ML17285A300
| ML17285A300 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 01/31/1989 |
| From: | Krajicek J SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML17285A298 | List: |
| References | |
| ANF-89-01, ANF-89-1, TAC-72251, NUDOCS 8903090450 | |
| Download: ML17285A300 (54) | |
Text
ANF 89 01 i Al@VAHCSSHUClLEARFQSlLS CORPORATION WNP-2 CYCLE 5 PLANT TRANSIENT ANALYSIS JANUARY 1989 8903090450 8
0397 pgR ADO K'gU p
ADVANCEDNUCLEARFUELS CORPORATION ANF-89-01 Issue Date:
1/16/89 WNP-2 CYCLE 5 PLANT TRANSIENT ANALYSIS Prepared by E. Krajicek BW Safety Analysis Licensing and Safety Engineering Fuel Engineering and Techni,cal Services January 1989
NUCLEAR REGULATORY COMMISSION REPORT DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT 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 pontrlbutlon to facilitate safety analyses by licensees of the U.S. Nuclear Regulatory Commission which utilize Ad.
vanced Nuclear Fuels Corporation fabricated reload fuel or other technical services provided by Advanced Nuclear Fuels Corporation for light water power reactors'and lt 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 compliance with the U.S. Nuclear Regulatory Commission's regulations.
Advanced Nuclear Fuels Corporation's warranties and representations con.
cerning 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 nor 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, or that the use of any information, apparatus, method, or process disclosed In this document willnot infringe privately owned rights, or B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any Information, ap.
paratus, method. or process disclosed In this document.
ANF1i5 629A (4/88)
ANF-89-01 Page i TABLE OF CONTENTS Section
1.0 INTRODUCTION
2.0
SUMMARY
3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN.
- 3. 1 Design Basis 3.2 Anticipated Transients 3.2. 1 Load Rejection Without Bypass 3.2.2 Feedwater Controller Failure 3.2.3 Loss of 'Feedwater Heating 3.3 Calculational Model 3.4 Safety Limit.
3.5 Final Feedwater Temperature Reduction
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,9 4.3 Closure of All Main Steam Isolation Va 5.0 RECIRCULATION FLOW RUN-UP
6.0 REFERENCES
APPENDIX A HCPR SAFETY LIMIT
- 4. 0 MAXIHUM OVERPRESSUR IZATION 4.1 Design Bases.............
4.2 Pressurization Transients 1 ves
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ANF-89-01 Page ii LIST OF TABLES Table 2.1 3.1 3.2 3.3 5.1 THERMAL MARGIN
SUMMARY
fOR WNP-2 CYCLE 5 DESIGN REACTOR AND PLANT CONDITIONS FOR WNP-2 SIGNIFICANT PARAMETER YALUES USED IN ANALYSIS FOR WNP-2 RESULTS OF SYSTEM PLANT TRANSIENT ANALYSES. '.
REDUCED FLOW MCPR OPERATING LIMIT FOR WNP-2
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10ll 14 28 LIST OF FIGURES
" ~iciure
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3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 3.10 5.1 A.l A.2 A.3 A.4 A.S A.6 LOAD REJECTION WITHOUT BYPASS
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LOAD REJECTION WITHOUT BYPASS
- RESULTS, RPT OPERABLE, NORMAL SCRAM SPEED LOAD REJECTION WITHOUT BYPASS RESULTS, RPT INOPERABLE, NORMAL SCRAM SPEED LOAD REJECTION WITHOUT BYPASS RESULTS, RPT INOPERABLE, NORMAL SCRAM SPEED LOAD REJECTION WITHOUT BYPASS
SPEC.
SCRAM SPEED LOAD REJECTION WITHOUT BYPASS RESULTS, RPT OPERABLE, TECH.
SPEC.
SCRAM SPEED LOAD REJECTION WITHOUT BYPASS RESULTS, RPT INOPERABLE, TECH.
SPEC.
SCRAM SPEED LOAD REJECTION WITHOUT BYPASS RESULTS, RPT INOPERABLE, TECH.
SPEC.
SCRAM SPEED FEEDWATER CONTROLLER FAILURE RESULTS FOR 47%
POWER AND 106%
FLOW WITH NORMAL SCRAM SPEED FEEDWATER CONTROLLER FAILURE RESULTS FOR 47%
POWER AND 106%
FLOW WITH NORMAL SCRAM SPEED REDUCED FLOW MCPR OPERATING LIMIT WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF-4 FUEL)
WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-3 FUEL)
WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-2 FUEL)
WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-1 FUEL)
WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS (GE FUEL)
RADIAL POWER HISTOGRAM FOR I/O CORE SAFETY LIMIT MODEL.
15 16 17 18 19 20 21 22 23 24 29 A-5 A-6 A-7 A-8 A-9 A-10
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ANF-'89-01 Page iii CKNOWLEDGMENT The author wishes to acknowledge the contribution made to this report by fellow Advanced Nuclear Fuels Corporation employees H.
E.
- Byram, S. J.
- Haynes, and D. J.
Braun.
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ANF-89-01 Page 1
1.0 INTRODUCTION
This report presents the results of the Advanced Nuclear Fuels Corporation (ANF) evaluation of system transient events for the Supply System Nuclear Project Number 2
(MNP-2) during Cycle 5 operation.
For this analysis the Cycle 5 core was, assumed to contain 572 ANF 8x8 and 192 GE P8x8R fuel assemblies.
This evaluation together with the analysis of final feedwater temperature reduction(
)
(FFTR) and the analysis of core transient events(
) determines the necessary thermal margin (HCPR limits) to protect against boiling transition during the most li'miting an'ticipated operational occurrence (AOO).
The evaluation also demonstrates the vessel integrity for the most limiting pressu'rization event.
This evaluation is applicable for core flows up to the maximum attainable with the recirculation flow control valve in its fully open position which is 106% of the rated core flow value at 100%
power.
The methodology used for these system transient analyses is detailed in References 3 and 4.
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r 2.0
SUMMARY
ANF-89-01 Page 2
The Minimum Critical Power Ratios (HCPR) calculated to protect against boiling transition during potentially limiting plant system transient events are shown in Table
- 2. 1 for powers that bound allowable values.
The system transient HCPR values of Table
- 2. 1 for the load rejection without bypass (LRNB) and feedwater controller failure (FWCF) transients were obtained using a
scram time based on WNP-2 measured values.
The loss of feedwater heating (LOFH) transient results shown in Table
- 2. 1 were obtained from a
bounding analysis which is discussed in Section 3.2.3.
The limiting AOO values for the cases of Table 2. 1 are for the LRNB transient at End of Cycle conditions; the limiting HCPR value's are 1.34 for GE fuel and 1.31 for ANF fuel.
For previous WNP-2 cycles, ANF performed an analysis for the LRNB event at a cycle exposure of EOC -2000 HWd/HTU.
Prior to the end of cycle, a large number of control blades are still inserted in the core.
These analyses showed that this LRNB system transient was bounded by the control rod
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withdrawal event (CRWE) by a
substantial margin.
- Thus, for the earlier
- cycles, plant operating limits were always based on the CRWE for cycle exposures up to EOC -2000 MWd/HTU.
Based on this prior experience, the Cycle 5
MCPR limit up to EOC -2000 has been determined only by the CRWE.(
)
- Thus, the Cycle 5
CRWE defined HCPR limit is applicable up to EOC -2000 HWd/HTU, and for exposures beyond EOC -2000 HWd/MTU the limits in Table 2. 1 are applicable.
Additional transient analyses were performed assuming the recirculation pump trip (RPT) was out of service, and using the technical specification scram speed (TSSS) and the results are reported herein.
The critical power results for these events are presented in Section 3.0.
The maximum system pressure was calculated for the containment isolation event which is a
rapid closure of all main steam isolation valves.
This analysis shows that for WNP-2 Cycle 5 operation, the safety valve response prevents the pressure from reaching 110/ of design pressure.
The maximum It system pressures predicted during the event are below the ASHE Code limit of
ANF-89-01 Page 3
110% of design pressure (1375 psig) and are shown in Table 2.1.
The analysis conservatively assumed six safety relief valves out of service.
The continued applicability of the previously established MCPR safety limit of 1.06 in Cycle 5
was confirmed for all fuel types using the methodology of Reference 6.
TABLE 2.1 THERMAL MARGIN
SUMMARY
FOR WNP-2 CYCLE 5 ANF-89-01 Page 4
Transient 5 Power
% Flow Delta CPR MCPR*
GE Fuel ANF Fuel Load Rejection**
Without Bypass 104/106 0.28/1.34 0.25/1.31 Feedwater Controller**
Failure 47/106 0.23/1.29 0.20/1.26 Loss of Feedwater***
Heating Not Applicable 0.09/1.15 0.09/1.15 MAXIMUM PRESSURE (PSIG)
Transient MSIV Closure Vessel Dome 1286 Vessel Lower Plenum 1315 Steam Line 1289
- MCPR value using the 1.06 safety limit justified herein.
- These transients were evaluated with normal scram
~em
ANF-89-01 Page 5
3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN 3.1
~Bi 3
System analyses were performed at the increased core flow condition 'of 106%
to determine the most limiting type of system transients for the establishment of thermal margins.,
As shown in Reference 5,
system transients from the increased core flow condition bound transients from the nominal (100/o) flow condition.
Analysis of the LRNB was performed at the rated design 104/
power/106%
flow point.
Since feedwater controller failure (FWCF) transients may be more severe at reduced power because of the larger change in feedwater flow, a
FWCF transient was performed at the minimum power (47%) that allowed for increased core flow.
The initial conditions used in the analysis for transients at the 104/
power/106% flow point are as shown in Table 3. 1.
The most limiting exposure in cycle was determined to be at end of full power capability when control rods are fully withdrawn from the core; the thermal margin limit established for end of full power conditions is conservative in
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h relation to cases where control rods are partially inserted.
The calculational models used to analyze these pressurization events include the ANF plant transient and core thermal-hydraulic codes as described in previous documentation.(
, >>")
Fuel pellet-to-clad gap condu'ctances used in the analyses are based on calculations with RODEX2.(
)
Recirculation pump trip (RPT) coastdown was input based on measured WNP-2 startup test data, and the COTRANSA system transient model for WNP-2 was benchmarked to appropriate WNP-2 startup test data.
The hot channel performance is evaluated with XCOBRA-T'( ) using COTRANSA supplied boundary conditions.
Table 3.2 summarizes the values used fo} important parameters in the analysis.
3.2 Antici ated Transients ANF transient analysis methodology for Jet Pump BWR's considers eight categories of potential system transient occurrences.(
)
The three most limiting transients for WNP-2 are presented in this section; these transients are:
ANF-89-01 Load Rejection Without Bypass (LRNB)
Feedwater Controller Failure (FWCF)
Loss of Feedwater Heating (LOFH).
I A summary of the transient analyses is shown in Table 3.3.
Other plant transient events are inherently nonlimiting or clearly bounded by one of the above events.
3.2.1 Load Re 'ection Without B
ass This event is the most limiting of the class of transients characterized by rapid vessel pressurization.
The generator load rejection causes a turbine control valve trip, which initiates a reactor scram and a recirculation pump trip (RPT).
The compression wave produced by the fast turbine control valve closure travels through the steam lines into the vessel and pressurizes the reactor'essel and core.
Bypass flow to the condenser, which would mitigate the pressurization
- effect, is 'conservatively not allowed.
The excursion of core power due to void collapse is primarily terminated by reactor scram and void growth due,to RPT.
Figures
- 3. 1 through 3. 10 depict the time variance of
'ritical reactor and plant parameters from the analyses of several load
'.rejection transients.
Transient analysi's cases include the design basis power and increased core flow point with a matrix of cases which involve normal scram
- speed, technical specification scram
- speed, and recirculation pump trip (RPT) in service and out of service.
Analysis assumptions are:
Control rod insertion time based on WNP-2 measured data (normal scram speed) or minimum technical specification scram speed.
Integral power to the ho't channel was increased by 10% for the pressurization transient",
consistent with Reference 9.
ANF-89-01 Page 7
Table 3.3 shows delta CPR values for a matrix of LRNB transients with the RPT out of service with both normal scram speed (NSS) and technical specification scram speed (TSSS).
ANF has previously analyzed the LRNB event for prior cy'cles at an exposure of EOC -2000 HWd/HTU.
Since a significant number of control rods are inserted into the core up to end-of-cycle (EOC) minus 2000 HWd/MTU, this prior analytical experience has shown'he CRWE to be clearly bounding from the beginning-of-cycle (BOC) up to this point.
That is, the limiting delta CPR or HCPR limit throughout the earlier part of the cycle was set by the CRWE from BOC to EOC
-2000 HWd/HTU.
For Cycle 5
an LRNB calculation at EOC
-2000 MWd/HTU has not been provided because the CRWE clearly sets the HCPR limit up to this exposure.
For Cycle 5 exposures greater than EOC minus 2000 HWd/HTU, HCPR values defined in Table 3.3 are applicable.
3.2.2 Feedwater Controller Failure
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Failure of the feedwater control system is postulated to lead to a
maximum increase in feedwater flow into the vessel.
As the excessive feedwater flow subcools the recirculating water returning to the reactor core, the core power will rise and attain a
new equilibrium if no other action is taken.
Eventually, the inventory of water in the downcomer will rise until the high vessel level trip setting is exceeded.
To protect against wet steam entering the turbine, the turbine trips upon reaching the high level setting, closing the turbine stop valves.
The compression wave that is created, though mitigated by bypass flow, pressurizes the core and causes a power excursion.
The power increase is terminated by reactor
The evaluation of this event was performed using the scram and integral power assumptions discussed in Section 3.2. 1.
Sensitivity results have shown that EOC conditions are bounding because rods are inserted for lower cycle exposures, and high flows are bounding because of higher axials in the core.
Reference ll showed that the LRNB is more limiting at full power than the FWCF.
Because the total change in feedwater flow is the greatest from reduced
ANF-89-01 Page 8' power condition, the FWCF was analyzed from reduced power conditions.
The FWCF was analyzed with the feedwater flow rate increasing at a rate between 10 and 25 percent of nuclear boiler rated (NBR) flow per second.
The FWCF transient event was analyzed from the lowest allowed power (47%) at increased core flow.
Figures 3.9 through 3.10 present key variables.
The delta CPR E
values for the co-resident fuel types for 47%
power/106% flow transient are shown in Table 3.3.
Table 3.3 shows that the delta CPR/MCPR value for the FWCF 'is less than the delta CPR/MCPR value for the 104/106 LRNB event with RPT operable and normal scram speed.
3.2.3 Loss of Feedwater Heatin Loss of Feedwater Heating (LOFH) events were evaluated for Cycle 5 with the ANF core simulator model XTGBWR(10) by representing the reactor in equilibrium before and after the event.
Actual and projected operating statepoints were used as initial conditions.
Final conditions were determined I
by reducing the feedwater temperature by 100'F and increasing core power such that the calculated eigenvalue remain unchanged.
I Based on a bounding value analysis, a
MCPR limit of 1. 15 for WNP-2 with a MCPR safety limit of 1.06 is supported (i.e.,
a delta CPR of 0.09).
As shown in Appendix A of this
- report, the WNP-2 MCPR safety limit for Cycle 5
continues to be 1.06;
- hence, the LOFH transient requires a
MCPR limit of 1. 15 for WNP-2.
3.3 Calculational Model The plant transient codes used to evaluate the pressurization transients (generator load rejection and feedwater flow increase) were the ANF advanced codes COTRANSA( ) and XCOBRA-T.(4)
This axial one-dimensional model predicted reactor power shifts toward the core middle and top as pressurization occurred.
This was accounted for explicitly in determining thermal margin changes in the transient.
All pressurization transients were analyzed on a
bounding basis using COTRANSA in conjunction with the XCOBRA-T hot channel model.
The XCOBRA-T code 'as used consistent with the benchmarking methodology.
ANF-89-01 Page 9
3.4 ~fLi The HCPR safety limit is the minimum value of the critical power ratio (CPR) at which the fuel could be operated where the expected number of rods in boiling transition would not exceed
- 0. 1% of the fuel rods in the core.
The operating limit HCPR is established such that in the event the most limiting anticipated operational transient
- occurs, the safety limit will not be violated.
The safety limit for all fuel types in WNP-2 Cycle 5 was confirmed by the methodology presented in Reference 6 to have the Cycle 2 value of 1.06.
The input parameters and uncertainties used to establish the safety limit are presented in Appendix A of this report.
3.5 Final Feedwater Tem erature Reduction Reference 1
presents final feedwater temperature reduction (FFTR) analysis with thermal coastdown for WNP-2 for Cycles 3
and 4.
The FFTR
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analysis was performed for a 65'F temperature reduction.
These FFTR analyses are applicable after the all rods out condition is reached with normal feed-water temperature.
The FFTR analysis results show that delta CPR changes for the LRNB and FWCF transients are conservatively bounded by adding 0.02 to the delta CPR values for these transients at normal feedwater temperatures.
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ANF-89-01 Page 10 TABLE 3.1 DESIGN REACTOR AND PLANT CONDITIONS FOR WNP-2 Reactor Thermal Power (104/)
Total Recirculating Flow (106%)
Core Channel Flow Core Bypass Flow Core Inlet Enthalpy Vessel Pressures Steam Dome Upper Plenum Core Lower Plenum Turbine Pressure Feedwater/Steam Flow Feedwater Enthalpy
'ecirculating Pump Flow (per pump) 3464 HWt 115.0 Hlb/hr 102.4 Hlb/hr 12.3 Hlb/hr 527.8 BTU/lb!0 1036. psia 1049. psia
" 1056. psia-1073. psia 978. psia 14.8 Hlb/hr 391.1 BTU/ibm 16.3 Hlb/hr 0
ANF-89-01 Page ll TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN ANALYSIS FOR WNP-2 High Neutron Flux Trip Void Reactivity Feedback Time to Deenergized Pilot Scram Solenoid Valves Time to Sense Fast Turbine Control Valve Closure Time from High Neutron Flux Time to Control Rod Motion Normal 126 2%
10% above nominal*
200 msec 80 msec I
290 msec Tech Spec Scram Insertion Times**
0.404 sec 0.660 sec 1.504 sec 2.624 sec 0.430 sec 0.868 sec 1.936 sec 3.497 sec to Notch 45 to Notch 39 to Notch 25 to Notch 5
Turbine Stop Valve Stroke Time Turbine Stop Valve Position Trip Turbine Control Valve Stroke Time (Total)
Fuel/Cladding Gap Conductance Core Average (Constant)
Safety/Relief Valve Performance Settings Relief Valve Capacity Pilot Operated Valve Delay/Stroke 100 msec 90% open 150 msec 587. BTU/hr-ft2-F Technical Specifications 228.2 ibm/sec (1091 psig) 400/100 msec
- For rapid pressurization transients a
10% multiplier on integral power is used; see Reference 9 for methodology description.
- Slowest measured average control rod insertion time to specified notches for each group of 4 control rods arranged in a 2x2 array.
ANF-89-01 Page 12 TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN ANALYSIS FOR WNP-2 (Continued)
MSIV Stroke Time HSIV Position Trip Setpoint Condenser Bypass Valve Performance Total Capacity Delay to Opening (801. open)
Fraction of Energy Generated in Fuel Vessel Watch Level (above Separator Skirt)
High Level Trip (L8)
Normal Low Level Trip (L3)
Maximum Feedwater Runout Flow Two Pumps Recirculating Pump Trip Setpoint 3.0 sec 85% open 990. ibm/sec 300 msec 0.965 73 ln 49.5 in 21 in I
5799. ibm/sec 1170 psig
/
Vessel
- Pressure,
ANF-89-01 Page 13 TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN ANALYSIS FOR WNP-2 (Continued)
Control Characteristics Sensor Time Constants Steam Flow Pressure Others Feedwater Control Mode Feedwater 100% Mismatch Water Level Error Steam Flow Equiv.
Flow Control Mode Pressure Regulator Settings Lead Lag Gain 1.0 sec 500 msec 250 msec Three-Element 48 in 100%
Manual 3.0 sec 7.0 sec 3 '%/psld
ANF-89-01 Page 14 TABLE 3.3 RESULTS OF SYSTEM PLANT TRANSIENT ANALYSES Event LRNB RPT Operable, NSS*
Maximum'eutron Flux
/ Rated 403 Maximum Core Average Heat Flux
/ Rated 121 Maximum System Delta CPR Pressure GE ANF Qi~si ~
Fuel 1169 0.28 0.25 LRNB RPT Inoperable, NSS LRNB RPT Operable, TSSS**
LRNB RPT Inoperable, TSSS 501 454 594 127 127 132 1181 0.35 1189 0.41 1174 0.35 0.31 0.31 0.35 FWCF (47% Power/106%
Flow),
NSS RPT Operable HSIV Closure With Flux Scram 163 708 133 1315 54 1026 0.23 0.20 N/A NOTE:
All results are for the design power and increased flow point (104%
power/106% flow) unless otherwise noted.
- Technical Specification Scram Speed (TSSS).
Cl tD
- i. NEUTRON FLUX LEVEL
- 2. HEAT FLUX
- 3. RECIRCULATION FLOW
- 4. VESSEL STEAM FLOW
- 5. FEEOWATER FLOW o~O I-O IX,'
Z O CV CJ CE LLJ CL Cl i2345 i23 4
Cl I
ClO Io.o 0.2 0.5 0.7 i.o i.2
- TIME, SEC i.5 i.7 2.0 2.2 2.5 FIGURE 3.1 LOAD REJECTION WITHOUT BYPASS RESULTS, RPT OPERABLE, NORMAL SCRAM SPEED
- i. VESSEL PRESSURE CHANGE (PSI)
- 2. VESSEL WATER LEVEL tIN) 1 1
CI 1
c4.0 0.2 0.5 0.7 1.0 1.2
- TIME, SEC 1.5 1.7 2.0 2.2
- 2.5 FIGURE 3.2 LOAD REJECTION WITHOUT BYP <
o lD oo lA
- i. NEUTRON FLUX LEVEL
- 2. HEAT FLUX
- 3. RECIRCULATION FLOH
- 4. VESSEL STEAM FLOH
- 5. FEEOHATER FLOH oo Cl~o
+o
~ m GC o
Ill Ol D
tK LLJ CL o 12345 123 3 4 oI oo c 0.0 0.2 0.5 0.7 1.0 1.2
- TIME, SEC 1.5 1.7 2.0 2.2 2.5 FIGURE 3.3 LOAD REJECTION WITHOUT BYPASS RESULTS, RPT INOPERABLE, NORMAL SCRAM SPEED
CI F
~H CI Ol
- l. VESSEL PRESSURE CHANGE (PSI)
- 2. VESSEL HATER LEVE (I
CI
'IPf t
W.O 0.2 0.5 0.7 1.0 1.2
- TXiilE, SEC i.7 2.0 2.2 2.6 FIGURE 3.4 LOAD REJECTION MITHOUT BYPASS
- SULTS, RPT INOPERABLE, NORMAL SCRAM SPEED
oo tD o
LQ
- 1. NEUTRON FLUX LEVEL
- 2. HEAT FLUX
- 3. RECIRCULATION FLOW
- 4. VESSEL STEAM FLOW
- 5. FEEOWATER FLOW o
o~o
+ o
~ m CC o
LU CV C3 CLill CL oo 12345 123 4
o I
oo
%VI0.0 0.2 0.5 0.7 1.0 1.2
- TIME, SEC 1.5.
1.7 2.0 2.2 2.5 FIGURE 3.5 LOAD REJECTION WITHOUT BYPASS RESULTS, RPT OPERABLE, TECH.
SPEC.
SCRAM SPEED
. i. VESSEL PRESSURE CHANGE (PSl)
- 2. VESSEL HATER LEVEL (I
Cl Cl Ol c4.0 0.2 0.5 Q.?
f.0 f.2 i.5
'XHE, SEC f.7 2.Q 2.2 2.5 FIGURE 3.6 LOAO REJECTION MITHOUT BYPASS RESULTS, RPT OPERABLE, TECH.
SPEC.
SCRAH SPEED
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- i. NEUTRON FLUX LEVEL
- 2. HEAT FLUX
- 3. RECIRCULATION FLOW
- 4. VESSEL STEAM FLOW
- 5. FEEOWATER FLOW CI Cl CD~o
+o W m CC OO OJ C7 LE ILI CL CI 12345 123 3 4 4
< 0.0 0.2 0.5 0.7 1.0 1.2 1.5
- TIME, SEC 1.7 2.0 2+2 2.5 FIGURE 3.7 LOAD REJECTION WITHOUT BYPASS RESULTS,
- RPT, INOPERABLE, TECH.
SPEC.
SCRAM SPEED
C) 1.
VESSEL PRESSURE CHANGE (PSI) 2.
VESSEL WATER LEVEL (IN)
Cl CI CI Ol i
c8.0 0.2 0.5 0.7 i.0'IME, SEC i.5 i.7 2.0 2.2 2.5 FIGURE 3.8 LOAD REJECTION WITHOUT BYPASS RESULTS, RPT INOPERABLE, TECH.
SPEC.
SCRAH SPEED 0
- 1. NEUTRON FLUX LEVEL
- 2. HEAT FLUX
- 3. RECIRCULATION FLOW
- 4. VESSEL STEAM FLOW,
- 5. FEEOHATER FLOW O~o CU IX 3
I-O CO C3IZ LLI Q.
12 12 12 4
C)
IP 10
- TIME, SEC 12 18 20 NO FIGURE 3. 9 FEEDWATER CONTROLLER FAILURE RESULTS FOR 47%
POWER AND 106%
FLOW WITH NORMAL SCRAM SPEED
CI Ol CI CI F1
- i. VESSEL PRESSURE
.CHANGE (PSI)
- 2. VESSEL WATER LEVEL (IN) 2 OJIQ 10
- TIME, SEC 12 14 16 18 20 jll I
tel CO (D LD I
FO O CIGURE 3 10 FEEDWATER CONTROLLER FAILURE RESULT FOR 47%
POWER AND 106%
ANF-89-01 Page 25
- 4. 0 MAXIHUH OVERPRESSURIZATION Haximum system pressure has been calculated for the containment isolation event (rapid closure of all main steam isolation valves) with an adverse scenario as specified by the ASHE Pressure Vessel Code.
This analysis showed that the safety valves of WNP-2 have sufficient capacity and performance to prevent pressure from reaching the established transient pressure safety limit of 110/ of the design pressure.
The maximum system pressures predicted during the event are shown in Table
- 2. 1.
This analysis also assumed six safety relief valves out of service.
4:1 ~iB The reactor conditions used in the evaluation of the maximum pressuriza-
=tion event are those shown in Table 3. 1.
The most critical active component (scram on HSIV closure) was assumed to fail during the transient.
The calculation was performed with the ANF advanced plant simulator code COTRANSA,( ) which includes an axial one-dimensional neutronics model.
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~
4.2 Pressurization Transients ANF has evaluated several pressurization events and has determined that closure of all main steam isolation valves (HSIVs) without direct scram is the most limiting.
Since the HSIVs are closer to the reactor vessel than the turbine stop or turbine control valves, significantly less volume is available to absorb the pressurization phenomena when the HSIVs are closed than when turbine valves are closed.
The closure rate of the MSIVs is substantially slower than the turbine stop valves or turbine control valves.
The impact of this smaller volume is more important to this event than the slower closure speed of the MSIV valves relative to turbine valves.
Calculations have determined that the overall result is to cause HSIV closures to be more limiting than turbine isolations.
4.3 Closure of All Hain Steam Isolation Valves This calculation also assumed that six relief valves were out,of service and that all four main steam isolation valves were isolated at the containment boundary within 3
seconds.
At about 3.3
- seconds, the reactor scram is
ANF-89-01 Page 26
'nitiated by reaching the high flux trip setpoints.
'Pressures reach the recirculation pump trip setpoint (1170 psig) before the pressurization has been reversed.
Loss of coolant flow leads to enhanced steam production as less subcooled water is available to absorb core thermal power.
The calculated maximum pressure in the steam lines,was 1289 psig, occurring near the vessel at about 5 seconds.
The maximum vessel pressure was 1315 psig, occurring in the lower plenum at about 5 seconds.
These results are presented in Tables
- 2. 1 and 3.3 for the design basis point.
t
ANF-89-01 Page 27 5.0 RECIRCULATION FLOW RUN-UP The HCPR full flow operating limit is established through evaluation of anticipated transients at the design basis state.
Due to the potential for large reactor power increases should an uncontrolled recirculation flow increase occur from a less than rated core flow state, the need exists for an augmentation of the operating limit HCPR (full flow) for operation at lower flow conditions.
Advanced Nuclear Fuels Corporation determined the required reduced flow MCPR operating limit by evaluating a bounding slow flow increase event.
The calculations assume the event was initiated from the 104% rod line at minimum flow and terminates at 120% power at 103% flow (flow control valve wide open).
This power flow relationship bounds that calculated for a
constant xenon assumption.
It was conservatively assumed that the event was quasi-steady and a flow biased scram does not occur.
The power distribution was chosen such that the HCPR equals the safety limit at the final power/flow run-up point.
The reduced flow HCPRs were then calculated by XCOBRA(6) at discrete flow points.
The recirculation flow run-up analysis performed for WNP-2 Cycle 2
was
- reviewed, and the assumptions and conditions used for Cycle 2 are applicable to Cycle 5 except for the six degree reduction in feedwater temperature at full power conditions.
- Thus, the reduced flow MCPR operating limit for WNP-2 Cycle 5
is changed slightly from earlier cycles.
For final feedwater temperature reduction (FFTR) conditions, the previously reported(I) reduced flow HCPR operating limit remains applicable.
The reduced flow MCPR operating limit for Cycle 5 is presented in Figure 5. 1 and tabulated in Table 5. 1.
The MCPR operating limit for WNP-2 shall be the maximum of this reduced flow MCPR t
operating 1 imit and the full flow HCPR oper ating limit as summarized in Reference 2.
ANF-89-01 Page 28 TABLE 5.1 REDUCED FLOW MCPR OPERATING LIMIT FOR WNP-2 Core Flow
~%Rated 100 90 80 70 60 50 40 Reduced Flow MCPR 0 eratin Limit 1.07 1.13 1.19 1.26 1.34 1.44 1.59
1.6 1.5 NOTE: The MCPR operating limitshall be the greater of the rated, condition MCPR operating limitor the value for reduced flour from this curve 1.4 p
1.3 O
C4 1.2 20 30 40 50 60 70 80 90 100 110 TOTAL CORE RECIRCULATING-FLOW (%RATED)
FIGURE 5.1 REDUCED FLOW HCPR OPERATING LIHIT
I I
ANF-89-01 Page 30 2.
3.
4.
5.
6.
7.
8.
9.
REFERENCES J.
E.
XN-NF-87-92 and XN-NF-87-92, Supplement 1,
Advanced Nuclear Fuels Corporation,
- Richland, WA 99352, June 1987 and Hay 1988.
J.
E. Krajicek, "Supply System Nuclear Project Number 2
(WNP-2) Cycle 5
Reload Analysis,"
ANF-89-02, Advanced'uclear Fuels Corporation,
- Richland, WA 99352, January 1989.
R.
H.
- Kelley, "Exxon Nuclear Plant Transient Hethodology for Boiling II R t,"
~XII-NF-79-71 P, R
I*i 2
(
001 t d),
E Nuclear Company, Inc., Richland, WA 99352, November 1981.
H. J.
Ades and B. C. Fryer, "XCOBRA-T:
A Computer Code for BWR Transient Tt I-Ryd I I 0
A ly I,l'NN~F-84-1 AJ, V I I,
1 I Epyl t I, Vl I
Eppl t
2 d
~XII-P-
-I A, Vl Supplement 4,
Advanced Nuclear Fuels Corporation,
- Richland, WA 99352, February 1987 and July 1987.
R J.
B.
- Edgar, Letter to
- WPPSS, Supplemental Licensing Analysis
- Results, ENWP-86-0067, Exxon Nuclear
- Company, Inc
- Richland, WA
- 99352, April 15, 1986.
T.
W. Patten, "Exxon Nuclear Critical Power Hethodology for Boiling Water R,"
~XN 4A, R
I I I,
E N
I 0
p y,
- Richland, WA 99352, November 1983.
T.
L.
Krysinski and J.
.C.
- Chandler, "Exxon Nuclear Methodology for Boiling Water Reactors; THERMEX Thermal Limits Methodology; Summary 0
Ipti,"
X~II-IIF- -I A, V I 2,
R I I 2,
I N
Company, Inc., Richland, WA 99352, January 1987.
K.
R.
- Herckx, "RODEX2 Fuel Rod Mechanical
Response
Evaluation Hodel,"
~REEF-,R I
I 2,
I II I
C 0
- Richland, WA 99352, March 1984.
S.
E.
- Jensen, "Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors:
Revised Methodology for Including Code Uncertainties'n Determining Operating Limits for Rapid Pressurization Transients in PNR,"~XM-
->>,R II 2,dppl I,l.,d,EN Company, Inc., Richland, WA 99352, March 1986.
10.
"Exxon Nuclear Hethodology for Boiling Water Reactors Neutronics Methods I
0 10 A lyi*," X~NF-
-I A, Vi I, Rppi t
I d
2, Exxon Nuclear Company, Inc., Richland, WA 99352, March 1983.
J.
E. Krajicek, "WNP-2 Cycle 2 Plant Transient Analysis,"
XN-NF-85-143, Revision 1,
Exxon Nuclear
- Company, Inc.,
- Richland, WA
- 99352, February 1986.
0 I
I I
APPENDIX A HCPR SAFETY LIMIT ANF-89-01 Page A-I A.l INTRODUCTION Bundle power limits in a boiling water reactor (BWR) are determined l
through evaluation of critical heat flux phenomena.
The basic criterion used in establishing critical power ratio (CPR) limits is that at least 99.9% of the fuel rods in the core will be expected to avoid boiling transition (critical heat flux) during normal operation and anticipated operational occurrences.
Operating margins are defined by establishing a minimum margin to the onset of boiling transition condition for steady state operation and calculating a transient effects allowance, thereby assuring that the steady state limit is protected during anticipated off-normal conditions.
This appendix addresses the calculation of the minimum margin to the steady state boiling transition condition, which is implemented as the HCPR safety limit in the plant technical specifications.
The transient effects allowance, or the
'imiting transient change in CPR (i.e., delta CPR), is treated in the body of this report.
The HCPR safety limit is established through statistical consideration of measurement and calculational uncertainties associated with the thermal hydraulic state of the reactor using design basis radial,
- axial, and local power distributions.
Some of the calculational uncertainties, including those introduced by the critical power correlation, power peaking, and core coolant distribution, are fuel related.
When ANF fuel is introduced into a core where it will reside with another supplier's fuel types, the appropriate value of the HCPR safety limit is calculated based on fuel-dependent parameters associated with the mixed core.
Similarly, when an ANF-fabricated reload batch is used to replace a
group of dissimilar fuel assemblies, the core average fuel dependent parameters change because of the difference in the relative number of each type of bundle in the core, and the HCPR safety limit is again reevaluated.
ANF-89-01 Page A-2 The design basis power distribution is made up of components corresponding to representative
- radial, axial, and local'eaking factors.
Where such data are appropriately available from the previous
- cycle, these factors are determined through examination of operating data for the previous cycle and predictions of operating'onditions during the cycle being evaluated for the HCPR safety limit.
If operating data are not available, either because the reactor has not been operated or because appropriate data cannot be supplied to ANF, the 'safety limit power distribution is determined strictly from the predicted operating conditions during the cycle being evaluated.
Operating data for WNP-2 during Cycle 4 and the predicted operating conditions for Cycle 5 were evaluated to identify the design basis power distributions used in the Cycle 5, HCPR safety limit analysis.,
ANF-89-01 Page A-3 A. 2 ASSUMPTIONS A.2.1 Desi n Basis Power Distribution The local and radial power distributions which were determined to be conservative for use in the safety limit analysis are shown in Figures A-1 through A-5.
A.2.2 H draulic Demand Curve Hydraulic demand curves based on calculations with XCOBRA were used in the safety limit analysis.
The XCOBRA calculation is described in ANF topical reports XN-NF-79-59(A),
"Methodology for Calculation of Pressure Drop in BWR Fuel Assemblies,"
and XN-NF-512(A), "The XN-3 Critical Power Correlation."
A.2.3 S stem Uncertainties System measurement uncertainties are not fuel dependent.
The values reported by the NSSS supplier For these parameters remain valid for the
~
~
~
~
~
insertion of ANF fuel.
The values used in the safety limit analysis are tabulated in the topical report XN-NF-524(A),
"Exxon Nuclear Critical Power Methodology for Boiling Water Reactors."
A.2.4 Fuel Related Uncertainties Fuel related uncertainties include power measurement uncertainty and core flow distribution uncertainty.
The values used in the safety limit analysis are tabulated in the topical report XN-NF-524(A),
"Exxon Nuclear Critical Power Methodology for Boiling Water Reactors."
Power measurement uncertainties are established in the topical report XN-NF-80-19(A), Volume 1, "Exxon Nuclear Methodology for Boiling Water Reactors; Neutronics Methods for Design and Analysis."
ANF-89-01 Page. A-4 A.3 SAFETY LIMIT CALCULATION A statistical analysis for the number of fuel rods in boiling transition was performed using the methodology described in ANF topical report XN-NF-524(A),
"Exxon Nuclear Critical Power Methodology for Boiling Water Reactors."
With 500 Monte Carlo trials it was determined that 'for, a minimum CPR value of 1.06 at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition with a confidence level of 95%.
ANF-89-01 Page A-5
.936
.977
- 1.023
- 1.015
- 1.011
- 1.041
- 1.076 :. 1.052
.977
- 1.011
.907
- 1.042
- ,, 1.035
.932
.962 1.075 1.023
.907
- 1.017 4988
.974
.996
.931 1.040
~
~
1.015
- 1.042
.988
.000
.850
.972
- 1.033 1.009 1.011
- 1.035
.974
.850
.000
.985
- 1.038
- 1.011 1.041
.932
.996
".972
.985
- 1.012
.901
- 1.043 1.076
.962
.931 1.033
- 1.038
.901
.976
- 1.078 1.052
- 1.075
- 1.040
- 1.009
- 1.011
- 1.043
- 1.078 1.054 FIGURE A.1 WNP-2 CYCLE 5 SAFETY LINIT LOCAL PEAKING FACTORS (ANF-4 FUEL)
ANF-89-01 Page A-6
~
.944
.962
- 1,011
- 1.044
- 1.043
- 1.010
.960
.943
~
.962
.980 o
~
- 1.064 I
.894
- '1.033
- 1.059
- 1.034
.961
- 1.011
- 1.064
- 1.010
.994 ":
.982
- 1.002
.915
- 1.010
~
~
- 1.044
.894.:
.994
.000
.907
.980
- 1.032
- 1.042
~
~
- 1.043
- 1.033
.982
.907
.000 ':
~
.988
.952
- 1.041
~
- 1.010
- 1.059
- 1.002
~
.980
.988
- 1.004
- 1.060
- 1.065
.960
- 1.034
.915
- 1.032
.952
- 1.060
.966
- 1.053
.943
.961
- 1.010
- 1.042
- 1.041
- 1.065
- 1.053
- 1.019 FIGURE A.2 WNP-2 CYCLE 5 SAFETY LIHIT LOCAL PEAKING FACTORS (ANF XN-3 FUEL)
ANF-89-01 Page A-7
.950
.963
- 1.000
- 1.027
- 1.026
.999
.963
.950
.963
.981 1.052
.920
- 1.033 :,1.049
- 1.020
~ t
.963 1.000
- 1.052
- 1.017
- 1.005
.997':
1.011
.936
- 1.000 1.027
.920
- 1.005
.000
.935,:
.996
- 1.033 1.'027 1.026
- 1.033
.997
.935
.000
- 1.002
.971
- 1.027
.999 1.049
- 1.011
.996
- 1.002':
1.016
- 1.054
- 1.042
.963 1.020
.936
- 1.033
.971
- 1.054
.973
- 1.029
.950
.963
- 1.000
- 1.027
- 1.027
- 1.042
- 1.029
- 1.003 FIGURE A.3 WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-2 FUEL)
ANF-89-01 Page A-8
.967
.969
.997
- 1.019
- 1.019
.996
.968
.966
.969
.981
- 1.044
.932
- 1.030
- 1.042
- 1.013
.968
.997
- 1.044
- 1.017
- 1.008
- 1.001
- 1.012
.944
.997 1.019
.932
- 1.008
.000
.947
- 1.000
- 1.030
- 1.019 1.019
- 1.030
- 1.001
.947
.000
- 1.006
,.976
- 1.020
.996
- 1.042
- 1.012
- 1.000
- 1.006 :,1.017
- 1.047
- 1.032
.968 1.013
.944
- 1.030
.976
- 1.047
.975
- 1.020
.966
.968
.997
- 1.019
- 1.020
- 1.032
- 1.020
- 1.003 FIGURE A.4 WNP-2 CYCLE 5 SAFETY LIHIT LOCAL PEAKING FACTORS (ANF XN-1 FUEL)
4 ANF-89-01 Page A-9
~
1.03 1.00
.99
.99
~
.99
.99 1.00 1.03 1.00
.97
~
.99 1.02 1.03 1.03
.99 1.00
~
4 0
.99
.99 1.02 1.01
~
1.02
.91 1.03
.99
~
~
.99 1.02 1.01
.91
.00 1.02 1.02
.99
.99 1.03
~
1.02
.00 1.02 1.01 ':
99
~
.99
~
~
99
~
1.03
.91 1.02 1.01
.98
.99
.99 1.00
.99 1.03 1.02
.99
.99
.97 1.00 1.03 1.00
.99
.99
.99
.99 1.00 1.03 FIGURE A.5 WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS (GE FUEL'
100 80 b3 60 Kl
<0 20 0
0 0.25 0.50 0.75, 1.25 BUNDLE PONER FFICTOR
ANF-89-01 Issue Date:
'I/I6y8g WNP-2 CYCLE 5 PLANT TRANSIENT ANALYSIS t
Distribution:
0.
C.
M. E.
R.
E.
S. J.
S.
E.
J.
E.
J.
L.
J.
N.
L. A.
G. L.
R.
B.
H; E'.
Brown Byram Collingham Haynes Jensen Krajicek Maryott Morgan Nielsen Ritter Stout Williamson Y. U.
Fresk/WPPSS (50)
Document Control (5)