactor Protection.......................
5 3.4 Transient Analysis and Thermal Margin Analysis.........
5 3.4.1 Transient Analysis.......................
5 3.4.2 Thermal Margin Analysis 7
3.5 Conclusion...........................
7 4.0 REACTOR COOLANT PUMP ROTOR SEIZURE / SHAFT BREAK ANALYSIS 21 4.1 Event Initiators........................
21 4.2 Event Consequences....................... 21 4.3 Reactor Protection.......................
21 4.4 Transient Analysis and Thermal Margin Analysis.........
21 4.4.1 Transient Analysis.......................
21 4.4.2 Thermal Margin Analysis 22 4.5 Conclusion...........................
22 5.0 CONTROL R00 MISOPERATION....................
24 5.1 Oropped Control Rod or Control Rod Bank 24 5.1.1 Event Initiators........................
24 5.1.2 Event Consequences.......................
24
l ii
.XN NF-86-91(NP)
TABLE OF CONTENTS (Cont.)
Section E191 5.1.3 Reactor Protection.......................
24.
5.1.4 Transient Analysis and Thermal Margin Analysis.........
24 5.2 Statically Misaligned Control Rod or Control Rod Bank 25 S.2.1 Event Initiators........................
25 5.2.2 Event Consequences.......................
25 5.2.3 Reactor Protection.......................
26 5.2.4 Transient Analysis and Thermal Margin Analysis.........
26 28 5.3 Single Control Rod Withdrawal 5.3.1 Event Initiators....................,....
28 5.3.2 Event Consequences.......................
28 5.3.3 Reactor Protection.......................
28 5.3.4 Transient Analysis and Thermal Margin Analysis.........
28 6.0 PTSPWR2 MODEL FOR CE 2X4 LOOP PLANTS..............
32
7.0 REFERENCES
34 0
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111 XN-NF-86-91(NP)
LIST OF TABLES Table Eagg 3.4.1-1 Summary of Initial Operating Conditions (39% Power - Three Primary Coolant Pumps Operating)
Loss of Forced Reactor Coolant Flow 9
3.4.1-2 Sequence of Events Summary (39% Power - Three Primary Coolant Pumps Operating)
Loss of Forced Reactor Coolant Flow 10 3.4.2-1 Summary of Input Parameters for the XCOBRA-IIIC Thermal Margin Analysis for the Loss of Forced Reactor Coolant Flow......................
11 4.4.2-1 Summary of Input Parameters for the XCOBRA-IIIC Thermal Margin Analysis for the Reactor Coolant Pump Rotor Seizure / Shaft Break Analysis.........................
23 5.4.4-1 Summary of Input Parameters for the XCOBRA-IIIC Thermal Margin Analysis for the Single Control Rod Withdrawal Event......................
30 o
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iv XN NF-86-91(NP)
LIST OF FIGURES Fiaure Eggg 3.4.1-1 Loss of Forced Reactor Coolant Flow - Reactor Power 12 vs Time 3.4.1-2 Loss of Forced Reactor Coolant Flow - Reactor Heat Flux vs Time..........................
13 3.4.1-3 Loss of Forced Reactor Coolant Flow -
Reactor Coolant Temperatures vs Time..............
14 3.4.1-4 Loss of Forced Reactor Coolant Flow - Core Flow vs Time 15 3.4.1-5 Loss of Forced Reactor Coolant Flow -
Steam Generator Pressure vs Time................
16 3.4.1-6 Loss of Forced Reactor Coolant Flow -
Idle Pump Loop.........................
17 3.4.1-7 Loss of Forced Reactor Coolant Flow - Parallel Loop 18 to Idle Pump Loop 3.4.1-8 Loss of Forced Reactor Coolant Flow - Pump on loop 3......
19 3.4.1-9 Loss of Forced Reactor Coolant Flow - Pump on Loop 4......
20 l
5.2.2-1 Power Dependent Insertion Limits for two,- three and four primary coolant pump operation.................
31 l
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1 XN-NF-86-91(NP) l l
1.0 INTRODUCTION
l The purpose of this report is to document the analyses performed to support a Technical Specification revision to the current low flow trip setpoint value of 71%(I) of four primary coolant pump flow for three primary coolant pump operation.
The revised low flow trip setpoint value is 60% of the four primary coolant pump flow for operation with only three primary coolant pumps.
Limiting thermal ma. gin transients directly protected by the low flow trip setpoint are the Loss of Forced Reactor Coolant Flow and the Reactor Coolant Pump Rotor Seizure events.(5)
The spectrum of Control Rod Misoperation and Uncontrolled Rod / Bank Withdrawal at Suberitical or Low Power events were identified during previous analyses (5) as the limiting class of non-flow related thermal margin transients. These events are all considered herein for three primary coolant pump operation at the reduced low flow trip setpoint.
l The revised low flow trip setpoint will allow an orderly plant shutdown, without violation of the low flow trip setpoint, should problems develop with one of the primary coolant pumps.
This revision is required since increased steam generator tube plugging has increased the potential for reverse flow through the loop containing the idle /out of service pump.
. As such, the current Technical Specification limit for the low flow trip prohibits three primary coolant pump operation when accounting for measu. ant uncertainties.
I
i l
2 XN-NF-86-91(NP) 2.0
SUMMARY
This report documents analyses which support a Technical Specification revision for Three Primary Coolant Pump Operation from the current low flow III trip setpoint value of 71% of the four primary coolant pump flow rate to 60% of the four primary coolant pump flow rate.
The Technical Specification (2.3.4) allows three primary coolant pump operation for a limited period of time (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) for repair / pump restart at a limit of 39% of rated thermal power.
Upon failure to repair or restart the pump, the plant is required to be in hot standby (or below) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
This reduction in power level to a maximum of 39% will provide sufficient thermal margin to offset that lost by the reduction in primary coolant flow due to the loss of one primary coolant pump.
Thus, the combined effects of the power and primary coolant flow reduction will result in an increase in the initial thermaT margin, which will in turn reduce the consequences of the limiting, thermal margin transients.
All of the limiting thermal margin flow induced transients were shown to be adequately protected utilizing a low flow trip setpoint of 60% of the four i
primary coolant pump flow rate, in conjunction with a flow uncertainty of 5%.
That is, a reactor trip was assumed to occur at 55% of rated flow.
l The non-flow induced thermal margin limiting transients were identified in i
Reference 5 to be the Uncontrolled Control Bank Withdrawal at subcritical or i
low power and the various Control Rod Hisoperation events.
The Uncontrolled Control Rod Withdrawal event was analyzed in Reference 5 for three primary I
coolant pump operation.
Since this transient does not trip on low flow, the i
results of the analysis are not altered by the change in the low flow setpoint and, consequently, remain bounding.
The Control Rod Misoperation events are l
also shown in the following sections to satisfy all applicable acceptance criteria.
l In summary, this Low Flow Trip Setpoint Technical Specification revision is
i 3
XN-NF-86-91(NP) j required since increased steam generator tube plugging has increased the potential fcr reverse flow through the loop containing the tripped /out of service pump, to the extent that the current Technical Specification limit for the low flow trip prohibits three primary coolant pump operation. The revised low flow trip setpoint (60% of four primary coolant pump flow rate) allows for an orderly plant shutdown without violation of the low flow trip setpoint, i' problems develop with one of the primary coolant pumps.
1
4 XN-NF-86-91(NP) 3.0 LOSS OF COOLANT FLOW ANALYSIS 3.1 Event Initiators A mechanical or electrical fault in one of the primary coolant pumps is postulated such that it is necessary to remove this pump from service.
Infrequent operation with three pumps is permitted by Technical SpecEications to provide a limited time for repair / pump restart, to provide for an orderly shutdown, or to provide for the conduct of reactor internals 2
noise monitoring test measurements. The maximum power level allowed for three primary coolant pump operation is 39% of rated power (Technical Specification 2.3.4).
As a result of the increased flow to power ratio relative to full power four primary coolant pump operation, loss of flow from the three prinary coolant pump configuration is in general less severe than the loss of flow from a four primary coolant pump configuration.
The normal power supplies for the primcry coolant pumps are from two buses connected to the turbine generator.
Two primary coolant pumps, in opposite loops, are powered from each bus.
If there is a turbine generator trip, the primary coolant pumps are automatically transferred to a bus supplied from the external power lines.
A turbine generator trip with the failure of this transfer could result in a loss of power durin5 three primary coolant pump operation.
For the partial loss of flow situations (one or two primary coolant pumps coasting down), the magnitude of the coastdown is less severe than the three primary coolant pump coastdown, and the consequences of a partial loss of flow are bounded by the three primary coolant pump loss of flow event.
Additionally for the partial loss of flow situations, there is always some degree of forecd reactor coolant flow.
3.2 Event Conseauences The immediate result of the postulated loss of forced coolant flow is an increase in the coolant temperature as it flows through the reactor core.
i 5
XN-NF-86-91(NP)
This ' temperature increase could challenge the specified acceptable fuel design limits.
Departure from nucleate boiling (DNB) could result with subsequent fuel damage if the reactor is not tripped.
3.3 Reactor Protection Reactor protection is provided by the low reactor coolant flow, thermal margin / low pressure, and high pressurizer pressure trips.
3.4 Transient Analysis and Thermal Marain Analysis 3.4.1 Transient Analysis - The transient analysis was performed using 3 modified version of the PTSPWR2 code.(2)
This version, which is discussed in Section 6.0, simulates two reactor coolant loops with two separate cold legs in each loop in order to account for the reverse flow in the cold leg with the idle /out of service primary coolant pump.
Table 3.4.1-1 presents a list of pertinent analysis input parameters.
Realistic operating conditions and the conservative values used in the analysis are presented for comparison I
purposes.
The nominal power value in Table 3.4.1-1 corresponds to 39% of l
rated thermal power and is the maximum allowed power during three pump 1
operation supported by this analysis.
The actual power used in the PTSPWR code includes a 2% uncertainty as shown in the table.
In conjunction with the I
39% power level, it was assumed that the Variable High Power Trip (VHPT) would be at a power level 10% higher than the operating power and has a 5.5% power i
measurement uncertainty.
This is consistent with the analysis performed in Reference 5 and results in a VHPT at 54.5% power, including the uncertainty.
j l
The inlet temperature shown in Table 3.4.1-1 was obtained by assuming a linear relationship between the HZP average coolant temperature of 532*F and the rated thermal power average coolant temperature of 570.58'F derived from the inlet temperature Limiting Condition of Operation (LCO).(5)
This inlet temperature and the three pump flow rate was then used in the i
6 XN-NF-86 91(NP)
PTSPWR2 code to balance the primary and secondary systems to obtain a steady state condition from which to initiate the transient.
In order to obtain this steady state, a power split between the two steam generators was assumed proportional to the flow rate to each steam generator.
The steady state balance between the primary and secondary systems was then found such that both steam generators produced the same pressure to the turbine.
The axial shape used in the analysis was chosen based on the axial shape index (ASI) limits of the inlet temperature LC0.(5) These limits are based on power and indicate an ASI limit for low power operation of
.342.
The limiting axial shape corresponding to this ASI limit was used in all PTSPWR2 and XCOBRA-IIIC calculatioi ;.
The transient is initiated by assuming that the three operating primary k
coolant pumps begin to coast down.
As the pumps coast down, the core flow is reduced, causing a reactor scram signal initiation at 3.5 seconds on low reactor coolant flow (55% of rated).
As the flow coasts down, primary coolant temperatures increase slightly (core average coolant temperature increases less than 3*F).
This increase in temperature would cause a subsequent increase in pwer, if the moderator temperature coefficient were assumed to be positive.
This effect was not simulated, but was accounted for in the thermal margin analysis by assuming an additional 5% uncertainty in the initial power level.
For full power initial conditions with a four primary coolant pump coastdown, the average core temperature increases approximately 3*F, which results in a power increase of approximately 2.0% of raced.(5)
Therefore, the above assumed 5% conservatively accounts for the power increase that would be predicted for this analysis.
The temperature increase causes an insurge into the pressurizer and resultant increase in the reactor coolant system pressure.
Credit was not taken for this pressure increase in the thermal margin analysis.
Table 3.4.1-2 gives a summary of the sequence of events for the Loss of Forced Reactor Coolant Flow event.
The transient response of pertinent parameters is shown in Figures
7 XN-NF-86-91(NP) 3.4.1-1 thiough 3.4.1-9.
Figures 3.4.1-6 through 3.4.1-9 give the calculated primary coolant loops flow rate vs. time.
These flow rates can be directly compared to the four primary coolant loop total flow rate of 120.3 Mlb/hr at 29.3% plugging.(5) 3.4.2 Thermal Marcin Analysis - The Minimum Departure from Nucleate Boiling Ratio (MDNBR) for this transient is calculated using thermal hydraulic boundary conditions from the transient analysis with PTSPWR2 as input to XCOBRA-IIIC.(3,4)
Table 3.4.2-1 contains a
summary of conservatively biased input parameters for the Thermal Margin Analysis.
In addition to the 5% power biasing that was mentioned above, the reactor core coolant inlet temperature was biased 5'F (this is in addition to a 5'F measurement uncertainty) in order to conservatively account for less than perfect mixing in the inlet plenum due to the initial ficw asymmetrics arid power split between the two steam generators.
Based on a no load to full power reactor coolant inlet temperature variation of approximately 0.1165'F/%
power, and the steam generator power split from the three pump steady state of approximately 30% power from the unaffected side and 10% power from the affected side, a maximum core coolant temperature tilt of approximately 2.3*F might be expected as compared to the 5'F value that was assumed.
3.5 Conclusion Using the XNB critical heat flux correlation, an initial steady state MDNBR of greater thm 3.0 was calculated.
The transient MDNBR was calculated to be approximately 2.68 which is well above the initial DNBR for all of the l
transients considered in Reference 5.
The XNB critical heat flux safety correlation limit of 1.17 is not violated, so the transient analysis results are acceptable with respect to the MONBR Specified Acceptable Fuel Design Limits (SAFDL).
Over pressurization potential decreases as the reactor heat flux / steam flow ratio following reactor trip is reduced.
Consequently, peak over pressurization is bounded by the rated power Four Primary Coolant Pump Loss of Flow event results presented in Reference 5.
Maximum peak pellet
8 XNNF-86-91(NP) linear heat generation rate (LHGR) for this transient analysis is less than 5.0 kW/ft, well below the incipient fuel centerline melt criterion of 21 kW/ft.-
Applicable acceptance criteria for the transient analysis are therefore satisfied.
t
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XN-NF-86-91(NP)
Table 3.4.1-1 Summary of Initial Operating Conditions (39% Power - Three Primary Coolant Pumps Operating)
Loss of Forced Reactor Coolant Flow k
Nominal PTSPWR Parameter Value Analysis Value Primary Side Total Core Power, MWt 986.7 1037.3 Reactor Coolant System Pressure, psia 2060.0 2052.8 i
Total Coolant Flow Rate Mlbs/hr 92.58*
92.06 Coolant Inlet Temperature, 'F 535.33 539.13 High Power Trip Setpoint, % of 49.0%
54.5%
Rated Power Sagondary Side Total Steam Flow, M1bs/hr 4.2 4.04 Steam Pressure, psia 833.7 856.0 i
p
- Three Primary Coolant Pumps flow rate based on Technical Specification of 74.7% of rated.(5) s
10 XN-NF-86-91(NP)
Table 3.4.1-2 Sequence of Events Summary (39% Power - Three Primary Coolant Pumps Operating)
Loss of Forced Reactor Coolant Flow Event Time (sec)
Primary Coolant Pumps Begin to Coastdown 0.0 Low Reactor Coolant Flow Signal 3.5 Reactor Trip Signal and Initial Shutdown Control Rod Motion 4.0 Minimum DNBR 4.66
l 11 XN-NF-86-91(NP) l l
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Table 3.4.2-1 Summary of Input Parameters for the XCOBRA-IIIC l~
Thermal Margin Analysis for the Loss of Forced l
Reactor Coolant Flow l
l i
Core Flow Parameter /Model Distribution Subchannel l
Pressure (psia)(5) 2010.0 2010.0 Enthalpy (Btu /lb) at 545.33*F(5) 541.75 541.75 l
G (Mlb/hr/ft )
0.9557(3)
.9079(1) 2 ALHGR (kw/ft) 2.34(4) 3.97(2)
I (1) Coolant inlet flow maldistribution factor f
(2) Maximum assembly radial peaking factor l
(3) Derived from PTSPWR output and allows for a 3% flow uncertainty, a 3% core bypass flow, and a 1.8% difference between the initial PTSPWR l
value and that allowed by Technical Specifications i
(4) Derived from the initial PTSPWR value and includes 3 reactor coolant pump pumping power (5) Conservatively assumed initial condition l
l
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i ALHGR - Average Linear Heat Generation Rate.
Low Flow Trip Signal assumed at flow of 55% of rated flow.
Flow at time of minimum DNBR; 50.4% of rated flow.
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21 XN NF 86-91(NP) 4.0 REACTOR COOLANT PUMP ROTOR SE12VRE/ SHAFT BREAK ANALYSIS 4.1 Event Initiators 4
This event is initiated either by the instantaneous seizure of a primary coolant pump rotor or by a failure in the pump shaft of a primary coolant pump, causing the pump impeller to spin freely.
4.2 Event Consecuences In either case, flow in the affected loop will be rapidly reduced, causing the core flow to also decrease rapidly.
The initial rate of flow reduction will be greater for the pump rotor seizure event; however, the final steady-s' tate core flow will be lower for the shaft break event because of the potential for higher reverse flow in the affected cold leg.
Because of the initial reduction in core flow, the coolant temperatures will increase, causing expansion of the primary coolant and subsequent pressurizer insurge flow and reactor coolant system pressurization.
As the pressure increases, pressur'zer sprays and safety valves would act to mitigate the pressure transient and maintain the pressure below the safety valve setpoint.
4.3 Reactor Protection Reactor protection for either of these events is provided by the low reactor coolant fl ow, thermal margin / low pressure, and high pressurizer pressure trips.
4.4 Transient Analysis and Thermal Marain Analysis e
4.4.1 Transient Analysis - Simulation of the system transient response was not perforried.
Conservative boundary conditions were assumed as input to the thermal margin analysis as discussed in the following section.
L-22 XN-NF-86-91(NP) 4.4.2 Thermal Marain Analysis The MDNBR is calculated using conservatively biased input to XCOBRA-IIIC.
The flow is assumed to decrease instantaneously to a two primary coolant pump flow rate of 44%.
This assumed value is less than the design flow rate given in the Palisades Technical Specifications, page 2-7, for two primary coolant pump operation.
Since a reactor trip on low coolant flow will occur earlier in this case than for a loss of forced reactor coolant flow transient, the remaining boundary conditions can be assumed to be the same as for the loss of forced reactor coolant flow; namely, linear heat generation rate (LHGR), reactor coolant inkt temperature, and reactor coolant system pressure.
As previously discussed in Section 3.0, the LHGR assumes a 5% power increase due to moderator feedback, the reactor coolant inlet temperature assumes a 5'F temperature tilt in addition to a 5'F measurement uncertainty, and the reactor coolant system pressure is assumed to remain constant at the initially assumed value of 2010.0 psia.
Using these conservative boundary conditions, the minimum DNBR was approximately 2.54.
This compares with the locked rotor MDNBR from full power four primary coolant pump operation of 1.409.(5) 4.5 Conclusion Using the XNB critical heat flux correlation, the transient MDNBR was calculated to be approximately 2.54.
The XNB critical heat flux safety correlation limit of 1.17 is not violated, so the transient analysis results are acceptable with respect te MONBR Specified Acceptable Fuel Design Limits (SAFDLs).
Overpressurization potential decreases as the reactor heat flux / steam flow ratio following reactor trip is reduced.
Consequently, peak overpressurization is bounded by the Reactor Coolant Pump Rotor Seizure / Shaft Break analysis initiated from a four primary coolant pump operational configuration.(5)
Maximum peak pellet linear heat generation rate (LHGR) for this analysis is less than 5.0 kW/ft, well below the incipient fuel centerline melt criterion of 21 kW/ft.
Applicable acceptance criteria for this analysis are therefore satisfied.
23 XNNF-86-91(NP)
Table 4.4.2-1 Summary of Input Parameters for the XCOBRA-I!!C Thermal Margin _ Analysis for the Reactor Coolant Pump Rotor Seizure / Shaft Break Analysis Core Flow Parameter /Model Distribution.
Subchannel Pressure (psia)(5) 2010.0 2010.0 Enthalpy (Btu /lb) at 545.33*F(5) 541.75 541.75 G (Mlb/hr ft )
0.8749(3)
.8311(I) 2 ALHGR (kW/ft) 2.34(4) 3.97(2) i (1) Coolant inlet flew maldistribution factor.
(2) Maximum assembly radial peaking factor.
(3) Derived from Technical Specifications for two pump flow and allows for 1) a 3% flow uncertainty, 2) a 3% core bypass flow, 3) 3% allowance for asymetrics in steam generator tube plugging and resultant two pump configuration (one pump in each loop operational or two pumps in only one loop operational), and 4) a 4% allowance for any differences between PTSPWR predicted two pump flow and that specified by Technical Specifications.
(4) Derived from the initial PTSPWR value and includes 3 reactor coolant pump pumping power.
(5) Conservatively assumed initial condition.
~__.--
'l_
24 XN NF-86-91(NP) 5.0 CONTROL R00 MISOPERATION The control rod misoperation event encompasses a number of transients and steady-state configurations resulting from different event initiators.
The specific events addressed under this event category include the following:
(1) Dropped control rod or control rod bank; (2) Statically misaligned control rod or control rod bank; and (3) Single control rod withdrawal.
5.1 Droceed control Rod or Control Rod Bank 5.1.1 Event Initiators - The dropped control rod and dropped control bank events are initiated by a de energized control rod drive mechanism or by a malfunction associated with a control rod bank.
5.1.2 Event Consecuences - In the dropped rod or dropped bank events, the reactor power initially decreases in response to the insertion of negative reactivity.
This results in a reduction of the moderator temperature due to a mismatch between core power generation and secondary system load demand.
The core power redistributes due to the local power effect of the dropped rod or bank.
The reactor power will return to essentially the initial power level due to the combined effects of a negative moderator temperature coefficient and Doppler reactivity feedback.
Because of the increased peaking and the potential return to the initial power level, the dropped control rod event poses a severe challenge to the DNB margin.
5.1.3 Reactor Protection - If the amount of reactivity is large enough (bank drop) to cause a significant reduction in core power, a reactor trip would be generated by the low pressurizer pressure trip, low steam generator water level trip, or thermal margin / low pressure trip.
5.1.4 Transient Analysis and Thermal Marain Analysis - To evaluate the
25 XN-NF-86-91(NP) challenge to the DNB margin, the dropped control rod was previously analyzed (5) at rated power operating conditions for which the initial DNB margin is at a minimum.
The consequences of the dropped control rod from rated power conditions bound the other power operation conditions for four reactor coolant pump operation.
4 In the case of three primary coolant pump operation with approximately 75% of rated flow and a potential for a 5'F reactor coolant inlet temperature tilt, the maximum allowed operating power of 39% provides a sufficient increase in the flow to power ratio compared to the rated pcwer four reactor coolant pump operetion that the rated power four reactor coolant pump operation condition remains limiting.
Therefore for three reactor coolant pump operation at 39%
power, this event remains bounded by the results presented in Reference 5.
5.2 Statically Misalioned Control Rod or Control Rod Bank The static misalignment events occur when a 5.2.1 Event Initiators malfunction of the Control Rod Drive acchanism causes a control rod to be out of alignment with its bank, i.e., either higher of lower than any of the other l
control rods in the same bank or when a bank (s) is out of alignment with the Power Dependent Insertion Limit (PDIL).
5.2.2 Event Consecuences - In the static control rod misalignment event, a control bank is inserted but one of the control rods remains in a withdrawn state.
This results in a local increase in the radial power peaking factor and a corresponding reduction in the DNB margin.
The most severe misalignment occurs with bank 4 inserted beyond its insertion limit and one of the control l
rods in bank 4 fully withdrawn.
The radial power redistribution consequences of a reverse misalignment, wherein one rod is inserted while the bank remains withdrawn, are essentially the same as the dropped rod event and would be bounded by such.
The bank misalignment event occurs when bank 4 is inserted or withdrawn beyond the PDIL; or when bank 4 and bank 3 are in violation of the PDIL. (Note: The Power Dependent Insertion Limits for control rod banks 3
26 XN-NF-86-91(NP) and 4 are significantly different for three and four primary coolant pump operation. See Figure 5.2.2-1.)
Reactor protection for the statically 5.2.3 Reactor Protection misaligned control rod or control rod bank is provided by operating procedures and Technical Specifications.
The procedures require that control rod insertion meet Technical Specifications for three primary coolant pump operation (Technical Specification Figure 3.6 (Figure 5.2.2-1 this report]).
Also, the thermal and hydraulic safety limit lines shown in Figure 2-?.(1)
(Technical Specifications) for three primary coolant pump operation define the limiting values of primary coolant pressure, reactor inlet temperature, and core power level for which the criteria on minimum DNBR and parallel channel flow stability are met.
Finally, per Technical Specification 2.3.4, for three primary coolant pump operation, power is limited to a maximum of 39% of rated power for a maximum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
During this mode of operation, the high power level trip in conjunction with the TM/LP (minimum setpoint - 1750 psia) and the secondary system safety values (set at 1000 psia) assure that the limits shown on Figure 2-2(I) will not be violated.
5.2.4 Transient An61vsis and Thermal Marain Analysis - In the analysis of a statically misaligned control rod, conservative steady state values for the primary coolant system pressure, core inlet temperature, and coolant flow rate; and a radial peaking augmentation fa: tor to bound the radial power redistribution of a misaligned control rod would be input into the XCOBRA-lllC code to calculate the MDNBR.
Thus, the statically misaligned control rod event is dependent upon the maxiraum augmentation factor that is calculated for a control rod bank inserted beyond its PDil with one rod stuck out of the core.
The PDIL for the three pump operation at 39% power allows only rod bank 4 to be inserted 40%.
The analysis presented in Reference 5 for a single control rod withdrawal from 50% power 91th four coolant pumps operating gives a maximum augmentation factor of 1.289 for rod bank 4 inserted to its PDIL of 80%, and rod bank 3 inserted to its PDIL of 20%. Due to this large difference in rod bank insertions between four pump operation and three pump operation,
l 27 XN-NF-86-91(NP) the 50% power single rod withdrawal augmentation factor will bound the augmentation fac. tor for three pump operation at 39% of rated power.
Consequently, since the values of the pressure, inlet temperature and flow rate used in the single ~ control rod withdrawal precented in Section 5.3 either bounds or is equal to the values for the statically misaligned control rod event, the results of the single control rod withdrawal will bound the i
statically misaligned control rod event.
In the case of bank misalignment considered in the Reference 5 analysis, augmentation factors for bank misalignments were calculated by XTGPWR using a 16 inch misalignment from PDIL at 50% and 65% power levels.
The 50% power level was chosen because the Technical Specifications allow the largest radial peaking at that level (25% higher than rated power peaking).
The 65% power level was chosen because according to the 4 primary coolant pump operation PDIL bank 3 begins insertion at that level.
The positions of banks 3 and 4 were individually withdrawn and inserted 16 inches beyond PDIL to determine the highest augmentation factor for each power level.
XCOBRA-IIIC calculations were made using these augmentation factors and rated power pressure, core inlet temperature and flow rate to determine the MDNBR.
The MDNBRs at 50% and 65% power were 2.72 and 2.09, respectively.(5)
For a given bank (s) insertion, the three primary coolant pump operation PDil for control bank 4 and 3 restricts the allowed power to approximately 50% of that allowed for four primary coolant pump operation (Figure 5.2.2-1).
As such, the power reduction to the allowed power for three primary coolant operation is sufficient to account for the approximately 25% flow reduction and the potential coolant inlet temperature tilt that would be predicted for three primary coolant pump operation.
Therefore, the consequences of control rod bank misalignments for three primary coolant pump operation (reactor power 39% of rated) are bounded by the control bank misalignment analysis presented in Reference 5.
28 XH-NF-86-91(NP) 5.3 Sinale Control Rod Withdrawal The single control rod withdrawal event is 5.3.1 Event Initiators initiated by an electrical or mechanical failure in the Rod Control System that causes the inadvertent withdrawal of a single control rod.
As the control rod is withdrawn from the 5.3.2 Event Consecuences reactor core, it causes an insertion of positive reactivity which results in an increasing power transient.
The movement of a single rod out of sequence from the rest of the bank also results in a local increase in the radial power peaking factor.
The combination of these two factors results in a challenge to the DNB margin.
5.3.3 Reactor Protection - Reactor protection for the Single Control Rod Withdrawal event is provided by the variable over power, thermal margin / low pressure, and high pressurizer pressure trips.
The transient 5.3.4 Transient Analysis and Thermal Harain Analysis system response for the Single Control Rod Withdrawal event and the Uncontrolled Bank Withdrawal at Power event is essentially the same.
The principal difference is the larger local peaking that would be expected for the Single Control Rod Withdrawal event.
Also, it is not necessary to perform a simulation of '.he system transient.
That is, conservative boundary l
conditions can be selected based on the steam generator power sharing, the steam generator safety valve opening pressure, the variable overpower trip setpoint, the initial pressure and coolant flow rate, and the maximum predicted augmantation factor.
The effects of steam generator power sharing and steam generator safety valve opening pressure, in conjunction with the variable overpower trip setpoint (54.5Y. of rated power), define the maximum potential coolant temperature increase (~5'F for mixing / power sharing effects and 33*F to open the steam generator safety valves).
The assumed coolant inlet temperature was 578'F.
The initial pressure and coolant flow rate are 2
conservatively assumed to be 2010.0 psia and 1.5229 Hibm/hr-ft, respectively.
I-29 XN-NF 86-91(NP)
Note:
The Power Dependent Insertion Limits for control rod banks 3 and 4 are significantly different for three and four primary coolant pump operation, Figure 5.2.2-1.
As such, the maximum predicted augmentation factor for 39%
power /three primary coolant pump operation should be less than that predicted for 52% power /four primary coolant pump operation.
However for conservatism, i
I
}
the maximum predicted augmentation factor for 52% power /four primary coolant pump operation was assumed for 39% power /three primary coolant pump operation, Inputting these values, see Table 5.4.4-1, into the XCOBRA-IIIC thermal margin t
methodology resulted in a minimum DNBR of 1.20, which is above the criteria of 1.17.
Also, these values are conservative relative to those that would be obtained from the safety limit lines given in Technical Specifications, Figure 2-2.(I)
I l
1
30 XN-NF-86 91(NP)
Table 5.4.4-1 Summary of Input Parameters for the XCOBRA-IIIC Thermal Margin Analysis for the Single Control Rod Withdrawal Event Core Flow Parameter /Model Distribution Subchannel i
Pressure, psia (4) 2010.0 2010.0 Enthalpy (Btu /lb) at 578'F(5) 583.82 583.82 G (Mlbm/hr-ft )(6) 1.4494 1.3769(3) 2 ALHGR (kw/ft)(7) 2.96 6.48(1,2)
(1) Maximum assembly radial peaking factor (2)
Radial peaking augmentation factor (3) Coolant inlet flow maldistribution factor (4) Conservatively assumed initial condition that would result in opening of the (5) Conservatively calculated T l
steamgeneratorsafetyvalvd9et (6) Derived from PTSPWR output for the three pump coastdown event and allows for 1) a 3% flow uncertainty, 2) a 3% core bypass flow, aad
- 3) a 1.8% difference between the initial PTSPWR value and that allowed by Technical Specifications (7) Core Average Linear Heat Rate at 54.5% power
4 31 XN NF-86-91(NP)
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i s
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MAXt&W POWER LEVEL y
i 1"
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3 marmob 'NDS WWWmDN PWEWWF l
Ceefmol 200 seeENT10B0 lasts m",,,gyema,a W f
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Figure 5.2.2-1 Power Depend)nt Instrtion Limits for two, three, and four primary coolsrit eume operation i
i 1
32 XN-NF-86-91(NP) 6.0 PTSPWR2 MODEL FOR CE 2X4 LOOP PLANTS Combustion Engineering (CE) split cold leg (2x4) plants consist of two hot legs, two steam generators, and four cold legs with two cold legs connecting each steam generator to the reactor vessel.
Each of the four cold legs includes a separate primary coolant pump.
During the flow coastdown initiated by a pump trip or locked rotor occurrence, the ficM rate in an affected cold leg can be appreciably different than the flow rate in the parallel unaffected leg.
Indeed under some conditions, reverse flow may exist in the affected cold leg, while forward flow is maintained in the hot leg connected to the two cold legs of interest.
The current approved version of the PTSPWR2 code (2) was not specifically programmed for this asymmetric flow situation in CE 2x4 type plants where no back flow exists in the affected side hot leg.
In order to provide the general capability to analyze transients for CE 2x4 plants with one pump inoperative, the PTSPWR2 code has been modified.
In particular, the hydraulics equations used to determine tic. primary loop flow rates were modified to allow for four cold legs instead -
two and to allow for steady state initialization with reverse flow in the loop with the inoperative pump.
Four simultaneous equations instead of two are now solved f
to obtain the flow rate derivatives for each of the four primary coolant cold l
legs.
These derivatives are obtained from the previous time step calculations and are integrated to determine the updated flow rate values.
This revised l
version of PTSPWR2, which was benchmarked against the RELAP5/M002 code, has been used in this analysis to calculate the three pump steady state operating conditions and the flow coastdown from three pump operation.
i In the generic ENC thennal margin methodology, the reactor coolant system pressure, reactor coolant flow rate, core average heat flux, and. coolant l
inlet temperature (or enthalpy) are typically transferred from the PTSPWR system transient code to the XCOBRA-IIIC taermal margin code.
Each of these variables are biased about the initial or transient value in a conservative l
manner for the specific analysis.
The biasing applied to each parameter is
.r r
33 XN-NF-86-91(NP) described in this report in Tables 3.4.2-1, 4.4.2-1, and 5.4.4-1.
34 XN-NF-86-91(NP)
7.0 REFERENCES
1.
Palisades Plant Technical Specifications.
2.-
Descriotion of the Exxon Nuclear Plant Technical Simulation Model for Pressurized Water Reactors (PTS-PWR),
XN-74-5(P)(A),
Rev.
2, Exxon Nuclear Company, October 1986.
3.
Acolication of Exxon Nuclear Comoany PWR Thermal Marain Methodoloav to Mixed Core Confiaurations, XN-NF-82-21(P)(A), E.v.xon Nuclear Company, September 1983.
4.
XCOBRA-IIIC Users Manual, RA-83-3, R.K. Henke and D.S. Rowe, April 30, 1983.
5.
Palisades Cycle 7 Safety Analysis Reoort. Analysis of Chaoter 15 Events, XN-NF-85-94(P), Supplement 2 (to be published).
l
~
XN-NF-86-91(NP)
Issue Date: 6/13/88 LOW FLOW TRIP SETPOINT AND THERMAL MARGIN ANALYSIS FOR THREE PRIMARY COOLANT PUMP OPERATION 0F THE PALISADES REACTOR Distribution J. S. Holm L. A. Nielsen L. D. O' Dell CPCo/HG Shaw (25)
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