ML20128A042
| ML20128A042 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 05/31/1985 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML18029A620 | List: |
| References | |
| NEDO-24088-2, NEDO-24088-2-R02, NEDO-24088-2-R2, NUDOCS 8507020551 | |
| Download: ML20128A042 (50) | |
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"""fss" M AY 1985 LOSS-OF-COOLANT ACCIDENT ANALYSIS FOR BROWNS FERRY l
NUCLEAR PLANT UNIT 2 l
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CK 60 GEN ER AL h ELECTR
NED0-24086-2 Revision 2 77NED360 Class I May 1985 J
LOSS-OF-COOLANT ACCIDENT ANALYSIS FOR BROWNS FERRY NUCLEAR PLANT 1
UNIT 2 i
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4 NUCLEAR ENERGY BUSINESS OPERATIONS + GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA 95125 GENERAL $ ELECTRIC
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I NEDO 24088*2 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT 4
Please Read Carefully The only undertaking of General Electric Company respecting infomation in this document are contained in the contract betueen the Tennessee Valley Authority and General Electric Company and nothing contained in this doctonent shall be construed as changing the contract. The use of this information by anyone other than the Tennessee Valley Authority or for any purpose other than that for uhich it is intended, is not authorized; and uith respect to any unauthorized use, General Electric Company makes no representation or varranty, and assiones no I
liability as to the completeness, accuracy, or usefulness of the infomation contained in this docw:ent.
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NEDO-24088-2 TABLE OF CONTENTS Page 1.
INTRODUCTION 1-1 2.
LEAD PLANT SELECTION 2-1 3.
INPUT TO ANALYSIS 3-1 4.
LOCA ANALYSIS COMPUTER CODES 4-1 4.1 Results of the LAMB Analysis 4-1 4.2 Results of the SCAT Analysis 4-1 4.3 Results of the SAFE Analysis 4-1 f
4.4 Results of the REFLOOD Analysis 4-2 4.5 Results of the CHASTE Analysis 4-3 4.6 Methods 4-4 5.
DESCRIPTION OF MODEL AND INPUT CHANGES 5-1 6.
CONCLUSIONS 6-1
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6.1 Results for 7x7 Fuel 6-2 6.2 Results for 8x8 Fuel 6-3 6.3 Applicability 6-4 7.
REFERENCES 7-1 ili/iv
NED0-24088-2 LIST OF TABLES Table Title Pm 1
Significant Input Parameters to the Loss-of-Coolant Accident Analysis 3-1 2
Summary of Results 4-5 3
LOCA Analysis Figure Summary 4-6 4A MAPLHGR Versus Average Planar Exposure 4-7 4B MAPLHGR Versus Average Planar Exposure 4-8 4C MAPLHGR Versus Average Planar Exposure 4-9 4D MAPLHGR Versus Average Planar Exposure 4-10 4E MAPLHCR Versus Average Planar Exposure 4-11 4F MAPLHGR Versus Average Planar Exposure 4-12 4G MAPLHGR Versus Average Planar Exposure 4-13 t
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IND0-24b88-2 LIST OF ILLUSTRATIONS Figure Title Page la Water Level Inside the Shroud and Reactor Vessel Pressure Following a Maximum Recirculation Line Suction Break, LPCI Injection Valve Failure, Break Area = 4.2 f t2 (DBA) 6-5 lb Water Level Inside the Shroud and Reactor Vessel Pressure Following a Maximum Recirculation Discharge Break, LPCI Injection Valve Failure, Break Area = 1.9 f t2 (LBM) 6-6 Ic Water Level Inside the Shroud and Reactor Vessel Pressure Following a Recirculation Line Discharge Break, LPCI Injection Valve Failure, Break Area = 1.3 f t2 (66% DBA) (LBM) 6-7
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2a Peak Cladding Temperature Following a Maximum Recirculation Line Suction Break, LPCI Injection Valve Failure, Break Area = 4.2 ft2 (DBA) 6-8 2b Peak Cladding Temperature Following a Maximum Recirculation Line Discharge Break, LPCI Injection Valve Failure, Break Area = 1.9 ft2 (LBM) 6-9 2c Peak Cladding Temperature Following a Recirculation Line
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Discharge Break, LPCI Injection Valve Failure, Break Area = 1.3 ft2 (66% DBA) (LBM) 6-10
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3a Fuel Rod Convective Heat Transfer Coefficient During Blowdown at the High Power Axial Node Following a Maximum Recirculation Line Suction Break, LPCI Injection Valve Failure, Break Area = 4.2 ft2 6-11 3b Fuel Rod Convective Heat Transfer Coefficient During Blowdown at the High Power Axial Node Following a Maximum Recirculation Line Discharge Break, LPCI Injection Valve Failure, Break Area = 1.9 f c2 6-12 3c Fuel Rod Convective Heat Transfer Coefficient During Blowdown at the High Power Axial Node Following a Recirculation Line Discharge Break b (PCI Injection L
Valve Failure, Break Area = 1.3 fe 66% DBA) 6-13
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4a Normalized Core Average Inlet Flow Following a Maximum j
Recirculation Line Suction Break, Break Area = 4.2 fc2 6-14
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4b Normalized Core Average Inlet Flow Following a Maximum j
Recirculation Line Discharge Break, Break Area = 1.9 ft2 6-15 l
4c Normalized Core Average Inlet Flo'r Following a Maximum l
Recirculation Line Discharge Break, Break Area = 80% DBA 6-16 l
I Sa Minimum Critical Power Ratio Following a Maximum Recirculation Line Suction Break, Break Area = 4.2 ft2 6-17 Sb Minimum Critical Power Ratio Following a Maximum r
i Recirculation Line Discharge Break, Break Area = 1.9 ft2 6-18 1
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. - - ~ _ _. _ _ _ _,... _. _ _ _ - - - _ _. _ _ _.. _ _ _ _ _ _. _ _. _ _
NEDO-24088*-2 LIST OF ILLUSTRATIONS (Continued)
Page Figure Title a
f Sc Minimum Critical Power Ratio Following a Maximum l
Recirculation Line Discharge Break, Break Area = 80% DBA 6-19 i
6a Variation With Break Area of Time for Which Hot Node Remains Uncovered (Discharge) 6-20 6b Variation With Break Area of Time for Which Hot Node Remains Uncovered (Suction) 6-21 I
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NED0'-24088-2 1.
INTRODUCTION The purpose of this document is to provide the results of the loss-of-coolant accident (LOCA) analysis for the Browns Ferry Nuclear Plant Unit 2 (BF-2).
The analysis was performed using approved General Electric (GE) calculational models.
This reanalysis of the plant LOCA is provided in accordance with the NRC require-ment (Reference 1) and to demonstrate conformance with the ECCS acceptance criteria of 10CFR50.46. The objective of the LOCA analysis contained 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.
Recently approved model changes (Reference 4) are described in References 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 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.
Also included in the reanalysis are current values for input parameters based on the LOCA analysis reverification program being carried out by GE.
The specific changes as applied to BF-2 are discussed in more detail in later sections of this document.
i Plants are separated into 8roups for the purpose of LOCA analysis (Reference 10).
i Within each plant group there will be a single lead plant analysis which provides the basis for the selection of the most limiting break size yielding the highest peak cladding temperature (PCT). Also, the lead plant analysis provides an I
I expanded documentation base to provide added insight into evaluation of the details of particular phenomena. The remainder of the plants in that group will l
have non-lead plant analyses referenced to the lead plant analysis. This document contains ther non-lead plant analysis for BF-2, which is a BVR/4 l
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NEDd-24088-2 with low pressure coolant injection (LPCI) system modification group plant and is consistent with the requirements outlined in Reference 2.
The same models and computer codes are used to evaluate all plants. Changes to these models will cause changes in phenomenological responses that are similar within any given plant group. The difference in input parameters are not expected to result in significantly different results for the plants within a given group.
Emergency core cooling system (ECCS) and geometric differences between plant groups may result in different responses for different 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 parameters to the computer codes.
Thus, the lead plant concept is still valid for this evaluation.
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NEDO-24088-2 2.
LEAD PLANT SELECTION Lead plants are selected and analyzed in detail to permit a more comprehensive review and eliminate unnecessary calculations. This constitutes a generic analysis for each plant of that type which can be referenced in subsequent plant submittals.
The lead plant for BF-2 is James A. FitzPatrick Nuclear Power Plant. The justification for categorizing BF-1 in this group of plants and the lead plant analysis for this group is presented in Reference 11.
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i NEDO-24088-2 3.
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 3440 MWt, which corresponds to 105% of rated steam flow 6
Vessel Steam Output 14.05 x 10 lbm/h, which corresponds to 105% of rated steam flow I
Vessel Steam Dome Pressure 1055 psia
.i Recirculation Line Break Area 2
for Large Breaks - Discharge 1.9 ft2 (DBA) 1.3 ft (66% DBA)
- Suction 4.2 ft Number of Drilled Bundles 744 Fuel Parameterst Peak Technical Initial Specification Design Minimum Linear Heat Axial Critical Fuel Bundle Generation Rate Peaking Power Fuel Type Geometry (kW/ft)
Factor Ratio
- A.
IC Type 1 & 3 7x7 18.5 1.5 1.2 B.
IC Type 2 7x7 18.5 1.5 1.2 C.
8DB274L 8x8 13.4 1.4 1.2 D.
8DB274H 8x8 13.4 1.4 1.2 E.
8DRB284L 8x8 13.4 1.4 1.2 F.
P8DRB284L 8x8 13.4 1.4 1.2 C.
P8DRB265H 8x8 13.4 1.4 1.2
- To account for the 2% uncertainty in the bundle power required by Appendix K, the SCAT calculation is performed 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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NED0-24088-2 4.
LOCA ANALYSIS COMPUTER CODES 4.1 RESULTS OF THE LAMB ANALYSIS This code is used to analyze the short-term blowdown phenomena 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 reactors. The LAMB output (core flow as a function of time) is input to the SCAT code for calculation of l
blowdown heat transfer.
The LAMB results presented are:
o Core Average Inlet Flow Rate (normalized to unity at the beginning of I
the accident) following a Large Break.
4.2 RESULTS OF THE SCAT ANALYSIS This code completes the transient short-term thermal-hydraulic calculation for 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 transfer correlations are built into SCAT which calculates heat transfer coefficients for input to the core heatup code, CHASTE.
The SCAT results presented are:
e Minimum Critical Power Ratio following a Large Break.
e Convective Heat Transfer Coefficient following a Large Break.
i 4.3 RESULTS OF THE SAFE ANALYSIS l
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.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 i
sizes. The code is used during the entirs 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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The SAFE results presented are:
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e Water Level inside the Shroud (up to the time REFLOOD initiates) and 1
Reactor Vessel Pressure 4.4 RESULTS OF REFLOOD ANALYSIS This code is used across the break spectrum to calculate the system inventories 4
i after ECCS actuation. The models used for the design basis accident (DBA) application ("DBA-REFLOOD") was described in a supplement to the SAFE code 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 l
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 levels change slowly for smaller breaks by comparison with the DBA. More precise l
modeling of coolant level behavior is thus requested principally to determine j
the contribution of vaporizatien in the fuel assemblies to the counter current l
flow limiting (CCFL) phenomenon at the upper tieplace. The differences from i
i 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.
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j (2) The non-DBA version includes a heatup model similar to but less detailed than that in CHASTE, designed to calculate cladding temper-I ature during the small break. This heatup model is used in calculating l
vaporization for the CCFL correlation, in calculating swollen level in i
l the core, and in calculating the peak cladding temperature.
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The REFLOOD results presented are:
e Water Level inside the Shroud 4
f Peak Cladding Temperature and Heat Transfer Coefficient for breaks e
l calculated with small break methods t
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4.5 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 l
oxidation, and core-wide metal-water reaction for large breaks. The detailed fuel model in CHASTE considers transient gap conductance, clad swelling and rupture, and metal-water reaction. The empirical core spray heat transfer and l
channel wetting correlations are built into CHASTE, which solves the transient l
heat transfer equations for the entire LOCA transient at a single axial plane in a single fuel assembly.
Iterative applications of CHASTE determine the maximum permissible planar power where required to satisfy the requirements of 10CFR50.46 acceptance criteria.
The CHASTE results presented are:
Peak Cladding Temperature versus Time e
Peak Cladding Temperature versus Break Area e
Peak Cladding Temperature and Peak Local Oxidation versus Planar e
Average Exposure for the most limiting break size e
Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) versus Average Exposure for the most limiting break size A summary of the analytical results is given in Table 2.
Table 3 lists the figures provided for this analysis. The MAPLHCR values for each fuel type in the BF-2 core are presented in Tables 4A through 4G.
4-3
NEDO-24088-2 4.6 METHODS In the following sections, it w'111 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 Methods. LAMB / SCAT / SAFE /DBA-REFLOOD/ CHASTE. Break size: DBA.
b.
Large Break Methods (LBM). 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 2
coefficients: nucleate boiling prior to core uncovery, 25 Btu /hr-ft _.7 after recovery, core spray when appropriate. Peak cladding temperature and peak local oxidation are calculated in non-DBA-REFLOOD. Break 2
sizes A 1 1.0 ft,
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NED0-24088-2 Table 2
SUMMARY
OF R2SULTS e Break Size Core-Wide e Location Peak Local Metal-Water e Single Failure PCT (*F)
Oxidation (%)
Reaction (%)
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e 1.3 ft2 (66% DBA) 2128(1)
Note 2 0.12
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e Recirc Discharge e LPCI Injection Valve e 1.9 ft (DBA) 2086( }
Note 2 Note 3 e Recire Discharge LPCI Injection Valve e
e 4.2 ft 2151(1) 5.7%
Note 3 2
e Recire Suction e LPCI Injection Valve 1.
PCT from CHASTE.
2.
Less than most limiting break (5.7%).
3.
Less than most limiting case (0.12%).
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IDCA ANALYSIS FIGURE SUMARY i
i Maximum Suction Break Maximum Discharge Break Limiting Discharge Break l
(LPCI Injection (LPCI Injection (LPCI Injection i
Valve Failure)
Valve Failure)
Valve Failure)
(4.2 ft2)
(1.9 ft2)
(66% DBA) 1 Water Level Inside Shroud and Reactor Vessel Pressure la lb ic i
Peak Cladding Temperature 2a 2h 2c i
Heat Transfer Coefficient 3a 3b 3c Core Average Inlet Flow 4a 4b 4c 8
Minimum Critical Power h
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Ratio 5a Sb Sc j
j Peak Cladding Temperature h
of Highest Powered Plane
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Experiencing Boiling
,i Transition 2a Uncovered Time Versus Break Area 6a,b I
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NEDO-24088-2 Table 4A MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Plant: BF-2 Fuel Type: Initial Core - Type 1 & 3 Average Planar Exposure MAPLHGR PCT 0xidation (Mwd /t)
(kW/ft)
(*F)
Fraction 200 15.0 1926 0.009 1,000 15.1 1902 0.008 5,000 16.0 1975 0.011 10,000 16.3 2047 0.015 15,000 16.1 2151 0.055 20,000 15.4 2136 0.054 25,000 14.2 2035 0.037 30,000 13.1 1922 0.023 35,000 11.8 1821 0.015 40,000 10.5 1640 0.003 4
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MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE t
Plant: BF-2 Fuel Type: Initial Core - Type 2 Average Planar Exposure MAPLHGR PCT 0xidation (Mwd /t)
(kW/ft)
(*F)
Fraction l
200 15.6 1973 0.010 1,000 15.5 1956 0.009 5,000 16.2 1973 0.010 10,000 16.5 2063 0.016 15,000 16.5 2143 0.057 20,000 15.8 2119 0.055
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25,000 14.5 2005 0.038 30,000 13.3 1886 0.024 35,000 11.9 1782 0.015 40,000 10.6 1615 0.003
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i Table 4C MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Plant: BF-2 Fuel Type: 8DB274L t
Average Planar Exposure MAPLHGR PCT 0xidation (Mwd /t)
(kW/ft)
(*F)
Fraction
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200 11.2 1673 0.004 1,000 11.3 1681 0.004 5,000 11.9 1744 0.005 i
10,000 12.1 1755 0.005
'15,000 12.2 1777 0.005 20,000 12.1 1778 0.005 25,000 11.6 1737 0.004 30,000 10.9 1666 0.003 35,000 9.9 1577 0.003 l
40,000 9.3 1520 0.003 E
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NEDO-24088-2 Table 4D MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Plant: BF-2 Fuel Type: 8DB274H Average Planar Exposure MAPLHGR PCT 0xidation (Mwd /t)
(kW/ft)
(*F)
Fraction
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200 11.1 1667 0.004 1,000 11.2 1670 0.004 5,000 11.8 1730 0.005 10,000 12.1 1756 0.005 15,000 12.2 1778 0.005 20,000 12.0 1775 0.005 25,000 11.5 1733 0.004 30,000 10.9 1667 0.003 35,000 10.0 1579 0.003 40,000 9.3 1522 0.003 I
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NED0-24088-2 Table 4E MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Plant: BF-2 Fuel Type: 8DRB284L Average Planar Exposure MAPLHCR PCT 0xidation F-(mwd /t)
(kW/ft)
(*F)
Fraction 200 11.2 1747 0.005 1,000 11.3 1746 0.005 5,000 11.8 1804 0.006 10,000 12.0 1815 0.006 15,000 12.0 1827 0.007 20,000 11.8 1814 0.006 25,000 11.2 1759 0.005 30,000 10.8 1708 0.004 35,000 10.2 1651 0.003 40,000 9.5 1574 0.002 4-11
NEDO-24088-2 Table 4F
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MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Plant: BF-2 Fuel Type: P8DRB284L Average Planar Exposure MAPLHGR PCT 0xidation (mwd /t)
(kW/ft)
(*F)
Fraction 200 11.2 1738 0.005 1,000 11.3 1740 0.005 5,000 11.8 1792 0.006 10,000 12.0 1801 0.006 15,000 12.0 1809 0.006 20,000 11.8 1801 0.006 25,000 11.2 1745 0.005 30,000 10.8 1684 0.004 35,000 10.2 1620 0.003 40,000 9.5 1546 0.002 45,000 8.8 1472 0.001 4-12
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NEDO-24088-2 Table 4G MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE
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Plant: BF-2 Fuel Type: P8DRB265H Average Planar Exposure MAPLHGR PCT Oxidation (Mwd /t)
(kW/ft)
(*F)
Fraction
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200 11.5 1755 0.005 1,000 11.6 1762 0.005 5,000 11.9 1795 0.006 10,000 12.1 1812 0.006 15,000 12.1 1821 0.007 20,000 11.9 1813 0.007 25,000 11.3 1755 0.005 30,000 10.7 1687 0.004 35,000 10.2 1623 0.003 40,000 9.6 1551 0.002 4-13/4-14
NED0-24088-2 5.
DESCRIPTION OF MODEL AND INPUT CHANGES 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 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 this context (except to the extent that they affect the shape of the break spectrum) because detailed limiting break size calculational results will be presented for each plant.
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 BF-2 are listed below.
1.
Input Changes a.
Corrected Vaporization Calculaticn - Coefficients in the vapor-ization 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 Core Power in REFLOOD - The core power in REFLOOD was corrected to 102% of rated power.
d.
Corrected guide tube thermal resistance, e.
Correct heat capacity of reactor internals head nodes.
2.
Model Change a.
Core CCFL pressure differential = 1 psi - Incorporate the assump-tion that flow from the bypass to lower plenum must overcome a 1 psi pressure drop in core.
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NEDO-24088-2 b.
Incorporate NRC pressure transfer assumption - The assumption used in the SAFE-REFLOOD pressure transfer when the pressure is increasing was changed.
A few of the changes affect the accident calculation irrespective of CCFL.
These changes are listed below.
1.
Input Change a.
Break Areas - The DBA break area was calculated more accurately, b.
Core Power - The core power in REFLOOD has been corrected to 102% of rated.
2.
Model Change Improved Radiation and Conduction Calculation - Incorporation of a.
CHASTE 05 for heatup calculation.
b.
Suction Line Friction in Discharge Valve Closure Assumption -
Took credit for friction due to irreversible losses in the suction line.
Modify Recirculation Discharge Valve Closure Assumption - Assume c.
the valve does not close for a discharge break.
Note: This analysis takes credit for flow through holes in fuel lower tieplate. Since the previous analysis, alternate flow paths have been incorporated in fuel lower tieplates.
5-2
NED0-24088-2 6.
CONCLUSIONS Analyses have demonstrated that failure of the LPCI is the most severe failure among the low pressure ECCS because, unlike the core spray which must pass through the CCFL regions at the top of the core, LPCI is injected into the lower plenum through the jet pumps. Thus, the LPCI injection valve is the worst single failure in the large break region. This is the case for a break occurring in either the suction or discharge piping. For a break in the dis-charge piping, this failure results in no LPCI flow, and for a suction line failure, LPCI flow is minimized.
Comparison of the calculated PCT's for the maximum size break in the suction and discharge piping determines which is the DBA and which is the second most limiting location. For BF-2, the suction break is the most limiting location (Table 3) for 7x7 fuel. 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 c.
the time of calculated boiling transition.
The time of calculated boiling transition increases with decreasing break size, since jet pump suctiop uncovery (which leads to boiling transition) is determined primarily by the breaE hize for,c particular plant. The calculated hot node uncovery efee also tenerally increases with decreasing break size, as it is s
primarily determined by the inventory loss during the blowdown. The hot node reflooding tLEe is determined by"a numbhr of interactin3 phenomena such as depressurization rate, counter current flow limiting and a combination of
,r available ECCS.
s The period betDeen hot nedA uncovery and reflooding is the period when the hot node has the lowest haat transfer. The break that, results in the lengest period during which eb.c' hot node remains uncovered usually results in the highest 6
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NEDO-24088-1 calculated 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).
6.1 RESULTS FOR 7x7 FUEL Figures 6a and 6b show the variat1on with discharge and suction break size of
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the calculated time the hot node remains uncovered for Browns Ferry Unit 2.
show-Thd shape of the discharge break curve is very similar to the lead plant, ing a maximum hot node uncovered time at a break less than the DBA (i.e.,
66% DBA). However, owing to the geometric differences of Browns Ferry Unit 2 relative to the lead plant (larger peripheral bypass area and smaller vessel volume to break area ratio), the reflooding times for all discharge break 2
sizes from 1.0 ft to the DBA are considerably lower than that of the lead plant. This resulted in the suction break being more limiting than any of the discharge breaks.
The determination of the suction DBA being the most limiting break was based on the reasoning discussed above and the procedure used for the lead plant.
From Figure 6a the 66% discharge DBA was determined to be the most limiting discharge break smaller than the discharge DBA. Also, from Figure 6b the suction DBA was determined to be che most limiting suction break. A CHASTE calculation was performed to compare the PCT for the 66% discharge DBA, the discharge DBA, and the suction DBAY The suction DBA was determined to result in the highest PCT of the three break sizes and, hence, was determined to be
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the most limiting break. The results for 7x7 fuel are presented in Tables 4A and 4B.
- The CHASTE calculation for the discharge DBA used the discharge DBA LAMB / SCAT and SAFE /REFLOOD results. The CHASTE calculation for the suction DBA used the suction DBA LAMB / SCAT and SAFE /REFLOOD results.
In accordance with the conservative approach used for the lead plant, the 80% discharge DBA LAMB / SCAT results were used with the 66% discharge DBA SAFE /REFLOOD results to determine the 66% discharge DBA results.
6-2
NED0-24088-2 The DBA (the complete severence of the recirculation suction piping) results are shown on Figures la through Sa.
These results have changed very little from the previous analysis because with LPCI flow available the flood time is relatively insensitive to parameters that affect CCFL.
The second most limiting location for the LOCA is the recirculation discharge line. The results of the maximum break in this piping are shown on Figures le through Sc.
These results have changed very little from the previous analysis because the decrease in flow through the bundles, due to increased CCFL resulting from the new pressure assumption, is offset by the flow through the lower tie plate holes.
The single failure evaluation showing the remaining ECCS following an assumed failure and the effects of a single failure or operator error that cause any manually controlled, electrically operated valve in the ECCS to move to a position that could adversely affect the ECCS are presented in Reference 12.
6.2 RESULTS FOR 8x8 FUEL For 8x8 fuel bundles, the worst break is the 66% discharge break rather than the 100% DBA suction break, which is the worst for 7x7 fuel. This difference is due to the higher stored energy of the larger fuel rods in the 7x7 fuel. The PCT of the 7x7 fuel is more sensitive to early boiling transition, while the PCT of the 8x8 fuel is more sensitive to the total uncovered time.
Since the core configuration has changed to all 8x8 fuel (8x8, 8x8R, and P8x8R) the ECCS analysis will be virtually identical to that of the Browns Ferry 3 analysis reported in Reference 13.
Although large PCT margin exists, some adjustments in MAPLHGRs were made to be consistent with fuel thermal-mechanical heat flux limits.
Analyses were performed for all of the 8x8 fuel types for the 66% DBA discharge break, and the results are presented in Tables 4C through 4G.
6-3 m
NED0-24088-2 6.3 APPLICABILITY 1.
Single-Loop Operation This analysis is valid for operation with one recirculation loop out-of-service with the following MAPLHGR reduction factors applied:
Fuel Type MAPLHGR Multiplier 7x7 0.70 8x8 0.83 8x8R/P8x8R 0.82 2.
Safety Relief Valve Out-of-Service This analysis is valid for operation with two safety relief valves out-of-service.
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NED0-24088-2 1
1 7.
REFERENCES 1.
Letter, A. Schwencer (NRC) to Godwin Williams, Jr (TVA), "Re: Browns Ferry Nuclear Plant, Units Nos. 1 and 2," dated March 11, 1977.
2.
Letter, Darrel C. Eisenhut (NRC) to E. D. Fuller (GE), " Documentation of the Reanalysis Results for the Loss-of-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 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 (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.
Letter, A. J. Levine (CE) to D. B. Vassallo (NRC), dated March 14, 1977,
" Request for Approval for Use of Loss of Coolant Accident (LOCA) Evaluations 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 (GE) to V. Stello (NRC), " Answers to NRC Questions on NEDE-21156-2," January 24, 1977.
10.
Letter, G. L. Gyorey (CE) to V. Stello, Jr., dated May 12, 1975, " Compliance with Acceptance Criteria for 10CFR50.46."
11.
Letter, George T. Berry (PASNY) to Robert W. Reid (NRC), " James A. FitzPartick Nuclear Power Plant ECCS Analysis Docket No. 50-333," dated July 29, 1977.
12.
" Emergency Core Cooling System Analysis Appendix K Requirements,"
NED0-20973, dated August, 1975.
~
13.
"LOCA Analysis for Browns Ferry Nuclear Plant Unit 3," NED0-24194A, July 1979.
7-1/7-2
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