ML20154K258
| ML20154K258 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 05/23/1988 |
| From: | Krajicek J SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML17284A527 | List: |
| References | |
| XN-NF-87-92-S01, XN-NF-87-92-S1, NUDOCS 8809230268 | |
| Download: ML20154K258 (22) | |
Text
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XN-NF-87-92
' SUPPLEMENT 1
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WNP-2 PL ANT TR ANSIENT AN ALYSIS I
TEMPER ATURE REDUCTION CYCLE 4 AN ALYSIS I
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MAY 1988 lI
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ADVANCED NUCLEARFUELS CORPORATION l
XN-NF 87-92 Supplement 1 i
Issue Date: 5/23/88 I
I WNP-2 PLANT TRANSIENT ANALYSIS WITH FINAL FEEDWATER TEMPERATURE REDUCTION CYCLE 4 ANALYSIS l
cr//Qual Prepared By:
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6 - [fJ. E. Krajicek tNR Safety Analysis Licensing and Safety Engineering l
Fuel Engineering and Technical Services i
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NUCLEAR REGULATORY COMMISSION REPORT DISCLAIMER I
iMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was certved through research and development pro-grams sponsored L/ Advanced Nuclear Fuels Corporation. It is being suomit.
ted by Advanced Nuclear Fuels Corporation to the U.S. Nuclear Regulatory C,ommission as part of a techn6 cal conthbutkn to facilitate safety analyses by licensees of the U.S. Nuclear Regulatory Commissio i which uttilze Ad-venced Nuclear Fuets Corporstmfabrtcated reload fuel or other technical services provk$ed by Advanced Nuclear Fuels Corporation for light water power reactors and it is true and correct to the best of Advanced Nuclear Fuets Corporation's knowledge, Informat6on, and bel 6ef. The information con-tasned herein may be used by the U.S. Nuclear Regulatory Commiss6on in its reytow of this rooort, and under the terms of the respective agreements, by l
licensees or appl 6 cants before the U.S. Nuclear Regulatory Commiss6on l
wh6ch are customers of Advanced Nuclear Fuels Comorttlon in their i
demonstration of compthnce wtth the U.S. Nuclear Regulatory Commistion's regulations.
j Advanced Nuclear Fuels Corporation's warranties and representations con-g l coming the subject matter of this document are those set forth in the scree-I ment between Advanced Nuclear Fue's Corporation and the custorner to wh6ch this oocument is issued. Accordingty, except as otherwise expressly provided in such agreement, neither Advanced Nuclear Fuele 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 informat6on contained in this docu-ment, of that the use of any informat6on, apparatus, method, or process disclosed in this document will not intnnge privatety owned rights, '
E B. Assumes any llaedities w6th re.pect to the use of, or for W
camages resulting from the use of, any informat6on, ao.
parttus, method F pfoceSs disclosed 6n this document.
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l XN-NF-87-92 Supplement 1 TABLE OF CONTENTS Section gagg
1.0 INTRODUCTION
1 I
2.0
SUMMARY
2 3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN............................
4 3.1 Design Basis.....................................................
4 I
3.2 Anticipated Transients...........................................
5 3.2.1 Load Rejection Without 8ypass....................................
6 3.2.2 Feedwater Controller Failure.....................................
6
4.0 REFERENCES
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. XN-NF-87-92 Supplement 1 LIST OF TABLES Table Pagt 2.1 Thermal Margin Summary For Operation With FFTR...................
3 3.1 Design Reactor And Plant Conditions For Cycle 4 FFTR Oparation...
8 3.2 Significant Parameter Values Used In Analysis For WNP-2..........
9 3.3 Resul ts Of System Plant Transient Analyses.......................
12 1
l LIST OF FIGURES I
Fiaure Plan 3
3.1 Cycle 4 FFTR Load Rejection Without B 3
Normal Scram Speed...................ypass, RPT Operabl e, 13 3.2 Cycle 4 FFTR Load Rejection Without B Normal Scram Speed...................ypas s, RPT Operabl e, 1
14 3.3 Cycle 4 FFTR Feedwater Controller Failure Results For 47% Power And 106% Flow, RPT Operable. Normal Scram Speed..................
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3.4 Cycle 4 FFTR Feedwater Controller Failure Results For 47% Power And 106% Flow, RPT Operable, Normal Scram Speed..................
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1 XN NF-87-92 Supplement 1
1.0 INTRODUCTION
ANF has performed analyses supporting final feedwater temperature reduction (FFTR) for WNP-2 as reported in Reference 1.
The change in minimum critical power ratio (MCPR) resulting from the FFTR was determined for the limiting transients.
Operating MCPR limits for WNP-2 with FFTR were then proposed based on these analyses.
The FFTR analyses in Reference I were performed using neutronic input data for WNP-2 Cycle 3 and ANF transient methodology.
This supplement presents additional FFTR analysis results using Cycle 4 neutronics data for the most limiting transients.
The analyses of Reference 1 addressed potentially limiting transients as determined for jet pump BWR's in XN NF-79-71(4).
For the Cycle 4 analysis, only the most limiting rapid pressurization events, the load rejection without bypass (LRNB) and feedwater controller failure (FWCF) transients which may be sensitive to cycle specific changes, were recalculated.
The FFTR analysis was performed consistent with the WNP-2 cycle specific reload analysis (References 2
and 3) employing the same methodology (References 4 and 5).
The models were appropriately adjusted to reflect the conditions for FFTR operation.
Thus, the differences in the results between the FFTR cases and the normal feedwater cases are due only to the feedwater temperature reduction and cycle specific neutronics, l
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2 XN NF-87-92 Supplement 1 2.0
SUMMARY
The effects on critical power ratio requirements for the potentially limiting plant system transient events with a FFTR of 65'F or less are shown in Table
{
2.1 for both Cycle 3 and Cycle 4.
Based on the Cycle 3 analysis, the effect of FFTR is to increase the delta CPR for the LRNB event by up to 0.02 and decrease the delta CPR for the FWCF event by as much as 0.01.
The Cycle 4 FFTR LRNB results showed no delta CPR change for ANF fuel and a delta CPR increase of 0.01 for GE fuel relative to the Cycle 4 avents with normal feedwater temperature.
The Cycle 4 FFTR FWCF results showed a delta CPR decrease of 0.01, which is the same as shown for Cycle 3.
Thus, the Cycle 3 FFTR results bound the Cycle 4 results.
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XN NF-87-92 Supplement 1 I
TABLE 2.1 THERMAL MARGIN
SUMMARY
FOR OPERATION WITH FFTR*
Effect of FFTR**
,G.ysig
% Power /% Flow On MCPR Limit GE Fuel ANF Fuel LRNB 3
104/106 0.02 0.02 l
LRNB 4
104/136 0.01 0.00 FWCF 3
47/106
-0.01 0.01 FWCF 4
47/106
-0.01
-0.01 I
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- These transients were evaluated with normal scram speed.
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4 XN-NF-87-92 Supplement 1 I
3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN 3.1 Desian Basis I
The system transient analysis, performed to determine the effects of feedwater temperature reduction on the MCPR operating limits, assumed the following:
1.
The plant is operating on a normal 12 month cycle.
I 2.
A feedwater temperature reduction of 65'F is used to extend the operating cycle.
3.
Conservatively, the plant is assumed to operate 18 days at full thermal power with the feedivater temperature reduced prior to l
any coastdown.
4.
The plant is operating with the recirculation flow control g
valves in the wide-open position, producing 106% of rated core flow at 100% core thermal power.
5.
Control rod insertion time is based on WNP-2 measured (normal)
I scram speed.
6.
Integral power to the hot channel was increased by 10% for the pressurization transient thermal margin evaluations, consistent with Reference 6.
I These assumptions are consistent with the assumptions for the Cycle 3 FFTR analysis.
The load rejection without bypass event was performed at the 104% power /106%
I flow point with normal (as measured) scram speed.
Since feedwater controller failure transients are more severe at reduced power because of the capability l
for a larger change in feedwater flow (and a corresponding increase in core thermal power prior to a high water level trip), the FWCF translent evaluation was performed at the minimum power (47%) allowed for increased core flow (106%).
The initial conditions used in the Cycle 4 analysis at the 104%
power /106% flow point are as shown in Table 3.1.
The most limiting exposure I
5 XN-NF-87 02 Supplement 1 I
in the extended cycle is at the end of full core thermal power capability with E
FFTR when the control rods are fully withdrawn from the core.
The thermal 5
margin limits established in this evaluation conservatively apply to feedweter temperature reductions of less than 65'F and to cases where the I
control rods are partially inserted.
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The calculational models and plant operational parameters used to determine j
the thermal margin are consistent with those used in the WNP 2, Cycle 3 reload and FFTR analy31s and the Cycle 4 reload and FFTR analysis. The models include the ANF plant transient and core thermal-hydraulic codes as described g
in References 5, 6, 7, and 8.
The plant operational parameters utili.ied in 5
this evaluation are sumarized in Table 3.2.
I 3.2 Anticioated Transients ANF considers eight categories of potential system transient occurrences for det Pump BWRs in XN NF-79-71(4).
The two most thermal limiting events determined for Cycle 3(2) and Cycle 4(3) were evaluated for the FFTR operating states.
I These transients are:
I Lead Rejection Without Bypass (LRNB)
Feedwater Controller Failure (FWCF) g The WNP 2 reactor was operated in the single loop mode (one active recirculation loop and reduced power) during significant portions of Cycles 1 and 2.
This affects the neutronics for Cycle 3 and makes it typical of cycles following single loop operation.
Cycle 3 operated with both recirculation pumps active (full power), and as a result the Cycle 4 end of cycle neutronics used herein are more typical of 2 loop WNP-2 operation.
Thus, the impacts of l
FFTR as calculated for Cycle 3 and Cycle 4 represent the effect of FFTR on MCPR limits for considerably different prior plant operations.
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A summary of the Cycle 3(2) and Cycle 4(3) transient analyses including FFTR is shown in Table 3.3.
l 3.2.1 Load Re.iection Without Bvoass g
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 I
trip (RPT).
Figures 3.1 and 3.2 depict the time variance of critieni reactor and plant parameters for the load rejection without bypass (LRNB) event with normal scram speed, and with FFTR as initiated from the 104% power /106% flow state point for Cycle 4.
I Table 3.3 shows the delta CPR values for the LRNB transients, with P.PT operable, and with normal scram speed (NSS).
The effect of the FFTR is to slightly increase the severity of the 104/106 I
LRNB event.
This result is attributed to a core void distribution that causes an axial power peak higher in the core.
Axial power peaks higher in the core for FFTR conditions relative to normal feedwater temperature conditions were observed both in the Cycle 3 and Cycle 4 analyses.
The 104/106 LRN3 event is l
limiting for the extended cycle FFTR conditions, and the Cycle 3 FFTR results bound the Cycle 4 FFTR results.
I 3.2.2 figdwater Controller Failure l
Failure of the feedwater control system is postulated to lead to a maximum l
increase in feedwater flow into the vessel, resulting in a power increase prior to a high water level reactor trip.
The power increase is terminated by reactor scram, RPT, and pressure relief from the opening of the turbine bypass valves.
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7 XN NF-87-92 Supplement 1 I
Because the total change in feedwater flow is the greatest from reduced power I
conditions, the FWCF is more severe from reduced power states.
The FWCF 5
transiant event was analyzed from the lowest power (47%) allowed for increased core flow operation.
Figures 3.3 and 3.4 present the time variation of key variables ir. the Cycle 4 analyses.
The delta CPR values for the FWCF event, with RPT, normal scram speed, and with and without (Cycle 3 and Cycle 4 reload l
analysis) FFTR are shown in Table 3.3.
The heat balance at the FFTR state condition shows a slightly lower steam flow than for normal feedwater conditions.
When the FWCF event occurs, the maximum feettater flow is limited only by the feedwater pumps head / flow characteristics.
Thus, the water level in the vessel rises faster and the l
high water level trip occurs at a lower power level in the FFTR modes.
This results in a smaller transient delta CPR for FWCF transients with FFTR.
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'AN-NF-87-92 Supplement 1 i
I TABLE 3.1 DESIGN REACTOR AND PLANT CONDITIONS l
FOR CYCLE 4 FFTR OPERATION I
Reactor Thermal Power (104%)
3464 MWt Total Recirculating Flow (106%)
115.0 Hib/hr Core Channel Flow 102.1 Hib/hr I
Core Bypass Flow 12.9 Hib/hr g
Core inlet Enthalpy 520.8 Btu /lbm Vessel Pressures Steam Dome 1023.2 psia Upper Plenum 1036.1 psia Core 1042.7 psia lower Plenum 1059.0 psia Turbine Pressur1 971.9 psia l
Feedwater/ Steam Flow 13.8 Hlb /hr feedwater Enthalpy 335.2 Btu /lbm Recirculating Pump Flow (per pump) 17.26 Mlb/hr Feedwater Ternperature Reduction 65.0'F Cycle Extension s18 days
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9 XN NF-87-92 Supplement 1 I
TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN ANALYSIS FOR WNP-2 I
High Neutron Flux Trip 126.2%
Void Reactivity Feedback 10% above nominal
- Time to De energize Pilot Scram Solenoid Valves 200 msec g
Time to Sense Fast Turbine Control Valve closure 80 msee Time from High Neutron Flux Time to Control Rod Motion 290 msec Scram Insertion Times (normal)**
0.404 see to Notch 45 0.660 see to Notch 39 1.504 see to Notch 25 l
2.624 see to Notch 5 Turbine Stop Valve Stroke Time 100 msec Turbine Stop Valve Position Trip 90% open l
Turbine Control Valve Stroke Time (Total) 150 msec l
fuel / Cladding Gap Conductance 581. Btu /hr-ft2 F
- for rapid pressurization transients a 10% critiplier on integral power is used, see Reference 6 for methodology descript,on.
- \\lc^ 1st maasured average Control Rod insertion time to specified notches for each group of 4 control rods arranged in a 2x2 array.
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TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN ANALYSIS l
FOR WNP-2 (Continued)
I Safety / Relief Valve Performance l
Settings Technical Specifications Relief Valve Capaci+y 228.2 lbm/sec (1091 psig)
Pilot Operated Valve Delay / Stroke 400/100 msee MSIV Stroke Time 3.0 see MSIV Position Trip Setpoint 85% open l
Condenser Bypass Valve Performance Total Capacity 990 lbm/see Delay to Opening (80% open) 300 msec Fraction of Energy Generated in Fuel 0.965 Vessel Water level (above Separator Skirt)
High Level Trip (L8) 73 in Normal 49.5 in Low level Trip (L3) 21 in Maximum Feedwater Runout Flow I
Two Pumps 5799 lbm/sec 3
Recirculating Pump Trip Setpoint 1170 psig 5
Vessel Pressure I
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11 XN NF-87-92 Supplement 1 I
TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN ANALYSIS 3
FOR WNP-2 (Continued)
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Control Characteristics Sonsor Time Constants Steam Flow 1.0 see I
Pressure 500 msec Others 250 msec g
Feedwater Control Mode Three-Element Feedwater 100% Hismatch Water Level Error 48 in Steam Flow Equiv.
100%
Flow Control Mode Manual Pressure Regulator Settings l
Lead 3.0 sec Lag 7.0 see Gain 3.3%/psid I
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I-12 XN-NF-87-92 Supplement 1 I
TABLE 3.3 RESULTS OF SYSTEM PLANT TRANSIENT ANALYSES I
I Maximum Maximun Maximum Core Average System Delta CPR Neutron Flux Heat Flu.x Pressure GE ANF Eyant
(% Rated)
(% Rated)
(osio)
M M
LRWB, Cycle 3, 295 115 1165 0.25 0.23 RPT Operable, NSS' I
LRWB, Cycle 3 308 118 1170 0.27 0.25 RPT Operable NSS, FFTR LRWB, Cycle 4, 373 119 1170 0.25 0.24 RPT Operable, NSS 1
l L3WB, Cycle 4, 334 119 1170 0.26 0.24 RPT Operable, NSS, FFTR FWCF, Cycle 3, 156 54 1015 0.26 0.24 l 3 (47% Power /106% Flow),
(47% Power /106% Flow).
FWCF, Cycle 4, 103 50 1010 0.12 0.11 (47% Power /106% Flow),
l NSS, RPT Operable FWCF, Cycle 4, 98 50 1001 0.11 0.10 (47% Power /106% Flow),
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- 5. FEEDWATER FLOW 3
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rg Figure 3.3 Cycle 4 FFTR Feedwater Controller failure Results Fee 47% Power And 106% Flow, RPT Operable, fiormal Scram Speed
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aa figure 3.4 Cycic 4 FFTR Feedwater Controller Failure Results for 47% Power And 106% Flow, RPI Operable,fiormal Scram Speed
17 XN NF-87-92 Supplement 1
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4.0 REFERENCES
1.
J.
E. Krajicek, 'WNP-2 Plant Transient Analysis With Final Feedwater Tem)erature Reduction," XN NF 87 92, Advanced Nuclear Fuels Corporation, Ric11and, WA 99352, June 1987.
2.
J. E. Krajicek, "WNP-2 Cycle 3 Plant Transient Analysis," XN NF-87-24, Advanced Nuclear Fuels Corporation, Richland, WA 99352, March 1987.
3.
J. E. Krajicek, "WNP-2 Cycle 4 Plant Tra. lent Analysis," XN NF 88-01, Advanced Nuclear Fuels Corporation, Richland, WA 99352, January 1988.
4.
R.
H.
Kelley, "Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors," XN-NF-79-71(P), Revision 2 (as supplemented), Exxon Nuclear Company, Inc., Richland, WA 99352, February 1987.
i 5.
M.
J.
Ades, 'XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis,' XN-NF-105f A), Volume 1. Volume 1 Suppleuent 1.
I Volume 1 Supplerent 2. Advanced Nuclear Fuels Corporation Richland, WA 99352, February 1987.
I 6.
S.
E. Jensen, "Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors: Revised Methodology for Rapid Pressurization Transients in BWRs," XN NF-79-71(A), Revision 2 Supplements 1, 2, and 3, Exxon Nuclear Company, Inc., Richland, WA 99352, Mash 1986.
7.
T. W. Patten, "Exxon Nuclear Critical Power Methodology for Boiling Water Reactors,"
XN NF-524fA),
Revision 1.
Exxon Nuclear Company, Inc.,
Richland, WA 99352, November 1983.
I 8.
T.
L.
Krysinski and J.
C.
C4 ndler, "Exxon Nuclear Methodology for I
Boiling Water Reactors; THERMEX ihermal Limits Methodology; Sumary Description,"
XN NF 80-19fA),
Vclume 3
Revision 2,
Exxon Nuclear l
Company, Inc., Richland, WA 9935I, January 1987, 1I l
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XN-NF-87-92 Supplement 1 Issue Date: 5/23/88
!I 1'
WNP-2 TRANSIENT. ANALYSIS WITH FINAL FEEDWATER TEMPERATURE REDUCTION CYCLE 4 ANALYSIS Distribution:
R. E. Collingham
'I R. A. Copeland J. G. Ingham l
S. E. Jensen j
I T. H. Keheley i
D. C. Kilian J. E. Krajicek J. L. Maryott I
G. L. Ritter H. E. Williamson I
Y. U. Fresk/WPPSS (50)
Document Control (5)
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