ML17199U455
| ML17199U455 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 09/28/1987 |
| From: | Swope D SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| References | |
| ANF-87-111, NUDOCS 8803150360 | |
| Download: ML17199U455 (49) | |
Text
-,
ANF-87-111 ADVANCED NUCLEAR FUELS CORPORATION LOCA-ECCS ANALYSIS FOR DRESDEN UNITS DURING SINGLE LOOP OPERATION WITH ANF FUEL SEPTEMBER 1987 AN AFFILIATE OF KRAFTWERK UNION
. 8803150366 880309
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ADVANCED NUCLEAR FUELS CORPORATION ANF-87-111 Issue Date: 9/28/87 LOCA-ECCS ANALYSIS FOR DRESDEN UNITS DURING SINGLE LOOP OPERATION WITH ANF FUEL Prepared By:
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D. R. Swope BWR Safety.Analysis Licensing and Safety Engineering Fuel Engineering and Technical Services AN AFFILIATE OF KRAFTWERK UNION g1ewu
CUSTOMER DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY Advanced Nuclear Fuels Corporation's warranties and representations con-cerning the subject matter of this document are those set forth in the Agreement between Advanced Nuclear Fuets Corporation and the Customer pursuant to which this document is issued. Accordingly, except as otherwise expressly pro-vided in such Agreement, neither Advanced Nuclear Fuels Corporation nor any person acting on its behalf makes any warranty or representation, expressed or implied, with respect to.the accuracy, completeness, or usefulness of the infor*
" matlon contained in this document, or that the use of any information, apparatus, method or process disclosed in this document will not infringe privately owned rlgh1a; or assumes any iiabllitles with respect to the use of any information, ap-paratus,. method or process disclosed in this document, The information contained herein is tor the sole use of Customer.
In order to 8YOld impairment of rights of Advanced Nuclear Fuels Corporation in patents or inventions which may be included In the information contained in this document. the *recipient, by its acceplanC8 of this document, agrees not to publish or make public use (in the patent use of the term) of such information until so authorized In writing by Advanced Nuclear Fuels Corporation or until after six (6) months following termination or expiration of the aforesaid Agreement and any extension thereof, unless otherwise expressly provided in the Agreement. No rights or licenses in or to any patents are implied by !he fumishing of this docu-ment.
XN-NF*FOCJ.765 (1/87)
- i -
ANF-87-111 TABLE OF CONTENTS Section Page
1.0 INTRODUCTION
1 2. 0
SUMMARY
2
3 3.1 LOCA During Single Loop Operation.................................
3 3.2 EXEM/BWR Application To Dresden Units *............................
3
- 4. 0 ANALYSIS RESULTS *.....**..**.*..****...*...*****.....**.**........
7
- 5. 0....... REFERENCES **..****.*****************.**...***********...******....
9
~
- ii -
ANF-87-111 LIST OF TABLES Table Page 3.1 Dresden Reactor System Data....................................... 1-0 3.2 Comparison Of ANF MAPLH~R's On Equivalent Planar Power Basis....... Jl 4.1 SLO LOCA Limiting Break Event Times...*........................... 12 4.2 MAPLHGR Results Under SLO Conditions - 9x9 Fuel................... 13 ure
.1 3.2 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 4.10 4.11 4.12 4.13 LIST OF FIGURES Page Dresden 2/3 Power-Flow Map SLO Conditions......................... 14 System Slowdown Nodal ization For BWR/3............................
15 Slowdown System Pressure... *************~*********..... ****~*..... 16 Slowdown Total Break Flow.......................................... 17 Blowdown Average Core Inlet Flow................ ~***************** 18 Blowdown Average Core Outlet Flow................................. 19 Blowdown Hot Channel Inlet Flow................................... 20 Blowdown Hot Channel Midplane Flow................................ 21 Blowdown Hot Channel Outlet Flow.............................. ~... 22 Blowdown Intact Loop Jet Pump Slowdown Intact Loop Jet Pump Slowdown Broken Loop Jet Pump Slowdown Broken Loop Jet Pump Slowdown Broken Loop Jet Pump Suction Fl ow........................
Exit Fl ow............................
Drive Fl ow..........................
Suction Fl ow........................
Exit Fl ow...........................
23 24 25 26 27 Blowdown Upper Downcomer Mixture Level........................... ~
28
Figure 4*.14 4.15 4.16 4.17 4.18 4.19 4.20
- i i i -
LIST OF FIGURES (Continued)
ANF-87.
Page Slowdown Middle Downcomer Mixture Level........................... 29 Slowdown Lower Downcomer Mixture Level............................ 30 Slowdown Lower Downcomer Liquid Mass............................ ~.
31 Slowdown Upper Plenum Liquid Mass................................. 32 Slowdown Upper Downcomer Liquid Mass.............................. 33 Slowdown Lower Plenum Liquid Mass................................. 34 Refill/Reflood System Pressure.. :................................. 35
- 4. 21 Ref i 11 /Re flood Lower Pl en um Mixture Leve 1.........................
36 4.22 Refill/Reflood Relative Core Midplane Entrainment.................
4.23
- Slowdown Hot Channel Center Volume Coolant Temperature............
4.24 Slowdown Hot Channel Heat Transfer Coefficient................*...
4.25 Slowdown Hot Channel Center Volume Quality........................
4.26 Typical Hot A~sembly Heatup Results For 9x9 Fuel.................. -
39
- .I 40 41
- iv -
ANF-87-111
- ACKNOWLEDGEMENT The assi st.ance of Chuck Hendrix of Intermountai n Technologies, Inc. ( ITI) in preparing this analysis is acknowledged.
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1 ANF-87-111
1.0 INTRODUCTION
The results of a LOCA-ECCS analysis for the Dresden 2 and 3 nuclear power plants during single loop operation (SLO} are reported in this document.
These calculations were performed with the generically approved Exxon Nuclear Company EXEM/BWR Evaluation Model (Ref. 1, 2} in accordance with Appendix K of 10 CFR 50 (Ref. 3), and the results comply with the U.S. NRC 10 CFR 50.46 criteria.
The initial operating condition selected for this analysis was 81.3%
power/58.0% core flow.
The analysis was performed using the Dresden Unit 3 conditions; however, the Dresden 2 and 3 Units are sufficiently similar so that the results of this analysis are applicable to both Units.
~is analysis establishes the multiplier that is to be applied to the
~APLHGR's of the ANF fuel during single loop operation.
the analysis was performed with 9x9 fuel to determine this multiplier and justification is provided for its applicability to both ANF 8x8 and 9x9 fuel.
2 ANF-87-111 2.0
SUMMARY
The results of the ECCS analysis presented herein support the use of a 0.91 SLO multiplier on the two-loop MAPLHGR's for ANF fuels when the Dresden Units are operating in single loop operation.
All calculations were performed with the NRC approved EXEM/BWR ECCS Evaluation Model, according to Appendix K of 10 CFR 50.
Single loop operation of the Dresden Units with a multiplier of 0.91 on the two-loop ANF fuel MAPLHGR's assures that the emergency core cooling systems for the Dresden Units will meet the U.S. NRC acceptance criteria for loss-of-coolant accident breaks up to and including the double-ended severance of a reactor coolant pipe; That is:
- 1.
The ca 1 cul ated peak fue 1 e 1 ement c 1 ad temperature does not exceed the 2200 F limit.
- 2.
The calculated total oxidation of the cladding nowhere exceeds 17%
times the total cladding thickness before oxidatiqn. *
- 3.
The calculated maximum hydrogen generation does not exceed 1% of the zircaloy associated with the active fuel cladding in the reactor.
- 4.
The LOCA cladding temperature transtent is calculated to be terminated at a time when the core is still amenable to cooling.
- 5.
The system long-term cooling capabilities provided for the initial core and subsequent reloads remain applicable* to ANF fuel.
3 ANF-87-111 3.0 JET PUMP BWR ECCS EVALUATION MODEL 3.1 LOCA During Single Loop Operation Response to a loss-of-coolant accident (LOCA) is described in the previous analysis for ~wo-loop operation (Ref. 4).
This report addresses the instance of a LOCA during single loop operation (SLO).
During SLO the recirculation pump in the inactive loop is not operating and is isolated from the remainder of the system by a valve.
Thus, there is no flow through the idle, intact loop.
A break might occur in either loop.
- However, a break in the inactive loop would behave essentially like a break during two-
~
. loop operation except that substantial break fl ow would come from only one side of the break (because of the closed valve in the loop).
System rformance would then be like that resulting from a somewhat sma 11 er break during two-loop operation.
This scenario is already covered by the previous LOCA break spectrum analysis(S). Further consideration in this report will be given only to the case where a break occurs in the active 1 oop.
This case differs from the two-loop case in one important respect:
there is no coastdown flow in th~ intact loop.
A previous SLO analysis (Ref. 6) assumed that the consequence of a lack of coastdown flow (which continues to supply liquid from the downcomer to the lower plenum, although at a decreasing rate, during two-loop operation) would be an almost immediate flpw stagnation in the core and a very early CHF (O.l sec); this resulted in poor heat transfer very early in the transient and required that a MAPLHGR reductirin factor of 0.84 be imposed for SLO conditions on the MAPLHGR for GE 8x8 fuel.
EXEM/BWR Application To Dresden Units EXEM/BWR ECCS Evaluation Model codes(l} were used for this SLO LOCA-ECCS calculation.
The codes which comprise the Evaluation Model consist of
. -- - -*-*-*-* :.: -* =-- -
4 ANF-87
- RODEx2(7), RELAx(8), FLEx(9), and HUXY/BULGEx(l0,11).
The latest versiohs of these codes were used for this analysis.
The only significant geometric modeling difference necessary was to simulate SLO in the RELAX model of the intact loop.
The isolation valve in the intact loop is closed.
The approved ANF generic break spectrum analysis(5) for the BWR/3 reactors of which the Dresden Units are typical, identified the limiting break as the double-ended guillotine break of the recirculation pump suction pipe with a discharge coefficient of 1.0 (1.0 DEG/PS).
This limiting break (including worst single failure) was used in this analysis.
If LOCA occurs during single loop operation, a rapid drop in ~ore flow will
- occur during the early phase because the intact loop pump is not operating. -
The core flow transient during the early phase (0 - 5 seconds) of a LO during single loop is the principal event which will distinguish such accident from a LOCA occurring during normal two-loop operation.
As the break size increases, the magnitude of the initial drop in core flow also would increase, and the critical heat flux (CHF) will occur earlier in time due to the decreased *core fl ow.
This trend of a larger drop in core fl ow with increasing break size has been confirmed by ANF in sensitivity analyses on similar plants.
Thus, the worst case LOCA for single loop operation would be expected t6 be with i break at least as large in area as the limiting break for two-loop operation.
Since the 1 imiting break for normal (two-loop) operation is the break of the largest possible area, this break will also be the limiting one for single 1 oop operation.
The NSSS vendor a 1 so determined that this same break is limiting for two a*nd single loop operation(6).
The 81.3% power/58% core flow (81.3/58) operati_ng point was selected as t-initial operating condition for this analysis.
This condition lies on t.
APRM single pump rod block 1 ine on the current operating power* flow map* for the Dresden Units.
This value was obtained by using the maximum core flow
5 ANF-87-111 under SLO conditions of 58% and then selecting the appropriate power from the power-flow map.
This condition was determined in Reference 6 to be a conservative upper bound. for single loop operation; Figure 3.1 shows the power-flow map and the selected single loop operating point.
The plant conditions used for the analysis are presented in Table 3.1.
This data app 1 i es to both Dresden Un it 2 and Unit 3 because of the similarity between the two pl. ants.
The system behavior duri.ng a LOCA is determined primarily by the LOCA. break parameters:
break 1 ocat ion, break size, and. break configuration, together with the system components and geometry.
Variations in core parameters produce only secondary effects on the system behavior.
- Thus, by,:-. using
.....
- bounding core neutroni c parameters, the *LOCA-ECCS results established by this analysis will apply for future cycles unless significant changes are made in e plant operating conditions, plant hardwa~e, or core design such that the nalysis no longer bounds the plant conditions.
The blowdown calculations made for this analysis differ from cthose for the previous two-loop analysis(4) in two important ways.
An additional v*a*-lve is included upstream of the pump in the intact loop and initial conditions are al~ered to correspond -to the max.imum power/flow conditions for SLO.
The nodalization is shown in Figure 3.2.
Consistent with plant procedures *for SLO, this is closed during the entire calculati-0n preventing any flow through the loop.
The average core blowdown calculations are followed by hot channel calculations.
The hot chann.el geometry is identical with that used in the two-loop analysis (Ref. 4). A bounding radial peaking factor of 1.76 was used to set the assembly power for the hot channel analysis.
he nodalization and geometry used in the reflood calculation are also identical to those of the two-loop analysis(4).
In the FLEX code the intact loop is not modeled in detail because intact loop flows are. insignificant: by
6 ANF-8 the time of rated spray; thus no changes were. required in the FLEX nodalization or geometry.
The initial conditions for the reflood calculation are entirely determined by the blowdown calculation.
The HUXY/BULGEX heatup calculation of the hot plane was done as in the*
previous two-loop analyses(4):
fuel stored energy, thermal gap conductivity and dimensions from RODEX2 as a function of power and exposure; time of.rated spray, decay power, heat transfer coeffi ci en ts and cool ant conditions from RELAX; and time of hot-node-reflood from FLEX.
Appendix K spray heat transfer coefficients are used for the spray cooling period.
Peak clad temperature (PCT) and the cladding oxidation percentage are determined for ANF 9x9 fuel.
Table 3.2 shows the ANF 8x8.and 9x9 MAPLHGR's for.the Dresden Units on the equivalent basis of bundle planar power {MAPLHGR times number of heated rods).
As shown in this table; the 9x9 MAPLHGR's at lower exposures are significa higher (on an equivalent bundle planar power basis) than 8x8 MAPLHGR's; a GWd/MTU the equivalent 9x9 MAPLHGR is about 13% higher than the equivalent 8x8
.MAPLHGR limit.
In preliminary SLO analyses it _was determined that the high planar power of the 9x9 MAPLHGR at 5 GWd/MTU was sufficient to cause an early CHF; thus a reduction in the 9x9 -MAPLHGR was
- Preliminary analyses show that a reduction in the peak 9x9 MAPLHGR value by 1 kW/ft (9%) to 10.75 kW/ft was sufficient to significantly delay CHF.
7 ANF-87-111 4.0.
ANALYSIS RESULTS Analyses were performed for the assembly exposure (0-20 GWd/MTM) for which the two-loop MAPLHGR's had the highest PCT's (see Table 3.2).
In all cases the MAPLHGR's justified in the single loop analysis were within 9% of the two-loop MAPLHGR's and the PCT's for the most limiting exposures (0-10 GWd/MTU) were not only less than acceptance criteria of 2200°F, but less than the PCT's from the two-,loop analyses.
Thus, a.91 multiplier on the two-loop 9x9 MAPLHGR curve assures that the 10 CFR 50.46 criteria are met under SLO conditions.
Calculated event time results'*are given in Table 4.1.
Syste*m blowdown results are presented in Figures 4.1 through 4.19.
System refill and reflood results if, are given. in Figures 4;20 through 4.22.
These system conditions are used as t*
boundary conditions for a series of exposure dependent maximum power assembly atup calculations.
Results.from a RELAX/HOT CHANNEL calculation are given
. in Figures 4*.23 through 4.25 for 9x9 fuel at MAPLHGR values of 10.75 kW/ft.
Resulting clad temperatures.as* calculated by HUXY/BULGEX are shown in Figure 4.26.
~'
An ~xamination of these plots;reveals.the following information:
- 1.
The sudden loss o'f' df'ive fluid in the jet pumps allowed a. sudden drop in lower plenum pressure of sufficient magnitude to allow flow through the inactive jet pumps. to "reverse" from their initial negative. flow to a positive flow in the earlier part of the blowdown.
- 2.
The lack of a pump coasting down in the intact loop a 11 owed fl ow through the suction and exit junctions of the operating (broken loop) jet pumps to remain in the positive direction (the drive, of course, reversed to supply fluid to the break) during the earlier part of the blowdown.
T"
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8 ANF-8
- 3.
There is an immediate plunge in core flow accompanied by a reduction in the heat transfer coefficient.
- 4.
- The plunge in core flow is not sufficient to cause CHF* to immediately occur at the high powered plane for the MAPLHGR's analyzed.
The flow in the high power planes continue in the upward direction as shown in* Figure 4.6.
CHF is delayed until about 9 seconds. This is because, while the hot channel inlet flow actually reverses, the lower part of the channel continues to supply a two phase liquid-vapor mixture to the high powered hot regions of the channel.
The peak MAPLHGR analyzed for the 9x9 fuel under SLO conditions (10. 75 kW/ft) is about 5% higher than the two-loop ANF 8x8 MAPLHGR on an equivaient pla power basis; thus these analyses demonstrate that early CHF would not expected to occur in the ANF 8x8 fuel ; thus application of the same SLO multiplier (.91) to the two-loop ANF fuel MAPLHGR will conservatively protect the ANF 8x8 fuel from exceeding *the 10 CFR 50.46 criteria during SLO conditions.
Use of this multiplier (9% reduction) and the lower equivalent power of the 8x8 MAPLHGR ( 5%f result in an equivalent ANF 8x8 MAPLHGR for SLO conditions 14% lower than the peak bundle planar power analyzed in this report.
This application of a.91 multiplier during SLO conditions to all ANF fuel MAPLHGR's in the Technical Specifications will protect ANF fuel in the Dresden Units *from exceeding 10 CFR 50.46 criteria.
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9 ANF-87-111
- 5. 0 REFERENCES
- 1.
"Exxon Nuclear Methodology for Boiling Water Reactors, Volume 2,
EXEM:
ECCS Evaluation Model Summary Description," XN-NF-80-19(P), Volume 2, Revision 1, Exxon Nuclear Company, June 1981.
- 2.
"Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," XN-NF-82-07(Al~ Revision 1, Exxon Nuclear Company, November 1982.
3..
"Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors," 10 CFR 50.46 and 10 CFR 50 Appendix K, Federal Register, Volume 39, Number 3, January 4, 1974.
- 4.
"Dresden Unit 3 LOCA-ECCS Analysis MAPLHGR Results for 9x9 Fuel," XN-NF-85-63(P), Exxon Nuclear Company, September 1981.
- 5.
"Generic Jet Pump BWR 3 LOCA-ECCS Analysis Using the ENC EXEM Evaluation Model," XN-NF-81-ll(P)(A) and Supplement l(P)(A), Exxon Nuclear Company, September 1982.
"Dresden Nuclear Power Stations Units 2 and 3 and Quad Cities Nuclear Power Station Units 1 and 2, Single Loop Operation," NED0-24807, General Electric, November 1980.
- 7.
0.RODEX2:
Fuel Rod Thermal-Mechanical Response Evaluation Model," XN-NF-81-58(A), Revision 2, Exxon Nuclear Company, February 1983.~
- 8.
"RELAX:
A RELAP4-Based Computer Code for Calculating Bl_.owdown Phenomena," XN-NF-80-19(P), Volume 2A, Revision 1, Exxon Nuclear COmpany, June 1981.
- 9.
"FLEX:
A Computer Code for the-Refill and Reflood Period of a LOCA," XN-NF-80-19(P), Volume 2B, Revision 1, Exxon Nuclear Company, June 1981.
- 10.
"HUXY:
A Generalized Multi rod Heatup Code with 10 CFR 50 Appendix K Heatup Option - User's Manual," XN-CC-33(A), Revision 1, Exxon Nuclear Company, November 1, 1975.
- 11.
"BULGEX:
A Computer Code to Determine the Deformation and the Onset of Bulging of Zircaloy Fuel Rod Cladding," XN-74-21, Revision 2, and XN-NF-27, Revision 2, Exxon Nuclear Company, December 31, 1974.
10 ANF-87 TABLE 3.1 DRESDEN REACTOR SYSTEM DATA*
Primary Heat Output, MW 2095.54 Total Reactor System Volume, ft**3
- Total Reactor Flow Rate, lb/hr Active Core Flow Rate, lb/hr Nominal Reactor System Pressure (upper plenum) psia Core Inlet Enthalpy, Btu/lb Recirculation Loop Flow Rate, lb/hr Steam Flow Rate, lb/hr Feedwater Flow Rate, *lb/hr Rated *Recirculation Pump Head, ft Rated Recirculation Pump Speed, rpm Moment of Inertia, lbm-ft**2/rad Recirculation.Suction Pipe I.D., 'in.
Recirculation Discharge Pipe I.D., in.
- 9x9 Fuel Assembly Rod Diameter, in 9x9 Fuel Assembly Rod Pitch, in 9x9 Active Core Height, in
- 81.3% of rated power/58% of rated core flow.
20160.0
- 56. 84 x.10**6 51.36 X*.10**6 995.46 506.87 15.83 x 10**6 7.92 x 10**6 7.91.X 10**6 570. 00*
1,670.00 10,950.00 25.59 25.59 0.424 0.572 145.24
11 TABLE 3.2. COMPARISON OF ANF MAPLHGR'S ON EQUIVALENT PLANAR POWER BASIS ANF-87-111 Bundle Average PCT MAPLHGR Equivalent Planar Exposure (oF)
(kW/ft)
Power (kW/ft)
(GWd/MTU) 8x8 9x9 8x8 9x9 8x8 9x9 0
1879 2006 13.0 11.4 819 900 5
1942 2045 13.0
- 11. 75 819 928 10 2123 1893 13.0 11.4 819 900 15 2159 1805 12.85 10.55 810 833 20 2074 1710 12.6 9.7 794 766 25 2011 1623 11.95 8.85 753 649 30 1895 1529
- 11. 2 8.00 706 632 35 1808' 1421 10.45 7.15 658 565 40.*
1309 6.30 498
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12 ANF-8 TABLE 4.1 SLO LOCA LIMITING BREAK EVENT TIMES Event Time (sec)
Start 0.00 Initiate Break 0.05 Feedwater Flow Stops 0.55 Steam Flow Stops 5.05 Low-Low Mixture Level 5.16 Jet Pumps Uncover 8.38 Recirculation Pipe Uncovers 12.05 Lower Plenum Flashes 14.16 HPCI' Flow Starts 15.16 LPCS Starts 57.06 Rated Spray Calculated 60.70 Depressurization Ends 114. 05 Start of Reflood 152.22 Time of Hot Node Re flood 163.05.
Peak Clad Temperature Reached 163.05
13 TABLE 4.2 MAPLHGR RESULTS UNDER SLO CONDITIONS 9X9 FUEL Single Loop Conditions Assembly Average Local Burn up MAPLHGR MWR (GWd/MTU)
(kW/ft)
_ru_
0 10.75 1.1 5
10.75 1.0 10 10.75 0.9 15 10.55 0.8 20 9.70 0.6 ANF-87-111
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Figure 3.1 Dresden 2/3*Power-Flow Map SLO Conditions 90 100 110
15 ANF-87-111 ADS 38 37
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Slowdown Bro Loop Jet Pump Suction Flow
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Figure 4.13 Blowdown r Downcomer Mixture Level 80 N
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Figure 4.15 Blowdown r Downcomer Mixture Level 80 w
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Figure 4.16 Blowdown Lower Downcomer Liquid Mass BO w -
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Figure 4.24
.Blowdown Hot Channel Heat Transfer Coefficient 80
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Blowdown Ho-nnel Center Volume Quality
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i 2000 1500 1000 so 1 2 J 4 5 6 7 8 9 2 1011121J14151617 J 1118192021 222324 4 12 1 9 25 26 27 28 29 JO 5 1J20 26 Jl J2 JJ J4 JS 6 1 4 21 27 J2 J6 J7 38 J9 7 15 22 28 JJ J7 40 41 42 8 162J29J4J841 4344 9 17 24 JO JS J9 42 44 45 0
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Figure 4.26 Typical Hot Assembly Heatup Results For 9x9 Fuel P1N26 P1N12 CANlSTER 280
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ANF-87-111 Issue Date: 9/28/87 LOCA-ECCS ANALYSIS FOR DRESDEN UNITS DURING SINGLE LOOP OPERATION WITH ANF FUEL Distribution D. A.. Adkisson D. J. Braun R. E. Collingham T. H. Keheley T. L. Krysinski J. L. Maryott
- J. N. Morgan D. F. Richey G. L. Ritter D. R. Swope C. J. Volmer J. A. White H. E. Williamson CECo/J. M. Ross (60)
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