ML20091B591
| ML20091B591 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 09/30/1977 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20091B589 | List: |
| References | |
| NEDO-24050, NUDOCS 9106100524 | |
| Download: ML20091B591 (34) | |
Text
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.. Dxumiib LOSS-0T-COOLANT ACCIDENT ANALYSIS REPORT FOR tt0NTICELLO NUCLEAR GENERATING PLANT I
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1 t 112CRTANT NOTICE REGARDING C05fENTS OF THIS REPORT ,
i i Please Read Caref% tty [
t l,
i The only undertakings of Generat Electrio company respecting infor-ution in 1
j this doowwnt are contained in the contraat between Northern states Power j Corgang and Generat Eteatric Conpany and nothing contained in this dooment i shall be construed ao changing the contract. The uso of this informtion 1
i by anyone other than Northern States Power Conyany or for any purpose other ,
I than that for uhtoh 't is intended, is not authorised; and with respect to j '
1 my unauthorized use, General Electric Conpany nakes no representation or
' wrranty, and asewss no liability as to the completenece, accuracy, or t i usef%tneee of the infomation contained in this doowent, e
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f . NEDO-24050
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i j TABLE OF CONTENTS 1
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- 1. INTRODUCTION 1-1 f,
- 2. LEAD PLANT SELECTION 2-1 i
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- 3. INPUT TO ANALYSIS 3-1 l
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! 4. LOCA ANALYSIS COMPUTER CODES 4-1 i
- 4.1 Results of the LAMB Analysis 4-1 1
- 4.2 Results of the SCAT Analysis 4-1 I 4.3 Results of the SATE Analyais 4-1 1
4.4 Results of RETLOOD Analysis 4-2 4.5 Results of the CHA8TE Analysis 4-3 4.6 Methods 4-4 l
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- 5. DESCRIPTION OF MODEL AND IN'UT CHANCES 5-1 1
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- 6. CONCLUSIONS 6-1
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- 7. REFERENCES 7-1 l
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NEDO-24050 l 4 8 l
i LIST OF TABLES f
i Table Title g 3
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1 1 Significant Input Parameters to the Loss-of-Coolant 3-1 I Accident l
) 4-5 j 2 Summary of Break Spectrum Resulta 4
3 LOCA Analysis Figure Summary - Non-Lead Plant 4-6
, 4A MAPLHGR Versus Average Planar Exposure 4-7 4B HAPulGR Versus Average Planar Exposure 4-8 j
4C HAPLHGR Versus Averago Planar Exposure 4-9 a
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NEDO-240$0 e i LIST OF ILLUSTRATIONS Fi gu re Title Page la Water Level Inside,the Shroud and Reactor Vessel Pressure Following a 1.6 ft' Recirculation Line Suction Break, LPCI Injection Valve Failure (40% DBA) (LBM) 6-3 lb Water f.evel Inside the Shroud and Reactor Vessel Pressure Following a 4.0 ft2 Recirculation Line Suction Break. LPCI Injection Valve Failure, (DBA) 6-4 2a Peak Cladding Temperature Following a 1.6 f t 2 Recirculation Line Suction Break LPCI Injection Valve Failure, Break-Area = (40% DBA) (LBM) 6-5 2b Peak Cladding Temperature Following a 4.0 f t 2 Recirculation Line Suction Break, LPCI Injection Valve Failure, (DBA) 6-6 3a Fuel Rod Convective Heat Transf er Coef ficient During Blowdown at the High Power Axial Node Following a 1.6 ft 2 Recirculation Line Suction Break, LPCI Injection Valve Failures, (40% DBA) 6-7 3b Fuel Rod Convective Heat Transfer Coefficient During Blowdown at the High Power Axial Node for a 4.0 ft2 Recirculation *ne Suction Break, LPCI Injection Valve Failure. (DBA, 6-8 4a Normalized Core Average Inlet Flow Following a 2.4 ft2 Recirculation Line Suction Break, (60% DBA) 6-9 4b Normalised Core Average Inlet Flow Following a 4.0 f t2 Recirculation Line Suction Break (DBA) 6-10
$a Minimum Critical Power Ratio Following a 2.4 ft2 Recirculation Line Suction Break (60% DBA) 6-11
$b Minimum Critical Power Ratio Following a 4.0 ft2 Recirculation Line Suction Break (DBA) 6-12 6 Variation with Break Area of Time for Which Hot Node Remains Uncovered 6-13 vii/viii I
!iEDO 24050 t 8 1 INT ROOL* CT ION The purponc of this documet.t in to provide the results of the loss-of-coolant accident (LOCA) analysis for the Monticello Nuclear Generating Plant (Honticello).
The analysis was perforced using approved General Electric (GE) calculatienal models.
This reanalysis of the plant LOCA is provided in accordance with the NRC re qui re ment (Re ference 1) and to demonstrate conformance with the ECCS ac ceptance criteria o f 10CTR50.46. The obje ctive o f the LOCA analysis con-tained herein is to provide assurance that the most limiting break size ,
break location, and single f ailure combination has been considered for the plant. The required documentaticn for demonstrating that these objectives have been satisfied is given in Reference 1. The documentation contained in this report is intended to satisfy taese requirements.
The general description of the 10CA evaluation models is contained in Re f e rence 3. Recently approved model changes (Reference 4) are described in Ref erences 5 and 6. These model changes are employed in the new 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 lover tieplate of the fuel bundle and the use of such fuel bundles in a full or partial core loading is described in Reference * *, C, and 9. This model was also approved in Reference 4. Also included in the reanalysis are current values for input parameters based on tha LOCA analysis reverification program being carried out by GE. The specific changes as applied to Monticello are discussed in more detail in later sections of this document.
Plants are separated into groups for the purpose of LOCA analysis (Reference 10). Within each plant group there vill be a single lead plant analysis which provides the basis for the telection of the most limiting break size yielding the highest peak cladding temperature (PCT) . Also, the lead plant analysis provides an expanded documentation base to provide added insight into evaluation of the details of particular phenomena. The remainder of 1-1
NED0-24050 a
the plants in that grou; util nave n:n-lead plant analyses ref e renced to the lead plant analysis. This docueent contains the n:n-lead plant analysis for Monticello, which is a BVR/3 group of plants and is consistent with the requirements outlined in Reference 2.
The same n.odels and conmuter codes ace used to evaluate all plants. Ch an ge s to these models will cause changes in phenomenological responses that are similar within any given plant group. The dif ference in input paramen te rs are not expected to result in significantly dif ferent results for the plants within a given group. Emergency core cooling system (ECCS) and geometric d 'f erences between plant groups may result in different responses for di.Jerent groups but within any group the responses will be similar. Input changes have been made in the new analysis which are essentially an upgrading of the input paramenters to the computer codes. Thus, the lead plant concept is still valid for this evaluation.
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1 NEDO-24050 I 2. LEAD PLAS". SELECTION j Lead plants are selected and analy:sd in detail to permit a more comprehen-I sive review and eliminate unnecessary calculations. This constitutes a generic a..alys!' for each plant of that type which can be referenced in subsequent plant s ubmitt als .
The lead plant for Monticello is Quad Cities. The justification for categorizing.
Monticello in this group of plants and the lead plant analysis for this group is y
presented in Reference 11.
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- 3. INPUT TO ANALYSIS A list of the significant plant input parameters to the LOCA analysis is presented in Table 1.
1, I Table 1
! !GNIFICANT 1,NPUT PARAMETERS TO THE
! LOES-OF-COOLANT ACCIDENT ANALYSIS Plant Pa rame t e rs :
Core Thermal Power 1703 MWt, which corresponds to i 102% of rated core power i
Vessel Steam output 6.91 x 106 lbm/h. which corre- ,
i sponds to 102% of rated core power l
i Vessel Steam Dome Pressure 1040 paia Recirculation Line 3reak 1.6 ft2 (40% DBA), 4.0 f t2 (DBA)
) Area for large Breaks - Suction i
Number of Drilled B.adles 0 1
Fuel Parameters :
Peak Technical Initial Specification Design Minimum Linear Heat Axial Critic al Tual Bundle Generation Rati Peaking Power
- Fuel Type Geome t ry (kW/ft) Factor Ratio
- A. 8D219 8x8 13.4 1.57 1.2 B. 8D250 8x8 13.4 1.57 1.2 C. 8D262 8x8 13.4 1.57 1.2 I
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- To account for the 2* uncertainty in bu idle power required by Appendix K, the SCAT calculation is perforr.ed with an MCPR of 1.18 (i.e. ,1.2 divided
- by 1.02) for a bundle with an initial MCPR of 1.20.
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!. . LOCA ANALYSIS C0."PUTER CODES j !. 1 RESULTS OF THE LA'!B ANALYSIS This code is used to analyze the short-term blowdown phenomens for large postu-laced pipe breaks (breaks in which nucleate boiling is lost before the water level drops and uncovers the active fuel) in jet pump reactv e The L#'3 output j (core slow as a function of time) is input to the SCAT c de 4r .alcw_ation of blowdown heat transfer.
1 The 1 rib results presented are:
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e Core Average Inlet Flow Rate (normalized to unity at t~i beginning of the accident) following a Large Break.
- 4.2 RESULTS OF THE SCAT ANALYSIS Thas code completas the transient short-term thermal-hydraulic calculation f or large breaks in jet pump reactors. The GEXL correlation is used to track the boiling transition in time and location. The post-critical heat flux heat transf er correlations are built into SCAT which calculates heat transfer coefficients for input to the core heatup code, CHASTE.
The SCAT results presented are:
i' e Minimum Critical Power Ratio following a Large Break.
e Convective Heat Transfer Coef ficient following a Large Break.
4.3 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.
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NEDO-24050 The SAfr. results presented are:
e Wtter Level inside the Shroud (up to the tine REFLOOD initiates) and Reactor Vessel Pressure 4.4 RESULTS OF 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 accident (DBA) application (" DEA-REFLOOD") was described in a supplement to the SAFE code description transmitted to the USSRC December 20, 1974 The "non-DBA REFLOOD" analysis is nearly identical to the DBA version and employs the same maior assumptions. The only differences stem from the fact that the core may be partially covered with coolanc at the time of ECCS initiation and coolant levels change slowly for smaller breaks by comparison with the DBA. More precise modeling of coolant leve' 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 DEA version in which greater precision is not necessary.
(2) The non-DBA version includes a heatt p model similar to but less detailed than that in CRASTE, designsd to calculate cladding temper-ature during the small break. This hiatup 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 4-2 l
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! 4.5 RESULTS OF THE CRASTE ANALYSIS l
t i This code is used, with suitable inputs from the other codes, to calculate the 1
fuel cladding heatup rate, peak cladding temperature, peak local cladding
]l oxidation, and core-wide metal-water reaction for large breaka. The detailed j fuel model in CRASTE considers transient gap conductance, clad swelling and 6 i rupture, and metal-water reaction. The empirical core spray heat transfer and channel wetting correlations are built into CRASTE, which solves the transient l
heat transfer equations for the entire LOCA transient at a single axial plane
} in a single fuel assembly. Iterative 'pplications of CHASTE determine the 4
- maximum permissible planar power where required to satisfy the requirements of
} 10CFR50.46 acceptance criteria.
i The CHASTE results presented are i
a j e Peak Cladding Temperature versra time i
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! e Peak Cladding Temperature versus Break Area i
1 e Peak Cladding Temperature and Peak Local Oxidation versus Planar
. Average Exposure for the most limiting break size I
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{ e Maximum Average Planar Heat Generation Rate 01APLEGK) versus Planar Average Exposure for the most limiting break size A summary of the analytical results is given in Table 2. Table 3 lists the j figures provided for this analysis. The MAPLHGR values for each fuel type in l the Monticello core are presented in Tables 4A through 4C.
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!. 6 !!ETHODS In the following sections, it will be useful to refer to the methods used to analyze DBA, large breaks, a 4 small breaks. For j et-pump reactors, these are defined as followst
- b. Large Break !!ethods (LEM). LAMB / SCAT / SAFE /non-DBA REFLOOD/ CHASTE.
2 Break sizes: 1.0 ft 1 A < DBA.
- c. Small Break Methods (SBM). SAFE /non-DBA REFLOOD. Heat transfer coefficients: nucleate boiling prior to core uncovery, 25 Btu /hr-f t *F
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after recovery, core spray when appropriate. Peak cladding temperature
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,? and peak local oxidation are calculated in non-DBA-REFLOOD. Break I sizes: A 1 1.0 ft ,
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Table 2
SUMMARY
OF BREAK SPECTRUM RESULTS e Break Size Core-Wide e Lccation Peak Local Metal-Water e Single Failure PCT (*F) Oxidation (%) Reaction (%)
2200(1) 0.23 e 1.6 fe (40% DBA) 3.4 e Recire Suction e LPCI Injection Valve e 4.0 ft (DBA) 2095 I1) Note 2 Note 3 e Recire Suction l e LPCI Injection Valve
- 1. PCT from CHASTE
- 2. Less than most limiting break (3.4%)
- 3. Less than most limiting break (0.23%)
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NEDo-24050 Table 3 LOCA ANALYSIS FIGURE
SUMMARY
- NON-LEAD PLANT Large Break Methods Limiting Maximum Suction Break Suction Break (LPCI Injection (LPCI Injection Valve Failure) Valve Failure)
(1.6 ft )2(40% DBA) (4.0 ft )2 (DRA) k'ater Level Inside Shroud la lb and Reactor Vessel Pressure Peak Cladding Temperature 2a 2b Heat Transfer Coef ficient 3a 3b Core Average Inlet Flow 4a 4b Minimum Critical Power Ratio 5a Sb Peak Cladding Temperature of 2a the Highest Powered Plane Experiencug Boiling Transition Variation with Break Area of 6 Time for htich Fot Node Remains Uncovertid 0
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- NECO-24050 4
Table 4A 1
, MAPLHGR VERSUS AVEPAGE PLANAR EXPOSURE 1
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i Plant: Monticello Fuel Type 8D219 4
I J Average Planar j Exposure MAPLHGR PCT 0xidation (mwd /t) (kW/ft) ('F) Fraction 200 10.7 2199 0.033 f
1,000 10.7 2199 0.033
' 5,000 10.8 2200 0.033 -
! 10,000 10.7 2196 0.033 15,000 10.7 2199 0.033 20,000 10.6 2194 0.033 l
- 25,000 10.6 2200 0.034 30,000 10.2 2138 0.028 4
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i Plant: Monticello Fuel Type SD250 J
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- (mwd /t) (kW/ft) ('F) Fraction i
- 300 10.6 2195 0.033
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1,000 10.7 2198 0,033
- 5,000 10.7 2195 0.033 1
,1 10,000 10.8 2194 0.032
' 15,000 10.7 2197 0.033 j
20,000 10.6 2196 0.033 25,000 10.6 2198 0.033 1 30,000 10.6 2199 0.034 t
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1 I Table 4C l MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE I
! Plant: Monticello Fuel Type 8D262 I
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Average Planar j Exposure MAPLHGR PCT 0xidation (mwd /t) (kW/ft) (*F) Fraction i
l 200 10.6 2197 0.033 i
l 1,000 10.7 2195 0.033 5,000 10.7 2196 0.033 i 10,000 10.8 2197 0.033 15,000 10.7 2199 0.033 a
.j 20,000 10.7 2198 0.033 1
j 25,000 10.6 2196 0.033
! 30,000 10.6 2198 0.034 i
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- 5. DESCRIPTION OT MODEL AND INPL'T CRA :GES l
l This section provides a general description of the input and model changes as they relate to the break spectrum calculations. It provides a general background so that the more specific calculated results shown in subsequent sections can be 1
more easily understood, particularly as they relate to how well trends observed in specific lead plant break spectrum analyses can be applied to the general nonlead plant case. The most limiting break size results are not discussed in j this context (except to the extent that they affect the shape of the break spectrum) because detailed limiting break size calculational results will be J
presented for each plant.
i The majority of the input and model changes primarily affect the amount of ECCS
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j flow entering the lower plenum as a result of the counter current flow limiting (CCFL) effect. These changes as applied to Monticello are listed below.
- 1. Input Changes j a. Corrected Vaporization Calculation - Coefficients in the vaporiza-tion correlation used in the REFLOOD code were corrected.
j b. Incorporated more accurate bypass areas - The bypass areas in the top guide were recalculated using a more accurate technique.
a c. Corrected guide tube thermal resistance.
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- d. Correct heat capacity of reactor internale head nodes.
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- a. Core CCFL pressure differential = 1 psi - Incorporate the assumption that flow from the bypass to lower plenum must overcome a 1 psi
- pressure drop in core.
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- b. Incorp arate NRC pressure transfer assumption - The asumption used in the SAFE-REFLOOD pressure transf er when the pressure is increasing j was changed.
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l A few of the changes affect the accident calculation irrespective of CCFL. These changes are listed below.
- 1. Input Change j
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- a. Break Areas - The DBA break area was calculated more accurately.
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- 2. Model Change
- a. Improved Radiation and Conduction Calculation - Incorporation of CRASTE 05 for heatup calculation.
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! 6. CONCLUSIONS 1
l l The LOCA analysis results in accordance with the requirements of Reference 2 for non-lead plants are presented in Figures la through Sa for the limiting suction break (40% DBA) and Figures 1b through 5b for the maximum suction break (DBA).
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! The characteristics that determine which is the most limiting break area at the DBA location are:
(a) the calculated hot node reflooding time, (b) the calculated hot node uncovery time, and
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(c) the time of calculated boiling transition.
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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 j is primarily determined by the inventory loss during the blowdown. The hot node reflooding time is determined by a number of interacting phenomena such as I depressurization rate, counter current flow limiting and a combination of available ECCS, I 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 results in the highest cal-culated PCT. If two breaks have similar times during which the hot node remains uncovered, then the larger of the two breaks will be limiting as it would have
, an earlier boiling transition time (i.e., the larger break would have a more severe LAMB / SCAT blowdown heat transfer analysis).
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d Figure 6 shows the variation with break size of the calculated time the hot i node remains uncovered for Monticello, Based on these results the 40% DBA I was determined to be the break that results in the highest calculated PCT in the 1.0 ft to DBA region. The determination of the 40% DBA being the l most limiting break was based on the reasoning discussed above and the pro-
! cedure used for the lead plant. The 40% DBA was determined to be the most 1
i limiting break smaller than the DBA from Figure 6. Then a CHASTE calcula-1 i tion was performed to compere the PCT for the DBA and the 40% DBA. The 40%
i i DBA was determined to result in a higher PCT compared to the DBA and, hence, t
j was determined to be the most limiting break.
I The conservative approach of using the 60% DBA LAMB / SCAT results with the 40%
< DBA SAFE /REFLOOD results for calculations for the 40% DBA was used in all cal-l culations for the analysis to determine the MAPLHGR's in Tables 4A through 4C.
l The DBA (the complete severence of the recirculation discharge piping) results
.I are shown on Figures 1b through Sb. The most significant change in these ;
l results from the previous analysis is that the reflooding time decreases f rom approximately 330 seconds to approximately 260 seconds. This is due to the input and model changes described in Section 5.
f The single failure evaluation showing the remaining ECCS following an assumed i failure and the effects of a single failure or operator error that causes any a
manually controlled, electrically operated valve in the ECCS to move to a j position that could adversely affect the ECCS are presented in Reference 12.
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1200 7
800 -
WATER LEVEL -
40 7
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(s VESSE L PRESSURE 0 300 400 500 0 1 T im im.
Figure la. Water Level Inside the Shroud and Reactor Vessel Pressure Following a 1.6 ft Recirculation Line Suction Break, LPCI Injection Valve Failure (40% DBA) (LBM)
_ _ . _ _ _ _ . _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ . . __ . . _ _ _ . _ _ _ - _ . . _ . _ . . . . . . . . . . _ _ _ . _ _ _ _ .. __.m._.._ . _ . . ..._. - -... . . _ - ... _ _._ _ _
e 1200
- m 800 -
WATER LEVEL -
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20 5 BAF r 1
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TIME Inect 4
Figure Ib. Water Level Inside the Shroud and Reactor Vessel Pressure Following a 4.0 f t Recirculation ,
Line Suction Break, LPCI Injection Valve Failure, (DBA) i
e 2500 HIGH POWER AXIAL PLANE REFLOODED~
Ir HIGHEST POWER AXIAL PLANE
- --- LOWEST AXI AL PLANE TO 2000 -- EXPERIENCE CPR 1.0 PRIOR TO KT PUMP UNCOVE RY
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g 1500 -
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s h Y s e / 8 i
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4 1000 -
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,,,,,. . == " " , m m = p/ k
- ONSET OF SPR AY COOLING a.
- HIGH POWE R AX8 AL PLANE a
500 - Ji UNCOVERS
- ONSET OF BOILING TRANSITION
, I , 1 0 50 100 93 2 5 10 T ME iswa Figure 2a. Peak Cladding Ten:perature Following a 1.6 f t Recirculation Line Suction Break, LPCI Injection Valve Failure, Break Area = (40% DBA) (LBM)
___._._.__m._ _-_...._..____._._____.___.m. _ _. m m ____..m...m.__ .__-.__..._..m..
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2500 HIGH POvvERED AXIAL 4
PLANE REFLOODED ~
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j SPRAY t- 1500 - COOLING -
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- HIGH POWER !
4 AXIAL PLANE UNCOVERS
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- ONSET OF
! BOILING TRANSITION O
2 5 to 50 100 ,
$m T#ML sm s Figure 2b. Peak Cladding Temperature Following a 4.0 f t Recirculation Line Suction Break, LPCI Injection !
l Valve Failure, (DBA)
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NED0-340$0 e .
100.000
-JET PUMP UNCOVE RY
-ONSET OF BOILING TRANSITION it 10.000 - if I
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i 5 07 wN I Of C!
- 3 1000 -
- 5E U
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- - ONSET OF LOWER i 10 PLENUM FLASHING J
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Figure 3a. Fuel Rod Convective Heat Transfer Coefficient During Blowdown I at the High Power Axial Node Following a 1.6 ft2 Reci:culation Line Suction Break, LPCI Injection Valve Failures, (40% DBA) i I
' 6-7 I
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Figure 3b. Fuel Rod Convective Heat Transfer Coefficient During Blowdown at the High Power Axial Node for a 4,0 f t2 Recirculation Line Suction i Break, LPCI Injection Valve Failure, (DBA) i s
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Figure 4b. Normalized Core Average Inlet Flow Following a 4.0 ft Recirculation Line Suction Break (L.,A)
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- Figure 6. Variation with Break Area of Time for Which Hv. Node Remains
- i. Uncovered 6-13/6-14 I
NED0-24050 g .e 6 7 RE FERENCES
- 1. Letter, Dennis L. Ziemann (NRC) to L.O. Mayer (NSP), "Re t Monticello Nuclear Generating Plan," dated March 11, 1977.
- 2. Letter, Darrell G. Eisenhut (NRC) to E.D. Fuller (GE), Documentation of the Reanalysis Results for the Loss-o't-Coolant Accident (LOCA) of Lead and Non-Lead Plants." June 30,1977
- 3. General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CRF50 f.ppendix K, NEDO-20566 (Draf t), submitted August 1974, and General Electric Refill Reflood Calculation (Supplement to SAFE Code Description) transmitted to the US AEC by letter, G.L. Gyorey (GE) to Victor Stello, Jr. (NRC), dated December 20, 1974.
- 4. " Safety Evaluation for General Electric ECCS Evaluation Mocel Modifications,"
letter from K.R. Goller (NRC) to G.G. Sherwood (GE), dated April 12, 1977.
- 5. Letter, A.J. Levine (GE) to D.F. Ross (NRC) dated January 27, 1977,
" General Electric (GE) Loss of Coolant Accident (LOCA) Analysis Model Revisions - Core Heatup Code CHASTE 05."
- 6. Lstter, A.J. Levine (GE) to D.B. Vassallo (NRC), dated March 14, 1977,
" Request for Approval for Use of Loss of Coolant Accident (LOCA) Evaluations Model uode REFLOOD05."
- 7. " Supplemental Information for Plant Modification to Eliminate Significant
'n-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, G.L. Gyorey (GE) to V. Stello, Jr. , dated May 12, 1975, " Compliance with Acceptance Criteria for 10CFR50.46."
- 11. " Loss of Coolant Accident Analysis Report For Dresden Units 2, 3 and Quad Cities Units 1, 2 Nuclear Power Stations (Lead Plant)," NEDO-24046, dated Augus t, 1977.
- 12. Letter, L.O. Mayer (NSP) to D.L. Ziemann (NRC), "Monticello Nuclear Generating Plant, Docket No. 50-263, License No. DP"-22, Transmittal of ECCS Analys ts," dated July 9,1975.
I 7-1/7-2
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iC poxu 195 - u.s.Nuct Am nsovLATony couuissioN oocxsv Nvusa n
- SO * .2 &
. pRC DISTNIBUTl'ON con PART 50 DOCKET MATERIAL
- FROM: oAtt or cocuusNT Northern States Power Company 9/15/77 Mr. Victor Stello Minneapolis, Minnesota 3 7, ,, e ,, y , ,
L. O. Mayer 9/19/77
.s t rt n CNotomizso Pmor iNPv7Pomu suusam or copies mactivso
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ConioiNAL CuscLAss:Pis o Jcorv / d d.
semirvios sNcLosums a
Loss of Coolant Accident Analysis Report )
f or Monticello Nuclear Generating Plant, NEDO 24050, September, 1977 - l
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(1-P) (21-P)
'. ANT NAME: Monticello RJL 9/19/77 '
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S AFETY FOR ACTION /INFORMATION l BRANCH CHTEF: (7) DA.Vis J
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