ML19337A822
| ML19337A822 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 07/31/1980 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19337A819 | List: |
| References | |
| TASK-15-19, TASK-RR 80NED281, NEDO-24085-1, NEDO-24085-1-R1, NUDOCS 8009300307 | |
| Download: ML19337A822 (32) | |
Text
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NEDO-24085-1 80NED281 Class I 1
July 1980 i
Revision 1 LOSS-OF-COOLANT ACCIDENT ANALYSIS REPORT FOR MILLSTONE UNIT 1 NUCLEAR POWER STATION NUCLEAR POWE.9 SYSTEMS OtVISION e GENERAL ELECTRIC COMPANY SAN JOSE. CALIFORNIA 96125 ggf3 ppg 7 GENER AL $ ELECTRIC
NED0-24085-1 1M*0RTANT NOTICE REGARDING CONTENTS OF THIS REM RT Please Read Carefully The only undertakings of General Electric Company respecting information in this document are contained in the contract between Northeast Utilities Company and General Electric Company and nothing contained in this document shali be construed as changing the contract. The use of this infomation by anyone other than Northeast Utilities Company or for any purpose other than that for whicit it is intended, is not authorized; and with respect to any unauthorized use, General Electric Company makes no representation or u1rranty and assumes no liability as to the completences, accuracy, or usefulness of the information contained in this document.
o 11
NED0-24085-1 CONTENTS Page 1.
INTRODUCTION 1-1 2.-
INPUT TO ANALYSIS 2-1 3.
'LOCA ANALYSIS. COMPUTER CODES AND RESULTS 3-1 3.1 Results of the LAMB Analysis 3-1 3.2 Results of the SCAT Analysis 3-1 3.3 Results of the SAFE Analysis 3-1 3.4 Results of the REFLOOD Analysis 3-2 3.5 Results of the CHASTE Analysis 3-3 3.6 Methods 3-3 4.
DESCRIPTION OF MODEL AND INPUT CHANGES 4-1 5.
CONCLUSIONS 5-1 6.
REFERENCES 6-1 f
4 i
ti/tv
NED0-24085-1 TABLES Table Title Pay 1
Significant Input Parameters to the Loss-of-Coolant Accident Analysis 2-1 2
Summary of Break Spectrum Results 3-5 3
LOCA Analysis Figure Summary 3-6 4A MAPLHGR Versus Average Planar Exposure: P8DRB282 3-7 4B MAPLHCR Versus Average Planar Exposure: P8DRB265H 3-7 4C MAPLHGR Versus Average Planar Exposure: 8DRB265H 3-8 4D MAPLHCR Versus Average Planar Exposure: 8DRB265L 3-8 t
4E MAPLHGR Versus Average Planar Exposure: 8DB274H 3-9 4F MAPLHCR Versus Average Planar Exposure: 8DB274L 3-9 4G MAPLHGR Versus Average Planar Exposure: 8DB262 3-10 v/vi i
NEDO-24085-1 ILLUSTRATIONS Figure Title Page 1
Water Level Inside the Shroud and Reactor Vessel Pressure Following a Maximum Recirculation Line Suction Break, LPCI Injection Valve Failure, Break Area = 4.34 f t2 (LBM) 5-3 2
Peak Cladding Temperature Following a Maximum Recirculation Line Suction Break, LPCI Injection Valve Failure, Break Area = 4.34 ft2 (LBM) 5-4 3
Fuel Rod Convective Heat Transfer Coefficient During Blowdpwn at the High Power Axial Node for a Maximum 2
Recirculation Line Suction Break (4.34 f t )
5-5 4
Normalized Core Average Inlet Flow Following a Maximum Recirculation Line Suction Break (4.34 ft2) 5-6 5
Minimum Critical Power Ratio Following a Maximum Recirculation Line Suction Break (4.34 ft'-)
5-7 6
Variation with Break Area of Time for thich Hot Node Remains Uncovered 5-8 7
Peak Cladding Temperature Versus Break Area 5-9 8
Water Level Incide the Shroud and Reactor Vessel Pressure Following a Small Recirculation Line Discharge Break, Gas Turbine Failure, Break Area = 0.10 ft2 (SBM) 5-10 9
Peak Cladding Temperature and Convective Heat Transfer Coefficient Following a 0.10 ft2 Recirculation Line Discharge Break, Gas Turbine Failure (sBM) 5-11 viircili
NEDO-24085-1 1.
INTRODUCTION The purpose of this document in to provide the results of the loss-of-coolant accident (LOCA) analysis for the Mi.'.lstone Unit 1 Nuclea-Power Station (Millstone-1), with a partial core loading of reload fuel with holes drilled in the lower tieplates, and considering the effects of a fourth automatic depressurization system (ADS) valve and failure of the gas turbine. The analysis was performed using approved General Electric (GE) calculational models.
This reanalysis of the plant LOCA is provided in accordance with the NRC requirement (Reference 1) and to demonstrate conformance with the ECCS acceptance criteria of 10CFR50.46. The objective of the LOCA analysis con-tained herein is to provide assurance that the most limiting break size, break location, and single failure combination has been considered for the plant.
The required documentation for demonstrating that these objectives have been satisfied is given in Reference 2.
The documentation contained in this report is intended to satisfy these requirements.
The general description of the LOCA evaluation models is contained in Reference 3.
Approved model changes (Reference 4) are described in References 5 and 6.
These model changes are employed in the REFLOOD and CHASTE computer codes which have been used in this analysis.
In addition, a model which takes into account the effects of drilling alternate flow path holes in the lower tieplate of the fuel bundle and the use of such fuel bundles in a full or partial core loading is described in References 7, 8 and 9.
This model was also approved in Reference 4.
Specific input changes as applied to Millstone-1 for this analysis are discussed in more detail in later sections of this document.
The existence of a gas turbine emergency power supply instead of a second station battery makes Millstone-1 unique with regard to single-failure con-siderations in the LOCA analysis. Accordingly, the entire recirculation line break spectrum (both large and small breaks) has been analyzed, and reference to a lead plant is not necessary.
1-1/1-2
NEDO-24085-1 2.
INPUT TO ANALYSIS A list of the significant plant input parameters to the LOCA analysis is presented in Table 1.
Table 1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS Plant Parameters:
Core Thermal Power 2051 MWe, which corresponds to 102% of rated core power 6
Vessel Steam Output 8.15 x 10 lbm/h, which corresponds to 102% of rated core power Vessel Steam P'~a Pressure 1050 psia Recirculation Line.,
Area 4.34 ft2 (DBA) for Large Breaks - Suction Number of Drilled Bundles 316 Fuel Parameters:
Peak Technical Initial Specification Design Minimum Linear Heat Axial Critical Fuel Bundle Generation Rate Peaking Power l
Fuel Type Geometry (kW/ft)
Factor Ratio
- A.
P8DRB282 8x8 13.4 1.57 1.24 B.
P8DRB265H 8x8 13.4 1.57 1.24 C.
8DRB265H 8x8 13.4 1.57 1.24 D.
8DRB265L 8x8 13.4 1.57 1.24 E.
8DB262 8x8 13.4 1.57 1.24 F.
8DB274H 8x8 13.4 1.57 1.24 G.
8DB274L 8x8 13.4 1.57 1.24 i
- To account for the 2% uncertainty in bundle power required by Appendix K, the SCAT calculation is performed with an MCPR of 1.22 (i.e.,1.24 divided by 1.02) for a bundle with an initial MCPR of 1.24.
2-1/2-2
NEDO-24085-1 3.
LOCA ANALYSIS COMPUTER CODES AND RESULTS 3.1 RESULTS OF THE LAMB ANALYSIS This code is used to analyze the short-term blowdown phenomena for large postulated pipe breaks (breaks in which nucleate boiling is lost before the water level drops and uncovers the active fuel) in jet pump reactors.
The LAMB output (core flow as a function of time) is input to the SCAT code for calculation of blowdown heat transfer.
The LAMB results presented are:
Core Average Inlet Flow Rate (normalized to unity at the beginning e
of the accident) following a Large Break.
3.2 RESULTS OF THE SCAT ANALYSIS This code completes the transient short-term thermal-hydraulic calculation for large breaks in jet pump reactors. The CEXL correlation is used to track the boiling transition in time and location. The post-critical heat flux heat transfer correlations are built into SCAT which calculates heat transfer coefficients for input to the core heatup code, CHASTE.
The SCAT results presented are:
Minimum Critical Power Ratio Following a Large Break e
e Convective Heat Transfer Coefficient following a Large Break.
3.3 RESULTS OF THE SAFE ANALYSIS This code is used primarily to track the vessel inventory and to model ECCS performance during the LOCA. The application of SAFE is identical for all break sizes. The code is used during the entire course of,the postulated accident, but after ECCS initiation, SAFE is used only to calculate reactor system pressure and ECCS flows, which are pressure dependent.
B 3-1
NED0-24085-1 The SAFE results presented are:
e Water Level Inside the Shroud (up to the time REFLOOD initiates) and Reactor Vessel Pressure 3.4 RESULTS OF THE REFLOOD ANALYSIS This code is used'across the break spectrum to calculate the system inventories after ECCS actuation. The models used for the design basis accidesc (DBA) l application (".DBA-REFLOOD") was described in a supplement to the SAFE code j
i description transmitted to the USNRC December 20, 1974. The "non-DBA REFLOOD" analysis is nearly identical to the DBA version and employs the same major assumptions. The only differences stem from the fact that the core may be l
partially covered with coolant at the time of ECCS initiation and coolant l.
levels change slowly for smaller breaks by comparison with the DBA. More l
precise modeling of coolant level behavior is thus requested principally to determine the contribution of vaporization in the fuel assemblies to the counter current flow limiting (CCFL) phenomenon at the upper tieplate. The differences from the DBA-REFLOOD analysis are:
(1) The non-DBA version calculates core water level more precisely than the DBA version in which greater precision is not necessary.
(2) The non-DBA version includes a heatup model similar to but less detailed than that in CHASTE, designed to calculate cladding tem-perature during the small break. This heatup model is used in calculating vaporization for the CCFL correlation, in calculating swollen level in the core, and in calculating the peak cladding temperature.
The REFLOOD'results presented are:
e Water Level Inside the Shroud e
Peak Cladding Temperature and Heat Transfer Coefficient for breaks calculated with small break methods 3-2
NED0-24085-1 3.5 RESULTS OF THE CHASTE ANALYSIS This code is used, with suitable inputs from the other codes, to calculate the fuel cladding heatup rate, peak cladding temperature, peak local cladding oxidation, and core-wide metal-water reaction for large breaks. The detailed fuel model in LHASTE considers transient gap conductance, clad swelling and rupture, and metal-water reaction. The empirical core spray heat transfer and channel wetting correlations are built into CHASTE, which solves the e
transient heat transfer equations for the entire LOCA transient at a single axial plane in a single fuel assembly.
Iterative applicaticns of CHASTE deter-mine the maximum permissible planar power where required to sati. fy the require-ments of 10CFR50.46 acceptance criteria.
The CHASTE results presented are:
e Peak Cladding Temperature versus time e
Peak Cladding Temperature versus Break Area Peak Cladding Temperature and Peak Local Oxidation versus Planar e
Average Exposur'e for the most limiting break size o
Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) versus Planar Average Exposure for the most limiting break size A summary of the analytical results is given in Table 2.
Table 3 liscs the figures provided for this analysis. The MAPLHGR values for each fuel type for Millstone-l are presented in Tables 4A through 4G.
3.6 METHODS In the following sections, it will be useful to refer to the methods used to analyze DBA, large breaks, and small. breaks. For jet-pump reactors, these are defined as follows:
a.
DBA Methoh. LAMB / SCAT / SAFE /DBA-REFLOOD/ CHASTE. Break size: DBA.
3-3 i
l i
b.
Lerne Break Methods (LBM). LAMB / SCAT / SAFE /non-DBA REFLOOD/ CHASTE.
I Break sizes:
1.0 ft2 < A < DBA.
c.
Small Break Methods (SBM). SAFE /non-DBA REFLOOD. Heat transfer coefficients: nucleate boiling prior to core uncovery, 25 Btu /
hr-ft
- F after recovery, core spray when appropriate. Peak l
cladding temperature and peak local oxidation are calculated in non-DBA-REFLOOD. Break sizes: A < 1.0 ft l
l l
l l
I 3-4
NEDO-24085-1 Table 2
SUMMARY
OF BREAK SPECTRUM RESULTS e Break Size Core-Wide e Location Peak Local Metal-Water e Single Failure PCT (*F)
Oxidation (%)
Reaction (%)
e
(2) e-Recirc Discharge e Gas Turbine (1) PCT from CHASTE (2) Less than DBA value e
4
. 5
NEDO-24085-1 Table 3 LOCA ANL J SIS FIGURE
SUMMARY
Large Break Methods Small Break Methods Maximum Suction Break DBA (LPCI Inj ection 0.10 ft Discharge Valve Failure)
Break (Gas (4.34 tt )
Turbine Failure) 2 Water Level Inside Shroud and 1
-8 Reactor Vessel Pressure Peak Cladding Temperature 2
9 Heat Transfer Coefficient 3
9 Core Average Inlet Flow 4
Minimum Critical Power Ratio 5
Peak Cladding Temperature of 2
the Highest Powered Plane Experiencing Boiling Transition Variation with Break Area of 6
Time for Which Hot Node Remains Uncovered 7
Peak Cladding Temperature Versus Break Area 3-6
NEDO-24085-1 Table 4A MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT: Millstone-1_
FUEL TYPE: P8DRB282 Average Planar Exposure MAPLHL' PCT 0xidation (mwd /st)
(KW/ft)
(*F)
Fraction 200 10.7 2198 0.036 1000 10.7 2197 0.035 5000 11.1 2198 0.034 10000 11.3 2197 0.033 15000 11.3 2199 0.033 20000 11.2 2196 0.033 25000 11.2 2198 0.033 30000 11.0 2198 0.082 35000 10.4 2084 0.057 40000 9.8 1976 0.038 Table 4B MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE
_ PLANT: Millstone-1 FUEL TYPE:
F8DRB265H Average Planar Exposure MAPLHGR PCT 0xidation (mwd /st)
(KW/ft)
M Fraction 200 10.8 2199 0.035 1000 10.8 2198 0.035 5000 11.1 2199 0.034 10000 11.3 2195 0.033 15000 11.3 2199 0.033 20000 11.2 2196 0.033 25000 11.2 2198 0.033 30000 10.7 2134 0.061 35000 10.2 2042 0.046 40000 9.6 1925 0.028 3-7
NEDO-24085-1 Table 4C MA?LHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT: Millstone-1 FUEL TYPE:
8DRB265H Average Planar Exposure MAPLHGR PCT 0xidation (mwd /st)
(KW/ft)
(*F)
Fraction 200 10.6 2196 0.035 1000 10.7 2196 0.035 5'000 10.9 2198 0.034 10000 11.1 2198 0.033 15000 11.1 2198 0.034 20000 10.9 2196 0.03t:
25000 10.9 2198 0.034 30000 10.7 2179 0.032 35000 10.2 2097 0.025 40000 9.2 2000 0.018 Table 4D MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT: Millstone-l FUEL TYPE: 8DRB265L Average Planar Exposure MAPLHGR PCT 0xidation (mwd /st)
(KW/ft)
(*F)
Fraction 200 10.6 2194 0.035 1000 10.7 2198 0.035 5000 11.0 2199 0.034 10000 11.1 2197 0.033 15000 11.1 2198 0.034 20000 13,0 2198 0.034 25000 10.9
.798 0.034 30000 10.7 2179 0.032 35000 10.2 2098 0.025 40000 9.6 1996 0.018 3-8
NEDO-24085-1 Table 4E MAPLHCR VERSUS AVERAGE PLANAR EXFOSURE PLANT: Millstone-1 FUEL TYPE:
8DB274H Average Planar Exposure MAPLHGR PCT Oxidation (mwd /st)
(KW/ft)
(*F)
Fraction 200 10.9 2197 0.033 1000 11.0 2198 0.032 5000 11.3 2198 0.032 10000 11.5 2199 0.031 15000 11.4 2200 0.031 20000 11.3 2199 0.031 25000 11.2 2197 0.031 30000 10.9 2161 0.028 35000 10.0 2020 0.018 40000 9.3 1928 0.013 Table 4F MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT: Millstone-1 FUEL TYPE: 8DB274L Average Planar Exposure MAPLHCR PCT 0xidation (mwd /st)
(KW/ft)
(*F)
Fraction
- 00 10.9 2198 0.033 1000 11.0 2198 0.032 5000 11.3 2198 0.032 10000 11.5 2197 0.031 15000 11.4 2199 0.031 20000 11.3 2198 0.031 25000 11.2 2193 0.031 30000 10.9 2158 0.028 35000 9.9 2016 0.017 40000 9.3 1924 0.012 3-9
Table 4G MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT: Millstone-1 FUEL TYPE:
8DB262 Average Planar Exposure MAPLHCR PCT 0xidation (mwd /st)
(KW/ft)
M Fraction 200 10.9 2200 0.033 1000 11.0 2198 0.032 5000 11.3 2199 b.032 10000 11.5 2196 0.030 15000 11.5 2199 0.031 20000 11.3 2198 0.031 25000 11.2 2199 0.031 30000 10.7 2128 0.025 35000 9.8 1988 0.016 40000 9.2 1895 0.011 A multiplier of 0.95 is to be applied to the values in Tables 4A through 4G when operating at less than 90% of rated core flow.
J l
3-10
NED0-24085-1 4.
DESCRIPTION OF MODEL AND INPUT CHANES This section provides a general description of the.nput and model changes as they relate to the break spectrum calculations.
It provides a general background so that the more specific calculated results rhown in subsequent sections can be more easily understood.
The majority of the input and model changes primarily affect the amount of ECCS flow entering the lower plenum as a result of the counter current flow limiting (CCFL) effect. These changes as applied to Millstone-1 are listed below.
1.
Input Changes a.
Corrected Vaporization Calculation - Coefficients in the vaporization correlation used in the REFLOOD code were corrected.
b.
Incorporated more accurate bypass areas - The bypass areas in the top guide were recalculated using a more accurate technique.
c.
Corrected guide tube thermal resistance.
2.
Model Change Core CCFL pressure differential = 1 psi - Incorporate the a.
assumption that flow from the bypass to lower plenum must overcome a 1 psi pressure drop in core.
b.
Incorporate NRC pressure transfer assumption - The assumption used in the SAFE-REFLOOD pressure transfer when the pressure is increasing was changed.
4-1
NEDO-24085-1 A few of the changes affect the accident calculation irrespective of CCFL.
These changes are listed below.
1.
Input thange a.
Break Areas - the DBA break area was calculated more accurately.
b.
Gas Turbine Failure - the systems remaining operable following a failure of the gas turbine coincident with low pressure
. coolant injection (LPCI) systems selection of the broken loop for injection was considered for small breaks less than 0.1 f t in area in the discharge side of the recirculation piping.
In this situation, the operating systems are one low pressure core spray, partial flow from two LPCI pumps, the isolation condenscr, and four ADS valves (1 LPCS + f x 2 LPCI + IC + 4 ADS).
For breaks larger than 0.1 ft in area, the LPCI system selects the unbroken loop for injection.
c.
Addition of the icolation condenser (IC) as an ECCS - the effect of the IC is primarily on small breaks less than 0.2 ft in area in hastening the vessel depressurization.
d.
The addition of a faurth ADS valve - serves to hasten vessel depressurization on small breaks.
2.
Model Change a.
Improved Radiation and Conduction Calculation - Incorporation of CHASTE 05 for the heatup calculation.
b.
Partial Drill Model - hastens reflooding for large breaks.
Application of this model is given in Reference 11 and approval for its use in Reference 10.
4-2
--n-,
NED0-24085-1 5.
CONCLUSIONS The results of the LOCA analysis in accordance with the requirements of Reference 2, for Millstone-1 with loop selection logic and fuel bundles with drilled lower tieplates in a partial core loading, is presented in Figures 1 Y
through 9.
This analysis is for the entire recirculation line break spectrum.
The characteristics that determine which is the most limiting break are:
(1) the calculated hot node reflooding time; (7) the calculated hot node uncovery time; and (3) the time of calculated boiling transition.
The time of calculated boiling transition increases with decreasing break size, since jet pump suction uncovery (which leads to boiling transition) is deter-mined primarily by the break size for a particular plant. The calculated hot node uncovery time also generally increases with decreasing break size, as it is primarily determined by tha inventory loss during the blowdown. The hot node reflooding time is determined by a number of interacting phenomena such as depressurization rate, counter current flow limiting and a combination of available ECCS.
The period between hot node uncovery and reflooding is the period when the hot node has the lowest heat transfer. Hence, the break that results in the longest period during which the hot node remains uncovered generally results in the b.ighest calculated PCT.
If two breaks have similar times during which the hot node remains uncovered, then the larger of the two breaks will be more limiting, as it would have an earlier boiling transition time (i.e., the larger break would have more severe results from the LAMB / SCAT blowdown heat transfer analysis).
Figure 6 shows the variation with break size of the calculated time the hot node 2
remains uncovered for Millstone-1.
Based on these calculations, the DBA (4.34 ft )
5-1 l
l NEDO-24085-1 l
l L
l was determined to be the break that results in the highest calculated PCT in the large break region (1.0 ft to DBA). The worst failure is failure of the LPCI injection valve.
i t
2 In the break region between 0.10 ft and 1.0 ft, the LPCI injection valve is i
also the worst failure., The calculated PCT in this break range is less than that calculated for the DBA, as shown in Figure 7.
i 1
2 l
For break areas of 0.10 ft and below, a discharge break with gas turbine failure and injection of LPCI into the broken recirculation loop is the most limiting break / failure combination. The CHASTE PCT for the limiting break size j
(0.10 ft ) is less than the PCT for the DBA.
A low flow MAPLHGR multiplier of 0.95 was derived for application to this LOCA analysis. This multiplier is consistent with Reference 13.
The single-failure evaluation showing the remaining ECCS following an l
assumed failure and the effects of a single failure or operator error that I
causes any manually controlled, electrically operated valve in the ECCS to j
move to a position that could adversely affect the ECCS are presented in i
Reference 12.
l l
I i
I 5-2
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1200 1 VESSEL PRESSURE Ipsiel 2 8AF - 17.59 i:
3 TAF - 29. lie it 4 WATER LEVEL Ift) l MILLSTONE SUCTION DSA LPCI-IV FAILURE
- soO 7 40 l
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l 2
20 4
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0 100 200 300 400 500 TIME (sec)
Figure 1.
Water Level Inside the Shroud and Reactor Vessel Pressure Following a Maximum Recirculation Line Suction Break LPCI Injection Valve Failure, Break Area = 4.34 f t2 (LBM)
m 31i00 i
l HIGHEST POWER AXIALPLANE 5000
--- LOWEST AXIAL PLANE TO EXPERIENCE CPR OF 1.C PRIOR TO JET PUMP UNCOVERY HIGH POWER AXIAL PLANE 1
REFLOODED 2s00 1
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t l
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l l
I l
l l
l o
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10 20 40 60 100 200 400 600 TIME (sec)
Figure 2.
Peak Cladding Temperature Following a Maximum Recirculation Line Suction Break, LPCI Injection Valve Failure Break Area = 4.34 f t2 (I.BM)
NED0-24085-1 100.000
~
- JET PUMP UNCOVERY
- ONSET OF SOlWNG TRANSITION 1 f 10,000 s
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- 1 8e U O
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4Y y 3 100 5
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{z PLANE UNCOVERED l
0
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- ONSET OF LOWER PLENUM FLASHING 10 1.0 0
10 20 30 40 TIME (sec)
Figure 3.
Fuel Rod Convective Heat Transfer Coefficient During Blowdown at the liigh Power Axial Node for a Maximum Recirculation Line Suction Break (4.34 f t2) 5-5
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Figure 5.
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Line Suction Break (4.34 ft )
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Figure 9.
Peak Cladding Temperature and Convective Heat Transfer Coefficient Following a 0.10 ft2 Recirculation Line Discharge Break, Gas Turbine Failure (SBM)
f NEDO-24085-1 6.
REFERENCES t
1.
Letter, George Lear (NRC) to Donald C. Switzer (NUSCO), " Millstone Nuclear Power Station Unit No. 1",
dated March 11, 1977.
2.
Letter, Darrell G. Eisenhut (NRC) to E. D. Fuller (GE), " Documentation of the Reanalysis Rasults for the Less-of-Coolant Accident (LOCA) of Lead and Non-Lead Plants." June 30, 1977.
3.
General Electric Ccapany Analytical Model for. Loss-of-Coolant Analysis in Accordance with 10CRF50 Appendix K, NEDO-20566 (Draft), submitted August 1974, and General Electric Refill Reflood Calculation (Supplement to SAFE Code Description) transmitted to the USAEC by letter, G. L. Gyorey (GE) to Victor Stello, Jr. (NRC), dated December 20, 1974.
4.
" Safety Evaluation for General Electric ECCS Evaluation Model Modifications,"
letter from K. R. Goller (NRC) to G. G. Sherwood (GE), dated April 12, 1977.
5.
Letter, A. J. Levine (CE) to D. F. Ross (NRC) dated Janusry 27, 1977,
" General Electric (GE) Loss of Coolant Accident (LOCA) Analysis Model Revisions - Core Heatup Code CHASTE 05."
6.
Letter, A. J. Levine (CE) to D. B. Vassallo (NRC), dated March 14, 1977,
" Request for Approval for Use of Loss of Cools.nt Accident (LOCA) Evalua-tions Model Code REFLOOD05."
7.
" Supplemental Information for Plant Modification to Eliminate Significant In-Core Vibrations," Supplement 1, NEDE-21156-1, September 1976.
8.
" Supplemental Information for Plant Modification to Eliminate Significant In-Core Vibrations," Supplement 2, NEDE-21156-2, January 1977.
9.
Letter, R. Engel (CE) to V. Stello (NRC), " Answers to NRC Questions on NEDE-21156-2," January 24, 1977.
10.
Letter, D. G. Eisenhut (NRC) to R. L. Gridley (GE) dated June 7, 1978.
11.
Letter, R. E. Engel (GE) to D. G. Eisenhut (NRC), " Loss of Coolant Accident Analysis Methods for BWR 2/3 With Drilled Lower Tieplates (NEDE-24094)," dated January 17, 1978.
12.
Letter, D. C. Switzer (NUSCO) to G. Lear (NRC), " Millstone Nuclear Power Station Unit No. 1 Loss-of-Coolant Accident Reanalysis Performed in Accordance with 10CFR50, Appendix K, and Revised Tecanical Specifications", dated July 9, 1975.
13.
Letter, R. L. Gridley (GE) to D.,G. Eisenhut (NRC), " Review of Low Core Flow Effects on LOCA Analysis for Operating BWR's", dated September 28, 1978.
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