ML20206P294
ML20206P294 | |
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Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
Issue date: | 06/30/1986 |
From: | Dasilva H, Fernandez R, St John K YANKEE ATOMIC ELECTRIC CO. |
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YAEC-1547, NUDOCS 8607020086 | |
Download: ML20206P294 (143) | |
Text
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VERMONT YANKEE BWR LOSS-OF-COOLANT ACCIDENT LICENSING ANALYSIS METHOD bY Dr. R. Thomas Fernandez Dr. Hugo C. da Silva, Jr.
I June 1986 I
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Yankee Atomic Electric Company Nuclear Services Division 1671 Worcester Road Framingham, Massachusetts 01701 l
8607020086 860627 PDR ADOCK 0500 1
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rL Prepared By:
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Def R. T. Fernandez, Principal En er (Date)
LOCA Analysis Group
/& C.d<~ W $.
G/2 G/8G Dr. M.'C. da Silva, Jr., Engin'eer (Date)
LOCA Analysis Group b
Reviewed By:
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Mr. K. E. St.Johrf Nuclear Engineer (Date)
Nuclear Evaluations and Support Group O E4/8b Dr. 'S.'P. Schult;/,' Managk (Date)
Nuclear Evaluations and Support Group L
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Ms. IC Sironen. 6'enior Engineer (Da(e)
VY Nuclear Engineering Coordinator Reactor Physics Group Reviewed and Approved By:
d 2b!9b Dr. A. Husaid, Manager
'(Dale)
LOCA Group Approved By:
N 2d M'r. P. A.'Berge'ron, P Acting Director
/ (D/te)
Nuclear Engineering partment h
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I DISCLAIMER OF RESPONSIBILITY This document was prepared by Yankee Atomic Electric Company I
(" Yankee"). The use of information contained in this document by anyone other than Yankee, or the Organization for which the document was prepared under contract, is not authorized and, with respect to any unauthorized use, neither I
Yankee nor its officers, directors, agents, or employees assume any obligation, responsibility, or liability or make any warranty or representation as to the accuracy or completeness of the material contained i
in this document.
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E ABSTRACT This report describes the BWR loss-of-coolant accident licensing analysis method for the Vermont Yankee Nuclear Power Station. This method will be used to perform full break spectra (size and location) and cycle I
exposure analyses that comply with USNRC regulations contained in 10CFR50.46 L
and Appendix K thereto. The method basically uses the RELAPSYA LWR System Thermal-Hydraulic Analysis Computer Program developed and extensively assessed r
by YAEC. The extensive code assessment work has established the validity of L
the code to predict the complex thermal-hydraulic phenomena encountered in twR LOCA events. In particular, the code assessuant has shown that use of the evaluation model features in RELAP5YA required by Appendix K yields reliable, conservative predictions when compared to experimental results such as from TLTA and Marviken tests. Therefore reliable, conservative predictions are obtained for the Vermont Yankee LOCA licensing analyses since these same features plus additional conservative assumptions are used.
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TABLE OF CONTENTS Page DISCLAIMER OF RESPONSIBILITY....................................
11 L
iii ABSTRACT........................................................
LIST OF FIGURES.................................................
v LIST OF TABLES..................................................
ix ACKN0WLEDGEMENTS................................................
X
1.0 INTRODUCTION
1 I
1.1 Purpose....................................................
1.2 Vermont Yankee NSSS Summary................................
2 6
1.3 RELAPSYA Summary...........................................
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l 2.0 NUCLEAR STEAM SUPPLY SYSTEM M0 DEL...............................
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2.1 NSSS Volumes, Junctions, and Heat Structures...............
16 S
2.2 Core Power.................................................
17 2.3 Trip Logic.................................................
18 F
2.4 Emergency Core Cooling Systems.............................
21 L
2.4.1 High Pressure Coolant Injection (HPCI).............
21 2.4.2 Automatic Depressurization System (ADS)............
21 2.4.3 Low Pressure Core Spray (LPCS).....................
21
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2.4.4 Low Pressure Coolant Injection (LPCI)..............
22 2.5 Initial Conditions.........................................
22 L
3.0 HOT CHANNEL M0 DEL...............................................
53 I
3.1 Obj ective and Description of the Model.....................
53 3.2 Boundary Conditions and Validation.........................
57 l
4.0 SAMPLE PROBLEMS 1 AND 2: LARGE RECIRCULATION LOOP BREAKS.......
61 4.1 Sample Problem 1: Large Break Case EA.....................
63 4.2 Sample Problem 2: Large Break Case EB.....................
83 s
5.0 SAMPLE PROBLEM 3: SMALL RECIRCULATION LOOP BREAK...............
103
6.0 CONCLUSION
S.....................................................
123
7.0 REFERENCES
125 APPENDIX A:
FUEL ROD INITIAL CONDITIONS FROM FROSSTEY CODE.....
127 m
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5 LIST Or FIGURES Number Title Page I
1.2-1 Vermont Yankee Reactor Vessel 10 2.1-1 Vermont Yankee NSSS Nodalization Diagram 24 58 3.1-1 VY HC Axial Nodalization 3.1-2 VY HC Rod Layout:
(a) Number and Location of the Rods 59 I
Represented and (b) As Modeled 70 4.1-1 Reactor Power History (LBLOCA-EA) 70 4.1-2 Net Reactivity (LBLOCA-EA) 4.1-3 Feed and Main Steam Flows (LBLOCA-EA) 71 4.1-4 Vessel Water Level (LBLOCA-EA) 71 72 4.1-5 Early Break Flow Rates (LBLOCA-EA) 72 4.1-6 Long-Term Break Flow Rates (LBLOCA-EA) 8 4,1-7 Vessel Pressure History (LBLOCA-EA) 73 73 4.1-8 ECCS Flow Rates (LBLOCA-EA) 74 4.1-9 Uet Flow Rate Into NSSS (LBLOCA-EA) 4.1-10 NSSS Fluid Mass inventory (LBLOCA-EA) 74 75 4.1-11 Bypass and Upper Plenum Fluid Mass (LBLOCA-EA) 4.1-12 CRGT and Lower Plenum Fluid Mass (LBLOCA-EA) 75 4.1-13 Outer and Central Core Fluid Mass (LBLOCA-EA) 76 76 4.1-14 High Power Assembly Fluid Mass (LBLOCA-EA) 77 4.1-15 High Power Bundle Clad Temperatures (LBLOCA-EA) 77 4.1-16 High Power Bundle Qualities (LBLOCA-EA) 78 4.1-17 Long-Term Heat Transfer Coefficients (LBLOCA-EA) 78 4.1-18 Degraded Heat Transfer Coefficients (LBLOCA-EA) 4.1-19 Maximum Bundle Clad Temperatures (LBLOCA-EA) 79 4.1-20 Bundle Heat Transfer Coefficients (LBLOCA-EA) 79 80 4.1-21 VY HC Average Rod Clad Temperatures (LBLOCA-EA) 8
-V-
LIST OF FICURES (CONTINUED)
Number Title Page 4.1-22 VY HC Hot Rod Clad Temperatures (LBLOCA-EA) 80 4.1-23 VY HC Average Rod Heat Transfer coefficients (LBLOCA-EA) 81 4.1-24 VY HC Hot Rod Heat T.*ansfer Coefficients (LBLOCA-EA) 81 4.1-25 VY HC Degraded Heat Transfer Coefficients (LBLOCA-EA) 82 4.1-26 VY HC High Power Bun 61e Qualities (LBLOCA-EA) 82 4.2-1 Reactor Power History (LBLOCA-EB) 89 89 4.2-2 Net Reactivity (LBLOCA-EB) 4.2-3 Feed and Main Steam Flows (LBLOCA-EB) 90 4.2-4 Vessel Water Level (LBLOCA-EB) 90 91 4.2-5 Early Break Flow Rates (LBLOCA-EB) 91 4.2-6 Long-Term Break Flowrates (LBLOCA-EB) 4.2-7 Vessel Pressure History (LBLOCA-EB) 92 4.2-8 ECCS Flow Rates (LBLOCA-EB) 92 4.2-9 Net Flow Rate Into NSSS (LBLOCA-EB) 93 4.2-10 NSSS Fluid Mass Inventory (LBLOCA-EB) 93 4.2-11 Bypass and Upper Plenum Fluid Mass (LBLOCA-EB) 94 4.2-12 CRGT and Lower Plenum Fluid Mass (LBLOCA-EB) 94 4.2-13 Outer and Central Core Fluid Mass (LBLOCA-EB) 95 4.2-14 High Power Assembly Fluid Mass (LBLOCA-EB) 95 4.2-15 High Power Bundle Clad Te'aperatures (LBLOCA-EB) 96 96 4.2-16 High Power Bundle Qualities (LBLOCA-EB) 4.2-17 Long-Term Heat Transfer Coefficients (LBLOCA-EB) 97 I
4.2-18 Degraded Heat Transfer Coefficients (LBLOCA-EB) 97 4.2-19 Maximum Bundle Clad Temperatures (LBLOCA-EB) 98 4.2-20 Bundle Heat Transfer Coefficients (LBLOCA-EB) 98
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l LIST OF FIGURES (CONTINUED)
)
Number Title Pate 4.2-21 VY HC Average Rod Clad Temperatures (LBLOCA-EB) 99 4.2-22 VY HC Hot Rod Clad TemperatJees (LBLOCA-EB) 99 P
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4.2-23 VY HC Average Rod Heat Transfer Coefficients (LBLOCA-EB) 100 4.2-24 VY HC Hot Rod Heat Transfer Coefficients (LBLOCA-EB) 100 4.2-25 VY HC Degraded Heat Transfer coefficients (LBLOCA-EB) 101 4.2-26 VY HC High Power Bundle Qualities (LBLOCA-EB) 101 4.2-27 Peak Clad Temperatures for Cases EA and EB 102 5.0-1 Reactor Power History (SBLOCA-EY) 109 i
j 5.0-2 Net Reactivity (SBLOCA-EY) 109 l
5.0-3 Feed and Hain Steam Flows (SBLOCA-EY) 110 5.0-4 Vessel Water Level (SBLOCA-EY) 110 L
5.0-5 Vessel Pressure History (SBLOCA-EY) 111 5.0-6 S/RV and ADS Flow Rates (SBLOCA-EY) 111 5.0-7 Break Flow Rate (SBLOCA-EY) 112 l
L 5.0-8 Break Junction Void Fraction (SBLOCA-EY) 112 H
5.0-9 LPCS Flow Rate (SBLOCA-EY) 113 Iu 5.0-10 LPCI A and B Flow Rates (SBLOCA-EY) 113 I
5.0-11 Net Flow Rate into NSSS (SBLOCA-EY) 114 L
I 5.0-12 NSSS Fluid Hass Inventory (SBLOCA-EY) 114 F
L 5.0-13 Bypass and upper Plenum rluid Ha. (SBLOCA-EY) 115 5.0-14 CRGT and Lower Plenum Fluid Hass (SBLOCA-EY) 115 f
5.0-15 Outer and Central Core Fluid Hass (SBLOCA-EY) 116 5.0-16 High Power Assembly Fluid Hass (SBLOCA-EY) 116 5.0-17 High Power Bundle Clad Temperatures (SBLOCA-EY) 117 5.0-18 High Power Bundle Qualities (SBLOCA-EY) 117
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E LIST OF FIGURES (CONTINUED)
Number Title Page 5.0-19 Long-Term Heat Transfer Coefficients (SBLOCA-EY) 118 5.0-20 Maximum Bundle Clad Temperatures (SBLOCA-EY) 118 5.0-21 Bundle Heat Transfer Coefficients (SBLOCA-EY) 119 5.0-22 Bundle Static Qualities (SBLOCA-EY) 119 5.0-23 VY HC Average Rod Clad Temperatures (SBLOCA-EY) 120 5.0-24 VY HC Hot Rod Clad Temperatures (SBLOCA-EY) 120 L
5.0-25 VY HC Average Rod Heat Transfer Coefficients (SBLOCA-EY) 121 4
l 5.0-26 VY HC Hot Rod Heat Transfer Coefficients (SBLOCA-EY) 121 5.0-27 VY HC High Power Bundle Qualities (SBLOCA-EY) 122 L
E S
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I LIST OF TABLES Number Title Page f
1.2.1 Vermont Yankee Operating Conditions 11
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1.3.1 RELAPSYA Model Development Tasks 12 1.3.2 Separate Effect Tests for RELAPSYA Model Assessment 13 1.3.3 Integral Tests for RELAPSYA Code Assessment 14 2.1.1 Vermont Yankee NSSS Nodalization Summary 25 2.1.2 Summary of VY NSSS Fluid Volumes 26 2.1.3 Summary of VY NSSS Junctions 34 2.1.4 Summary of VY NSSS Model Valves 42 2.1.5 Summary of VY NSSS Heat Structures 43 2.4.1 hPCI Steam Turbine Flow Rates 50 2.4.2 LPCS Injection Velocities and Flow Rates 51 2.4.3 LPCI Injection Velocities and Flow Rates 52 3.1.1 Differences Between VY NSSS and VY HC Numbering Scheme for 60 Heat Structures and Hydrodynamic Components 4.0.1 Summary of Vermont Yankee Large Break Accident Assumptions 62 4.1.1 Sequence of Events for Large Break Case EA 68 87 m
4.2.1 Sequence of Events for Large Break Case EB 5.0.1 Summary of Vermont Yankee Small Break Accident Assumptions 107 108 5.0.2 Sequence of Events for Small Break Case EY A.1 FROSSTEY Component Models 131 I
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ACKNOWLEDGEMENTS The Vermont Yankee LOCA licensing analysis method is the result of f
support. and cooperation from many organizations and individuals.
L The Vermont Yankee, Yankee, Maine Yankee, and New Hampshire Yankee organizations provided continued financial support and motivation throughout the developrant of the RELAPSYA code. The RELAPS MOD 1 base computer program was originally developed and made publicly available by the USNRC and EG&G Idaho Inc. Several of the extensive modifications to the base program were achieved through assistance from Intermountain Technologies, Inc. Research programs sponsored by EPRI, USNRC, and GE provided valuable test data for the code assessment effort.
The development of the Vermont Yankee Nuclear Steam Supply System and hot channel models resulted from the technical support provided by many departments within the VYNPC and YAEC organizations. These include the y{
following:
Vermont Yankee Nuclear Power Station B. Buteau, P. Donnelly, J. Edelhauser, J. Pelletier, R. Lodwick, R. Wanczyk, and T. Watson.
Vermont Yankee Nuclear Training Center A. Chesley, M. Krider, R. Slauenwhite, and D. Tuttle.
17 Vermont Yankee Proiect Office
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R. Capstick, A. Doyle H. Hyams, R. January, P. Johnson, M. Marian, L_,
S. Miller, R. Smith, and D. Yasi.
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YAEC Nuclear EnRineering Department I
B. Baharynejed, K. Burns, J. Cronin, J. Loomis, S. Mihaiu, K. Mitchell, K. St. John, M. Sironen, R. Wochlke, and W. Yeung.
I The Word Processing Center (especially Pam Vierstra, Susan Henchey, and Debbie Stanton for the excellent typing, and also Virginia Hellmuth and r"
Danielle Golove for the excellent proofreading), Computer Services Department, L
and the Reproduction Department provided invaluable services during this project. Finally, this effort was achieved because of the continued support frcm YAEC and VYNPC management, especially from D. Hunter, A. C. Kadak, and
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W. P. Murphy.
We gratefully acknowledge the support from all of these individuals
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and organizations.
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1.0 INTRODUCTTON 1.1 Purpose This report describes the BWR Loss-of-Coolant Accident (LOCA) licensing j
L analysis method for the vermont Yankee Nuclear Power Station. This method will be used to perform cycle independent LOCA-ECCS (Emergency Core Cooling Systems) licensing analyses that comply with USNRC regulations contained in 10CFR50.46 and Appendix K thereto. The development and application of this f
method has been motivated by several considerations:
L Supporting Vermont Yankee's goal to achieve extended fuel cycles of a.
approximately 18 months starting with Cycle 14.
b.
Retaining cycle independent LOCA-ECCS licensing analysis results.
Utilizing new LOCA-ECCS technology, both analytical and c.
experimental, developed by U.S. nucicar institutions over the past decade.
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The Vermont Yankee Nuclear Steam Supply System (VY NSSS) and reactor L
ccre are summarized in Section 1.2.
The Vermont Yankee LOCA licensing analysis method uses the RELAp5YA LWR System Thermal-Hydraulic Computer j
L Program, summarized in Section 1.3, and two base input models. The first is the Vermont Yankee NSSS model described in Section 2.0.
This model will be used to perform the break spectrum study. Hinor modifications are made to the base input data to model various break sizes, locations, and single failure f
assumptions. The second is the Vermont Yankee Hot Channel (VY HC) model described in Section 3.0.
This model uses core inlet and outlet thermal-hydraulic results from the corresponding Vermont Yankee NSSS cases as boundary conditions. This model will be used to perform the burnup study for each fuel assembly type as a function of exposure. Modifications to the base deck will account for variation of fuel assembly design and exposure conditions. Together, results from these models will be used to establish the Design Basis Accident (i.e., most limiting LOCA case) and to show compliance with the LOCA-ECCS criteria in 10CFR50.46 for the VYNPS. Three LOCA sample }
problems are presented in this report to demonstrate the application of the LOCA licensing analysis method to the VYNPS. Section 4.0 describes two large break cases and Section 5.0 describes a small break case.
1.2 Vermont Yankee NSSS Summary The Vermont Yankee Nuclear Power Station is a BWR4 with a Mark I containment that began commercial operation in November 1972. Table 1.2.1 shows the rated and design operating conditions. The reactor is currently licensed to produce 1.593 MWt; however, the turbine generator unit is designed for 1,664 MWt.
The plant nominally produces 540 MWe at rated conditions.
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The nuclear system generally includes those systems most closely f
associated with the reactor vessel which are designed to contain, or be in h
communication with, water coming from or going to the reactor core. The nuclear system includes the following (Section 1.2 of Reference 1-1):
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a.
Reactor vessel g
b.
Resctor vessel internals c.
Reactor core I
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Main steam lines from reactor vessel to the isolation valves outside the primary containment r
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Neutron Monitoring System u
f.
Reactor Recirculation System I
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Control Rod Drive System h.
Residual Heat Removal System J
- i. Reactor Core Isolation Cooling System (RCIC).
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- j. Csro Sttndby Cooling Syntcm (Emergincy Coro Cooling Systcms, ECCS) k.
Reactor Water Cleanup System 1.
Feedwater System piping between the reactor vessel and the first valve outside the primary containment.
The reactor vessel, internals, and core are nhown in Figure 1.2-1.
The reactor vessel has an internal diameter and heightlof 205 inches and 757.5 inches, respectively. The reactor vessel int.ernais include the following components:
89 -
Control rods and guide tubes
,I 30 -
Incore instrumentation tubes I
129 -
Standpipes and steam separators E
6 Chevron type dryers with an annular skirt 10 -
Internal riser pipes connected to the jet pumps su 20 -
Internal jet pumps 4
Sets of low pressure core spray pipes A shroud that defines the lower plenum, core, and upper plenum regions I
various internal support structures for the core, jet pumps, separators, and dryers I
f The reactor core consists of 368 fuel assemblies. Currently, each fuel assembly contains an 8 x 8 array of 62 zircaloy clad fuel rods and two water rods within a zircaloy channel. The axial length of the fuel zone is 150 I
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inches of which 14e inches contain enriched UO2 and six inches at the top contain natural UO. Vermont Yankee has a uniquely large core bypass region r
(inside the cors chroud and external to the fuel assemblies) for a BWR4 due to the er.,latively large vessel diameter and the small number of fuel assemblies.
A typical BWR4 (e.g., Monticello) with a 205-inch ID vessel would have 484 L
fuel assemblies and a core power of 1,670 MWt.
Alternatively, the Duane Arnold plant has a 183-inch vessel ID, 368 fuel assemblies, and was originally licensed for a core power of 1,593 MWt. The latter two values are the same as those for Vermont Yankee.
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Each of the two recirculation loops contain the following components:
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28-inch suction pipe from the vessel to the pump F
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20-inch RHR suction pipe and isolation valve 1
4-inch reactor water cleanup system suction pipe and isolation valve e
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32,250 spm variable speed recirculation pump l
1 28-inch discharge pipe from the pump to the header motor-operated discharge valve immediately downstream of the 1
recirculation pump 1
17.5-inch ID throat venturi in the discharge pipe F
l 24-inch RHR return pipe and isolation valve that also function 1
for LpCI injection 1
22-inch recirculation loop header pipe 5
12-inch external riser pipes from the header to the vessel -
Each of the two independent feedwater systems include the following components from the drywell to the vessel:
16-inch check valves (inside and outside the drywell wall) 2 16-inch feedwater pipe from the drywell to e tee 1
10-inch feedwater pipes from the tee to the vessel 2
During normal operation, the feedwater control system adjusts flow control valves located upstream from the drywell. The feedwater flow rate is set to match the main steam line flow rate plus an amount proportional to the L,
difference between the desired and the actual reactor water level. If a loss r'
of auxiliary power occurs, the condensate and feedwater pumps coast down, the I
flow control valves freeze in position, and the check valves close in about six seconds.
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Each of the four main steam lines include the following components from the vessel to the drywell:
1 18-inch steam line 1
8.7-inch ID throat venturi L
2 Main Steam Line Isolation Valves (MSIVs; one each inside and outside the drywell wall)
P Pilot-operated Safety / Relief Valve (S/RV) that discharges to L
1 the wetwell suppression pool Two of the main steam lines have a spring-loaded safety valve that F
discharges to the drywell. One main steam line has a 10-inch pipe and two motor-operated valves to deliver steam to the HPCI steam turbine. A different steam line has a 3-inch pipe and two motor-operated valves to deliver steam to the RCIC steam turbine. The four main steam lines then run from the drywell, through the steam line tunnel, to the main turbine.
Each has a Turbine control Valve (TCV) and a Turbine Stop Valve (TSV) immediately upstream of the lI l
turbina. Th2 four main ctcam linza era crea3-c:nn:ctGd by th3 proceurs cveraging manifold upstream of the turbine valves. Also, two sets of two main cteam lines each connect to a header that has five Turbine Bypass Valves (TBVs). These mechanical-hydraulically driven valves operate quickly following a change in turbine load (e.g., turbine trip).
The Vermont Yankee Nuclear Power Station has the following Emergency
(
Core Cooling Systems (ECCSs) that are also referred to as Core Standby Cooling Systems (CSCSs):
1 High Pressure Coolant Injection (HPCI) System that injects into one of the two main feed water lines outside the drywell wall.
1 Automatic Depressurization System (ADS) that utilizes the four g
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safety / relief valves to vent reactor vessel steam to the suppression pool.
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2 Low Pressure Core Spray (LPCS) Systems that inject coolant into the upper plenum directly above the core region.
2 Low Pressure Coolant Injection (LPCI) Systems, each which injects into one of the two recirculation loop discharge pipes.
1.3 RELAPSYA Sunnary RELAPSYA, a computer program for light-water reactor system thermal-hydraulic analysis, has been adapted by Yankee Atomic Electric Company for Loss-of-Coolant Accident (LOCA) analyses. RELAPSYA provides a consistent, integral analysis capability of the system and core response to LOCA events and other plant transients. YAEC will use this program as a major part of its method to enalyze the entire BWR break spectrum and the PWR small break spectrum in a manner that conforms to U.S. Nuclear Regulatory Commission requirements contained in 10CFR50.46 and Appendix K.
YAEC has extensively applied this program for other conservative and realistic analyses of LOCA events and transients. I
I The RELAP5YA hydrodyn-mic modal is bnnd upon a cns-dimenziennl, two-fluid, nonequilibrium model. This model accounts for single-phase liquid and gas flows, and for velocity and temperature differences between two fluid phases that frequently occur in LWR Systems. It allows for the presence of a noncondensible gas, such as nitrogen, mixed with the vapor phase. It also allows for a dissolved nonvolatile constituent, such as sodium pentaborate, I
mixed with the liquid phase. The hydrodynamic model contains the necessary conservation of mass, momentum, and energy equations, thermodynamic state relations, and constitutive equations to complete the physical and mathematical description of a generalized fluid system.
I The hydrodynamic model is implemented in the RELAPSYA code through the selection of hydrodynamic components. These components serve as building blocks and provide a high degree of flexibility for the user to simulate a variety of LWR and other thermal-hydraulic systems. RELAPSYA uses the concept I
of hydrodynamic volumes (fluid control volumes) and hydrodynamic junctions (momentum control volumes). Hydrodynamic volumes contain averaged state properties. These include primary dependent variables (p, X, X, pB'
- u. P) and auxiliary variables (ag, a,T T, X, and volume g
averaged phasic velocities). The wall heat transfer rate, vapor generation I
rate, and mechanical energy dirsipation terms are accounted for within these volumes. Hydrodynamic junctions contain the averaged phasic velocities and properties required for the flux terms at the open ends cf hydrodynamic volumes. Fluid inertia, convected momentum, gravitational forces, interphase I
drag, wall friction, localized pressure losses, and critical flow are accounted for within hydrodynamic junctions.
Twelve types of hydrodynamic components are available in RELAPSYA for constructing a simulation model of a thermal-hydraulic system. These components are the vehicle for entering input data, selecting user options, and applying the hydrodynamic model to fluid regions within the system or I
specifying fluid system boundary conditions. Five components (SNGLVOL, SNGLJUN, PIPE, ANNULUS, and BRANCH) apply the basic hydrodynamic model to l
I internal fluid regions. Another five components (SEPARATOR, PUMP, JETPMP, VALVE, and ACCUM) modify the basic hydrodynamic model to account for unique f
I I I
I hydecdynamic pt.cnonenn within esperators, ccntrifussi cnd jst perpa, vslvss, and accumulators. Finally, two components (TMDPVOL and TMDPJUN) allow the user to specify hydrodynamic boundary conditions for a system model.
I The RELAPSYA code also contains models for simulating other transient processes pertinent to light-water reactors and thermal-hydraulic systems.
These include the following:
Point reactor kinetics with decay heat, moderator density, Doppler, a.
and SCRAM reactivity efftets.
I b.
Thermal power sources and sinks.
I Dynamic fuel rod behavior with swelling, rupture, and metal-water c.
reaction.
d.
Conduction heat transfer within solid materials.
I Convective heat transfer between solid materials and the e.
surrounding fluid.
f.
Radiation heat transfer between participating solid material I
surfacer.
g.
Generalized control systems and trip logic, RELAP5YA Version 18V has been developed from the RELAPS HOD 1 Cycle 18 code that was originally developed by EG&G Idaho, Inc., under USNRC sponsorship, and publicly released. Substantial modifications have been made to RELAPS M001 in order to:
I Extend and improve upon the code simulation capabilities.
a.
J b.
Provide options that conform to 10CFR50, Appendix K requirements.
Correct certain errors identified by YAEC and by EC&G Idaho, Inc.,
c.
in their updates for Cycle 19 to Cycle 29.
I I i
Tha majcr modificcticns cra id:ntifisd in Tablo 1.3.1.
R;forsnce 1-2 provldes a complete description of the RELAPSYA Computer Program. Its focus is primarily on modifications by YAEC to the original RELAPS MOD 1 Code.
Reference 1-3 provides a user's manual for RELAP5YA. Reference 1-4 provides I
an extensive assessment of RELAPSYA calculations compared to many separate effect tests identified in Table 1.3.2, and integral tests identified in Table 1.3.3.
This assessment establishes the viability of the RELAP5YA code to predict complex thermal-hydraulic phenomena encountered in LWR System analyses of LOCA events and other transients. References 1-5 through 1-9 contain additional information from YAEC's response to the 197 general RELAPSYA questions from the USNRC. Reference 1-10 contains YAEC's response to the 39 BWR-related questions about RELAP5YA.
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i, a,,'.
t i
e.
I is g.
m >
m
[
E[
I :
t ] )
1 j 'q :
,Uk qlE m,,.......,7, h, I
Figure 1.2-1: Vermont Yankee Reactor Vessel I 1 I
TABLE 1.2.1 Verr.sont Yankee Operating Conditions Rated Design f
Core Power (MWt) 1,593 1,664 Vessel Pressure (psia) 1,020 1,035 Vessel Water Level (inches)a 512 512 Feedwater Temperature (OF) 372 376 Feedwater Flow (MLBM/hr)b 6.40 6.72 I
Recirculation Loop Flow (MLBM/hr) 12.3 12.3 (each of two loops)
Control Rod Drive Flow (MLBM/hr) 0.03 0.03 Core Flow (MLBM/hr) 48.0 48.0 Main Steam Line Flow (MLBM/hr) 6.43 6.75 (four combined lines)
NOTES:
Inches above the top of the lower reactor vessel invert.
a.
b.
Million pounds mass per hour.
I I
I I
' I,
- I
l TABLE 1.3.1 RELAPSYA Model Development Tasks Analysis Requirement A.
Hydrodynamic Models BE, EM 1.
Interphase Drag EM 2.
Moody Two-Phase Critical Flow BE, EM 3.
Jet Pump B.
Heat Transfer Models 1.
Forced Convective Boiling BE, EM 2.
Critical Heat Flux BE, EM 3.
Rewet and Quench BE, EM 4.
Multi-Surface Radiation BE, EM 5.
Heat Transfer Logic Options EM C.
Fuel Behavior Models I
1.
Internal Gas Pressure BE, EM 2.
Rod Deformation and Rupture BE, EM 3.
Transient Gap Conductance BE, EM EM 4.
Zircaloy-Water Reactions l I I
I I
I I
I I
TABLE 1.3.2 Separate Effect Tests for RELAP5YA Model Assessment Hydrodynamic Models Assessment Cases Interphase Drag 4
Frigg Loop Tests 1
GE Level Swell Test Two-Phase Critical Flow 7
Parametric Calculations 1
Marviken Test Jet Pump 233 1/6 Scale EG&G SS Tests 2
1/6 Scale Blowdown Tests VY Jet Pump M-N Curve Accumulator 1
Maine Yankee Test Heat Transfer Models Forced Convcetive Boiling 3
Bennett Tests Critical Heat Flux 5
Columbia Tests A
GE Nine Rod Tests I
3 ORNL THTF Tests Radiation 2
Analytical Solutions 4
Parametric Calculations EM Logic Options 9
Parametric Calculations Fuel Behavior Models Metal-Water Reaction 1
Analytical Solution All Models 1
T00DEE2-EM Run I
I I
I I I
TABLE 1.3.3 Integral Tests for RELAPSYA Code Assessment Reference Thermal-Hydraulic Test Facility (ORNL-NRC) 1-4 3
Steady-State Film Boiling Tests 1-4 1
Transient Film Boiling Test I
1-4 1
Quasi Steady-State Boiloff Test 1-4 2
Reflood Tests Two Loop Test Apparatus (CE-EPRI-NRC)
Large Break With ECCS, Test 6425/2 1-4, 1-10 1-4 Large Break Without ECCS, Test 6426/1 I
Small Break With Degraded ECCS, Test 6323/1 1-4, 1-10 1-4 System Bolloff With Recovery, Test 6441/6 LOFT (EG&G-NRC)
Small Break With Pumps On, Test L3-6 1-4, 1-9 Severe Ccre Transient With Pumps Tripped, Test L8-1 1-4 1-9 Small Break With Pumps Off Test L3-1 Semiscale Test Facility (EG&G-NRC) 1-9 Single-Phase, Two-Phase, and Reflux Modes of Natural Circulation. Test Series S-NC-2 Yankee Plant (YAEC) 1-7 Reactor Coolant Pump Trip Test I
I I I
E 2.0 NUCLEAR STEAM SUPPLY SYSTEM MODEL i
I This section describes the RELAP5YA base input model for Vermont Yankee NSSS LOCA licensing analyses. This model contains a relatively detailed representation of the following:
a.
Reactor vessel, internals, and core, b.
Each independent recirculation loop including the associated bank of ten jet pumps.
l I
Feedwater lines from the check valves just outside the drywell wall c.
to the reactor vessel.
d.
Main steam lines and valves from the vessel to the turbine stop I
valves and bypass valves.
e.
Emergency Core Cooling Systems.
f.
Control systems and trip logic.
g.
Core power.
The base input model is set up for a double-ended guillotine break in I
the discharge pipo of Recirculation Loop A just upstream of the header pipe.
This case was selected because it is the current Design Basis Accident (DBA) for VYNPS (Section 6.5, Reference 1-1).
Important accident assumptions include the following:
j a.
Coincident loss of normal auxiliary power.
b.
Failure of LPCI-A injection due to the proximit*/ of the break location to the injection pipe and valve.
I Failure of the Recirculation Loop A discharge valve to close on c.
demand due to the proximity of the break location to the valve.
' I
d.
Failura of thi LPCI-B injr.cticn VSlva to cpan on d: mand (single failure criterion).
l other break sizes, locations, and single failure assumptions are modeled by making relatively minor changes to this input model. These changes will be identified and documented for each case.
I l
The Vermont Yankee NSSS model summarized below results from a considerable amount of engineering insight, sound judgements, and experience I
gained over the past six years. This includes incorporation of the following:
l Current information for the Vemont Yankee NSSS such as drawings, a.
Technical Specifications, plant performance data, and guidance from l
supporting software.
l b.
Modeling techniques found successful for simulating the response of l
many separate effects and integral system tests identified in Tables 1.3.2 and 1.3.3.
l Improvements to the model that have evolved from previous RELAPSYA c.
e analyses of the Vermont Yankee NSSS.
l 2.1 NSSS Volumes Junctions, and Heat Structures 1
Figure 2.1-1 shows the Vermont Yankee NSSS nodalization diagram for the base input model. Table 2.1.1 identifies the corresponding regions or systems, and sumraarizes the number of volumes, junctions, and heat structures used in the Vermont Yankee NSSS model. Active volumes mean all volumes except Time-Dependent Volumes (TMDPVOLs). Likewise, active junctions mean all
,5 junctions except Time-Dependent Junctions (TMDPJUNs). Active heat structures l
mean those heat structures that model axial segments of fuel rods. Passive r
heat structures are those that model reactor vessel and pipe walls, and reactor vessel internal structures. Tables 2.1.2 through 2.1.5 summarize important input parameters for these volumes, junctions, valves and heat structures.
I I l
1
I A brief explanation is providad for ths ccmpon:nt numbsring systzm in the nodalization diagram to enhance easy recognition of components. Even digit numbers are generally used for all active volumes. Odd digit numbers are generally used for all active junctions and selected, where possible, to have values between the associated volume numbers. Heat structures generally retain the number corresponding to the fluid volume with which they interact.
Blocks of numbers have been assigned for subregions, systems, and boundary conditions that are evident from Table 2.1.1.
The numbering sequence generally traces the normal flow directions, beginning from the lower plenum.
2.2 Core Power The total core power is determined by the point reactor kinetics model in RELAP5YA. Conservative input data are entered for this model in order to compute the fission power and decay heat per 10CFR50.46 Appendix K I
requirements. The model accounts for moderator void, Doppler, and SCRAM reactivity effects. Minimum plausible input values have been selected for the sample problems based upon a review of SIMULATE computer code (References 2-1 and 2-2) data for Vermont Yankee.
I All core power is conservatively assumed to be generated in the fuel, i.e., none is deposited in moderator, cladding, or passive heat structures.
I This power is distributed according to the Nodal Power Factor (NPF) entered for each active heat structure (27) that represents a portion of UO f"*1*
2 I
Each nodal power factor is the product of three terms:
NPF = F xF xF where:
i i
F
- E B"
F = Core region radial power factor.
R F = Core region axial power factor.
I l
SIMULATC computcr ccda (Rcfsrcnccs 2-1, 2-2, c.nd 2-3) dtte for Varmont Yankee Cycles 9, 10, and 11 were reviewed to determine conservative values for each term in order to achieve bounding cycle-independent values. The following values were selected from this review:
I Core Region FB R
2 Peripheral Low Power 0.315217 0.5851 1.40 Chopped Cosine Central Average Power 0.673913 1.1860 1.40 Chopped Cosine Central High Power 0.010870 1.5000 1.53 Chopped Cosine This yields a maximum local peaking factor in the high power bundles of 2.295 (1.50 x 1.53).
This value is 11.4% larger than the highest value found I
during the review. This peak occurs in axial Node 5 located between 72 and 90 inches above the bottom of the fuel zone. Finally, the power within each fuel node is distributed according to flux depression factors obtained from FROSSTEY computer code (Appendix A) results.
2.3 Trio Logic I
The Vermont Yankee model uses a combination of 93 logical and variable trips to account for significant signals that will or might occur for a I
spectrum of LOCAs. Trip setpoints are generally obtained from the Vermont Yankee Technical Specifications if available. The trips account for delays built in to certain circuits and delays due to instrument and logic circuit response times. The trip functions and signals are summarized below.
a.
Accident Initiation o
Terminate normal operation.
l Begin LOCA with loss of normal auxiliary power.
o b.
Reactor SCRAM o
RPS MG set underfrequency time, OR High drywell pressure time, OR o
o RPV low level (127 inches), OR I I
I o
RPV high proa::ura (1,055 paig), OR o
MSIV closure (<90% open).
c.
Turbine Stop Valves (Junction 549) Closure l
l l
o High drywell pressure time, OR o
Reactor SCRAM plus 20 seconds.
i l
l d.
Turbine Bypass Valves (Junction 571) Opening i
o Turbine stop valve closure plus 0.1 second, e.
MSIV (Junction 547) Closure l
1 o
RPS MG set underfrequency time, OR o
High main steam line flow (140%), OR o
RPV low low level (82.5 inches).
I f.
Recirculation Pump Motor Trips o
RPV high pressure (1,150 psig), OR o
RPV low low level (82.5 inches) plus 10 seconds, OR I
o Recirculation loop MG set underfrequency time (17 seconds).
g.
Safety / Relief and Safety Valves f
Open (psid)
Close (esid) o S/RV1 (Junction 551) 1,080.0 1,047.6 o
S/RV23 (Junction 553) 1,090.0 1.057.3 o
S/RV4 (Junction 555) 1,100.0 1,067.0 o
SV12 (Junction 557) 1,240.0 1,202.8 o
All S/RV junctions close if ADS opens.
I I
h.
Automatic Depes curizaticn Systcm (Juncticn 559) Actuaticn I
o High drywell pressure time, AND o
Current RPV low low level (82.5 inches), AND o
120-second delay, AND t
i o
At least one low pressure ECCS pump running.
Initiates on high drywell pressure time OR RPV low low level (82.5 inches).
I I
o Terminates on RPV high level (177 inches) OR low main steam line pressure (90 psia).
l i
I
- j. Low Pressure Core Spray.
Pumps are at rated speed when emergency diesels are running (13 o
seconda) plus 15 second load and startup time AND high drywell pressure time OR RPV low low level (82.5 inches) occurs.
1 Injection valves are open when high drywell pressure time OR o
RPV low low level signal arrives AND RPV pressure permissive (315 psia) plus 8 seconds occurs.
I k.
Low Pressure Coolant Injection Pumps are at rated speed when emergency diesels are running (13 l
o seconds) plus 10 second load and startup time AND high drywell pressure time OR RPV low low level (82.5 inches) occurs.
l Injection valves start to open when high dyrwell pressure time o
or RPV low low level AND RPV pressure permissive (315 psia)
I occurs. I
I o
Racirculcticn 1ccp dicchnrgs valvas etcrt to cices whtn LPCI initiation signal arrives AND RPV pressure permissive (315 psia) occurs.
1.
Rewet and Quench Model Initiation Trip o
LPCS or LPCI is ready to inject.
2.4 Emergency Core Coolint Systems 2.4.1 High Pressure Coolant Iniection (HPCI]
I The HPCI system consists of one high pressure steam turbine assembly and a constant-flow pump assembly with associated piping, valves, controls, and instrumentation. This system is capable of delivering 4250 gpm over a broad range of vessel pressures (1,120 to 75 psid vessel to containment). The steam supply to the HPCI turbine is modeled with TdDPJUN 561. Values in Table 2.4.1 are used when the HPCI stcr'.up trip becomes true. The HPCI ECC flow is modeled using TMDPJUN 701. The flow versus time values for this table account for a 20.33-second startup time followed by a 5-second ramp to 587 lbm/sec (100 F water) after the HPCI initiation signal occurs.
2.4.2 Automatic Depressurization Eystem (ADS)
The ADS is modeled as a trip valve (Junction 559) using the combined nozzle flow areas of the four safety / relief valves. When the ADS opens, the corresponding SR/Vs (if open) are quickly ramped closed. A two-phase discharge coefficient of 0.848 is used for these valves in order to better approximate their rated conditions.
2.4.3 Low Pressure Core Spray (LPCS)
The two independent core spray systems are modeled as one combined system by TMDPJUN 721. A control variable (CV721) monitors the upper plenum to wetwell pressure difference (P206-P720). This parameter is used as the independent variable in the TMDPJUN table to determine the LPCS flow after the injection valve has opened. Table 2.4.2 contains the data for this I
I component. These values h va besn cencarvctively calcetsd from cetunl VYNPS pump data. Specifically, the lower of the two-pump performance curves was used and the flow rates were reduced by 3% to allow for the estimated measurement uncertainties.
2.4.4 Low Pressure Coolant Iniection (LPCI)
I Each independent LPCI System is modeled separately since they inject into different recirculation loops. These are modeled by TMDPJUNs 741 and 761. Each has an associated control variable that monitors the injection pipe to wetwell pressure difference:
CV741 = P742 - P740 CV761 = P762 - P760 I
Each CV parameter is then used as the independent variable for the corresponding TMDPJUN table. The two tables are identical. Values for TMDPJUN 741 are given in Table 2.4.3.
The flows assume that two pumps are operating in each system and have been reduced by 150 gpm to allow for the estimated uncertainties.
2.5 Initial Conditions I
The initialization of the NSSS model has been accomplished in several steps. First, an earlier version of the NSSS model was modified as follows:
a.
Accident trips were set to false.
I b.
All passive heat structures were deleted to avoid long thermal time l
constants associated with their heatup, l
c.
The NSSS was filled with hot stagnant liquid to the desired vessel l.
level with saturated steam above at approximately 988 psia. All fluid velocities were set to zero.
I I
I d.
A pump cpud v:r;us time tcbis wra untd to ramp th1t escirculation pumps from 0 to rated speed over 50 seconds and hold constant, thereafter, A power versus time table was used to ramp the core power from zero e.
at 10 seconds to full power at 50 seconds and hold constant, I
thereaftec.
f.
Perfect separation of the two-phase mixture exiting the steam separators was imposed.
I g.
The feedwater temperature was set at the desired value. The feedwater flow rate used control variables to be set at the main I
steam line flow rate plus an amount proportional to the desired minus actual vessel water level.
An " accelerated startup" transient was then run from 0 to 150 seconds.
This allowed 100 seconds after rated pump speed and core power for the NSSS to stabilize. This provided a reasonable set of volume and junction initial conditions at full power operation. These were copied through files and added to a Vermont Yankee NSSS input deck as initial condition replacement cards.
I A series of steady-state null transient runs were made with the Vermont Yankee NSSS model. The accident conditions were set to false and the i
recirculation pump speed was held constant. These runs allowed certain input
(
parameters to be adjusted to achieve desired steady-state initial ccnditions.
Also, the steady-state initialization flag for all heat structures is activated so they are initialized to operating temperatures. Slight adjustments were made to the fuel radius in each core region until the volume average fuel temperatures matched those calculated by the FROSSTEY code (see Appendix A) for the peak axial power location.
I I I
.1
661 l l
659 l
240 657 a'a-I 655 i lb F7 l
l 234 653 l 224.
4 542 L J 551 F~7 l
232 l 52 l
651 l 222 1 223
>- J
[Le L _#
-g-v 2
e "I
T.{
544 r- - 1 1 226 ;
256 228 256 g
t_J l260 l 220 l 260 l 3
571
% 6 6 e 570 M 264 2
264 549 266 266 g
,20
_, -2 l 26s I h 54s l 26 l 9
7 9
9 539 e
g g
g g
e 547 l270 l *s l 270 l
~
s s
a s,'#
Y 7
7 7
o e
1 2
2 1
~
~
3 s
I
/ 402-1
~
l I
t,~ll i
A l5 E
i 112o a0ti.oi 3g
{
l 012 l
a a
e j
0 h
i I
l 010 l O 352 7
l
?
4 7
i o
o s
j 008 l k E
+
T
?
6 T-E 802
{ 006 l h l
006
)
p 9
4 383-1 k
h 330-1 7
004 l
004
/
4 N
k 4
A 76M 762
'I 742 E
i 2.>
i
,.0 a
t t
[
+
f i
I
[
w 3,,
310 360 32 e 313 363
[
308 356 Figure 2.1-1:
Verraout Yankee NSSS Nodalization Diagrara E
M M
M M
M M
M M
M M
M M
M M
M TABLE 2.1.1 Vermont Yankee NSSS Nodalization Sumnary Region Nodalization Number of Number of Number of or Diagram Volumes Junctions Heat Structures System Numbers Active TMDPVOL Active TMDPJUN Active Passive Reactor Pressure Vessel Lower Plenum 002 to 012 6
0 10 0
0 6
Control Rod Cuide Tubes 022 to 024-5 6
0 6
0 0
6 Core Bypass 100 to 102-9 10 0
10 0
0 10 Core Fuel Assemblies 116 Peripheral Low Power 120 to 122-9 10 0
11 0
9 10 248 Central Average Power 140 to 142-9 10 0
11 0
9 10 4 Central High Power 160 to 162-9 10 0
11 0
9 10 Upper Plenum 206 to 208 2
0 2
0 0
2 Standpipes and Separators 210 to 228 3
3 5
2 0
3 Intermediate Steam 232 1
0 2
0 0
0 Steam Dryer Assembly 234 1
0 1
0 0
1 Steam Dome 240 1
0 1
0 0
1 250 to 290 15 0
14 0
0 13 Downcomer Other Systems Fecirculation Loop A 302 to 338-1 12 0
12 0
0 12 Jet Pump Bank A 340 to 342-3 4
0 6
0 0
4 Recirculation Loop B 352 to 388-1 12 0
12 0
0 12 Jet Pump Bank B 390 to 392-3 4
0 6
0 0
4 Feedwater Lines 400 to 402-3 3
1 3
1 0
0 Main Steam Lines 540 to 550 6
1 7
0 0
5 Steam Relief and Supply 551 to 571 0
1 6
1 0
0 Containment 614 to 663 0
8 0
0 0
0 l
ECCS 700 to 763 2
4 2
4 0
0 DEG Breai 801 to 802 0
0 2
0 0
0 118 18 140 8
27 109 TOTALS
M M
M M
M M
M M
M M
M TABLE 2.1.2 Summary of VY NSSS Fluid Volumes component Area Length volume Vert. Angle Elev. Change Rough.
Hyd. Diam.
Flags Number Type (ft )
(ft)
(ft )
(degrees)
(ft)
(ft)
(ft) 2 3
Lower Plenum Region 002 SNGLVOL (70.60) 3.5160 248.24
+90.0
+3.5160 1.5E-4 1.827 00 004 BRANCH (119.89) 2.3180 277.90
+90.0
+2.3180 1.5E-4 1.389 00 006 BRANCH (158.97) 2.2285 354.38
+90.0
+2.2285 1.5E-4 1.811 00 008 SNGLVOL (81.42) 2.7726 225.77
+90.0
+2.7726 1.5E-4 1.049 00 010 SNGLVOL (81.42) 2.7726 225.77
+90.0
+2.7726 1.5E-4 1.049 00 012 BRANCH (75.42) 2.7725 209.13
+90.0
+2.7725 1.5E-4 0.635 00 Control Rod Guide Tube Region (89 Combined) 022-1 PIPE (10.27) 14.5830 149.77
+90.0
+14.5830 1.5E-4 0.2058 00 024-1 PIPE (48.88) 2.3180 113.30
+90.0
+2.3180 1.5E-4 0.6070 00 024-2 PIPE (48.88) 2.2285 108.95
+90.0
+2.2285 1.5E-4 0.6070 00 E
024-3 PIPE (48.88) 2.7726 135.54
+90.0
+2.7726 1.5E-4 0.6070 00 i
024-4 PIPE (48.88) 2.7726 135.54
+90.0
+2.7726 1.5E-4 0.6070 00 024-5 PIPE (48.88) 2.7725 135.54
+90.0
+2.7725 1.5E-4 0.6070 00 Core Bypass Region 100 BRANCH (60.51) 0.9115 55.15
+90.0
+0.9115 1.5E-4 0.6380 00 102-1 PIPE (67.07) 1.5000 100.61
+90.0
+1.5000 1.5E-4 0.2639 00 102-2 PIPE (67.07) 1.5000 100.61
+90.0
+1.5000 1.5E-4 0.2639 00 102-3 PIPE (67.07) 1.5000 100.61
+90.0
+1.5000 1.5E-4 0.2639 00 102-4 PIPE (67.07) 1.5000 100.61
+90.0
+1.5000 1.5E-4 0.2639 00 102-5 PIPE (67.07) 1.5000 100.61
+90.0
+1.5000 1.5E-4 0.2639 00 102-6 PIPE (67.07) 1.5000 100.61
+90.0
+1.5000 1.5E-4 0.2639 00 102-7 PIPE (67.07) 1.5000 100.61
+90.0
+1.5000 1.5E-4 0.2639 00 102-8 PIPE (67.07) 1.5000 100.61
+90.0
+1.5000 1.5E-4 0.2639 00 102-9 PIPE (63.32) 1.6234 102.80
+90.0
+1.6234 1.5E-4 0.1979 00 l
l l
1
m n
n n
M rw r7 o
o o
n n
n n
n n
m r
TAGLE 2.1.2 l
(Continued) l Summary of VY NSSS Fluid Volumes Component Area Length Volume Vert. Angle Elev. Change Rough.
Hyd. Diam.
Flags Number Type (f t )
(ft)
(ft )
(denrees)
(ft)
(ft)
(ft) 2 3
Core Fuel Assembly Region Peripheral Low Power (116 Assemblies) 120 BRANCH (11.62) 0.9115 10.59
+90.0
+0.9115 1.5E-4 0.3570 00 122-1 PIPE (12.75) 1.5000 19.12
+90.0
+1.5000 5.0E-6 0.04461 00 l
122-2 PIPE (12.75) 1.5000 19.12
+90.0
+1.5000 5.0E-6 0.04461 00 l
122-3 PIPE (12.75) 1.5000 19.12
+90.0
+1.5000 5.0E-6 0.04461 00 122-4 PIPE (12.75) 1.5000 19.12
+90.0
+1.5000 5.OE-6 0.04461 00 l
l 122-5 PIPE (12.75) 1.5000 19.12
+90.0
+1.5000 5.0E-6 0.04461 00 1
122-6 PIPE (12.75) 1.5000 19.12
+90.0
+1.5000 5.0E-6 0.04461 00 l
122-7 PIPE (12.75) 1.5000 19.12
+90.0
+1.5000 5.0E-6 0.04461 00 122-8 PIPE (12.75) 1.5000 19.12
+90.0
+1.5000 5.0E-6 0.04461 00 122-9 PIPE (13.42) 1.6234 21.78
+90.0
+1.6234 5.0E-6 0.04461 00 Central Averate Power (248 Assemblies) 140 BRANCH (24.84) 0.9115 22.64
+90.0
+0.9115 1.5E-4 0.3570 00 142-1 PIPE (27.26) 1.5000 40.88
+90.0
+1.5000 5.0E-6 0.04461 00 y
142-2 PIPE (27.26) 1.5000 40.88
+90.0
+1.5000 5.0E-6 0.04461 00 i
142-3 PIPE (27.26) 1.5000 40.88
+90.0
+1.5000 5.0E-6 0.04461 00 142-4 PIPE (27.26) 1.5000 40.88
+90.0
+1.5000 5.0E-6 0.04461 00 142-5 PIPE (27.26) 1.5000 40.88
+90.0
+1.5000 5.0E-6 0.04461 00 142-6 PIPE (27.26) 1.5000 40.88
+90.0
+1.5000 5.0E-6 0.04461 00 142-7 PIPE (27.26) 1.5000 40.88
+90.0
+1.5000 5.0E-6 0.04461 00 142-8 PIPE (27.26) 1.5000 40.88
+90.0
+1.5000 5.0E-6 0.04461 00 142-9 PIPE (28.68) 1.6234 46.56
+90.0
+1.6234 5.0E-6 0.04461 00
M M
M M
M m
m a
e e
a e
g TABLE 2.1.2 (Continued)
Summary of VY NSSS Fluid Volumes Component Area Length Volume Vert. Angle Elev. Change Rough.
Hyd. Diam.
Flags Number Type (ft )
(ft)
(ft )
(degrees)
(ft)
(ft)
(ft) 2 3
Central High Power (4 Assemblies) 160 BRANCH (0.4007) 0.9115 0.3652
+90.0
+0.9115 1.5E-4 0.3570 00 162-1 PIPE (0.4396) 1.5000 0.6594
+90.0
+1.5000 5.0E-6 0.04461 00 162-2 PIPE (0.4396) 1.5000 0.6594
+90.0
+1.5000 5.0E-6 0.04461 00 162-3 PIPE (0.4396) 1.5000 0.6594
+90.0
+1.5000 5.0E-6 0.04461 00 162-4 PIPE (0.4396) 1.5000 0.6594
+90.0
+1.5000 5.0E-6 0.04461 00 162-5 PIPE (0.4396) 1.5000 0.6594
+90.0
+1.5000 5.0E-6 0.04461 00 162-6 PIPE (0.4396) 1.5000 0.6594
+90.0
+1.5000 5.0E-6 0.04461 00 162-7 PIPE (0.4396) 1.5000 0.6594
+90.0
+1.5000 5.0E-6 0.04461 00 162-8 PIPE (0.4396) 1.5000 0.6594 490.0
+1.5000 5.0E-6 0.04461 00 l
162-9 PIPE (0.4626) 1.6234 0.7510
+90.0
+1.6234 5.0E-6 0.04461 00 Upper Plenum Region 206 BRANCH (165.90) 1.8349 304.41
+90.0
+1.8349 1.5E 14.50 00 h
1 208 BRANCH (151.58) 1.8542 281.06
+90.0
+1.8542 1.5" 11.54 00 Standpipe Region (129 Combined) 210 SNGLVOL (25.88) 4.2292 109.46
+90.0
+4.2292 1.5E-4 0.5054 00 Separator Region (129 Combined) 220 BRANCH (77.72) 3.6667 284.98
+90.0
+3.6667 1.5E-4 0.8758 00 222 TMDPVOL (1.0E+4) 1.0E+2 1.0E+6 0.0 0.0 1.5E-4 0.0 10 224 TMDPVOL (1.0E+4) 1.0E+2 1.0E+6 0.0 0.0 1.5E-4 0.0 10 226 TMDPVOL (1.0E+4) 1.0E+2 1.0E+6 0.0 0.0 1.5E-4 0.0 10 228 BRANCH (71.72) 3.6667 284,98
+90.0
+3.6667 1.5E-4 0.8758 00 1
I
M M
M M
M m
a e
a e
a e
g TABLE 2.1.2 (Continued) l Summary of VY NSSS Fluid Volumes component Area Length Volume Vert. Angle Elev. Change Rough.
Hyd. Diam.
Flags Number Type (ft2)
(ft)
(ft )
(degrees)
(ft)
(ft)
(ft) 3 l
Intermediate Steam Region 232 BRANCH (209.14) 2.2917 479.28
+90.0
+2.2917 1.5E-4 9.970 00 Steam Dryer Region 234 BRANCH (188.44) 5.1666 973.58
+90.0
+5.1666 1.5E-4 0.5180 00 Steam Dome Region 240 BRANCH (229.21) 6.6528 1524.89
+90.0
+6.6528 1.5E-4 17.083 00
)
Downcomer Region 250 BRANCH (25.21) 5.1666 130.23
-90.0
-5.1666 1.5E-4 0.9084 00 e
252 BRANCH (13.90) 2.2917 31.865
-90.0
-2.2917 1.5E-4 0.3711 00 U$
254 BRANCH (13.90) 3.6667 50.985
-90.0
-3.6667 1.5E-4 0.3711 00 256 BRANCH (149.76) 3.6667 549.12
-90.0
-3.6667 1.5E-4 1.5540 00 260 BRANCH (95.45) 3.6667 349.99
-90.0
-3.6667 1.5E-4 0.7400 00 262 BRANCH (13.76) 3.6667 50.45
-90.0
-3.6667 1.5E-4 0.3673 00 264 BRANCH (179.92) 4.2292 760.92
-90.0
-4.2292 1.5E-4 2.2760 00 266 BRANCH (59.29) 3.6891 218.73
-90.0
-3.6891 1.5E-4 1.7020 00 268 SNGLVOL (65.22) 2.9092 189.74
-90.0
-2.9092 1.5E-4 2.6350 00 270 BRANCH (67.09) 2.7142 182.09
-90.0
-2.7142 1.5E-4 1.5900 00 274 SNGLVOL (64.76) 3.7958 245.82
-90.0
-3.7958 1.5E-4 1.6160 00 276 SNGLVOL (64.23) 3.6417 233.90
-90.0
-3.6417 1.5E-4 1.6030 00 278 SNGLVOL (61.12) 3.8542 235.58
-90.0
-3.8542 1.5E-4 1.3430 00 280 BRANCH (68.20) 2.5800 175.96
-90.0
-2.5800 1.5E-4 1.5820 00 290 SNGLVOL (57.57) 3.3575 193.30
-90.0
-3.3575 1.5E-4 1.3200 00
M M
M M
M m
a e
g g
g TABLE 2.1.2 (Continued)
Summary of VY NSSS Fluid Volumes Component Area Length Volume Vert. Angle Elev. Change Rough.
Hyd. Diam.
Flags Number Type (ft )
(ft)
(ft )
(degrees)
(ft)
(ft)
(ft) 2 3
Recirculation Loop A (Broken Loop) 302 BRANCH (3.6395) 5.6123 20.426 0.0 0.0 1.5E-4 2.153 00 304-1 PIPE (3.6395) 20.1276 73.254
-90.0
-19.0017 1.5E-4 2.153 00 304-2 PIPE (3.6395) 20.1276 73.254
-90.0
-19.0017 1.5E-4 2.153 00 l
308 BRANCH (3.6395) 9.9/80 36.316 0.0 0.0 1.5E-4 2.153 00 310 PUMP 3.6395 (13.3411) 48.555
+90.0
+4.3333 1.5E-4 2.153 00 312 BRANCH (3.6395) 9.2480 33.658 0.0 0.0 1.5E-4 2.153 00 320-1 PIPE (3.6395) 8.6667 31.542
+90.0
+8.6667 1.5E-4 2.153 00 320-2 PIPE (3.6395) 8.6667 31.542
+90.0
+8.6667 1.5E-4 2.153 00 326 BRANCH (3.6395) 8.6666 31.542
+90.0
+8.6667 1.5E-4 2.153 00 330-1 PIPE (4.2550) 17.8680 76.029 0.0 0.0 1.5E-4 2.328 00
+9.
0
+11.750
- 1. 5 E-4 0.9478 00 0
336-1 PIPE (3.5280) 11.7500 41.454 338-1 PIPE (3.5280) 15.6888 55.350
+58.943
+9.792 1.5E-4 0.8774 00 Jet Pump Bank A (10 Combined) 340 JETPMP 2.0462 4.4233 (9.0612)
-90.0
-4.4283 2.0E-6 0.5104 00 e
342-1 PIPE (2.5411) 3.5717 9.0760
-90.0
-3.5717 3.62E-6 0.5688 00 l
$ 342-2 PIPE (7.3374) 4.6358 34.0147
-90.0
-4.6358 5.25E-6 0.7870 00
)
342-3 PIPE (11.075) 4.5934 50.872
-90'.0
-4.5934 5.25E-6 1.1875 00 l
M M
M M
M M
M M
M M
M M
M M
M TABLE 2.1.2 (Continued)
Summary of VY NSSS Fluid Volumes Component Area Length Volume Vert. Angle Elev. Change Rough.
Hyd. Diam.
Flags Number Type (ft )
(ft)
{ft )
(degrees)
(ft)
(ft)
(ft) 2 3
Recirculation Loop B 352 BRANCH (3.6395) 5.6123 20.426 0.0 0.0 1.5E-4 2.153 00 354-1 PIPE (3.6395) 20.1276 73.254
-90.0
-19.0017 1.5E-4 2.153 00 354-2 PIPE (3.6395) 20.1276 73.254
-90.0
-19.0017 1.5E-4 2.153 00 358 BRANCH (3.6395) 9.9780 36.316 0.0 0.0 1.5E-4 2.153 00 360 PUMP 3.6395 (13.3411) 48.555
+90.0
+4.3333 1.5E-4 2.153 00 362 BRANCH (3.6395) 9.2480 33.658 0.0 0.0 1.5E-4 2.153 00 370-1 PIPE (3.6395) 8.6667 31.542
+90.0
+8.6667 1.5E-4 2.153 00 370-2 PIPE (3.6395) 8.6667 31.542
+90.0
+8.6667 1.5E-4 2.153 00 376 BRANCH (3.6395) 8.6666 31.542
+90.0
+8.6667 1.5E-4 2.153 00 380-1 PIPE (4.2550) 17.8680 76.029 0.0 0.0 1.5E-4 2.328 00 386-1 PIPE (3.5280) 11.7500 41.454
+90.0
+11.750 1.5E-4 0.9478 00 i
388-1 PIPE (3.5280) 15.6888 55.350
+58.943
+9.792 1.5E-4 0.8774 00 Jet Pump Bank B (10 Combined) 390 JETPMP 2.0462 4.4283 (9.0612)
-90.0
-4.4283 2.0E-6 0.5104 00 392-1 PIPE (2.5411) 3.5717 9.0760
-90.0
-3.5717 3.62E-6 0.5688 00 392-2 PIPE (7.3374) 4.6358 34.0147
-90.0
-4.6358 5.25E-6 0.7870 00 8
392-3 PIPE (11.075) 4.5934 50.872
-90.0
-4.5934 5.25E-6 1.1875 00
MU M mm e
uma e
sum um use e
muu aus em aus e
uma e
m l
TABLE 2.1.2 l
(Continued)
Summart <l VY NSSS Fluid Volumes Component Area Length Volume Vert. Angle Elev. Change Rough.
Hyd. Diam.
Flags l
Number Type (ft )
(ft)
(ft )
(degrees)
(ft)
(ft)
(ft) 2 3
Feedwater Lines (2 + 4 Combined) l 400 TMDPVDL (1.0E+5) 10.0 1.0E+6
+90.000
+10.0 1.5E-4 1000.0 11 l
402-1 PIPE (2.0441) 49.429 101.036
+14.810
+12.635 1.5E-4 1.1407 00
)
l 402-2 PIPE (2.0740) 34.466 71.483 0.000 0.0 1.5E-4 0.8125 00 l
402-3 PIPE (2.0740) 44.005 91.267
+52.937
+35.115 1.5E-4 0.8125 00 Main Steam Lines (4 Combined). Turbine and Condenser 1
542 SNGLVOL 5.6720 48.7517 (276.52)
-83.252
-43.5938 1.5E-4 1.3437 00 544 SNGLVOL 5.6720 25.4879 (144.57)
-12.600
-4.6563 1.5E-4 1.3437 00 I
546-1 PIPE 5.6720 22.9434 (130.14)
-25.463
-7.4011 1.5E-4 1.3437 00 di 546-2 PIPE 5.6720 22.9434 (130.14)
-25.463
-7.4011 1.5E-4 1.3437 00 7
548 PIPE 5.6720 216.9808 (1230.72)
+1.606
+6.0800 1.5E-4 1.3437 00 540 SNGLVOL 5.6720 25.00 (141.80) 0.0 0.0000 1.5E-4 1.3437 00 l
550 TMDPVOL (1.0E+5) 10.00 (1.0E+6)
+90.0
+10.0000 1.5E-4 1.0E+3 11 570 TMDPVOL (1.0E+5) 10.00 (1.0E+6)
+90.0
+10.0000 1.5E-4 1.OE+3 11 Primary Containment (Drywell and Wetwell) Steam Sinks 614 TMDPVOL (1.0E+5) 10.0 1.0E+6
+90.0
+10.0 1.5E-4 1.0E+3 11 616 TMDPVOL (1.0E+5) 10.0 1.0E+6
+90.0
+10.0 1.5E-4 1.0E+3 11 651 TMDPVOL (1.OE+5) 10.0 1.0E+6
+90.0
+10.0 1.5E-4 1.0E+3 11 653 TMDPVOL (1.0E+5) 10.C 1.OEt6
+90.0
+10.0 1.5E-4 1.0E+3 11 655 TMDPVOL (1.0E+5) 10.0 1.OE+6
+90.0
+10.0 1.5E-4 1.0E+3 11 657 TMDPVOL (1.0E+5) 10.0 1.OE+6
+90.0
+10.0 1.5E-4
- 1. 0E +3 11 l
659 TMDPVOL (1.0E+5) 10.0 1.0E+6
+90.0
+10.0 1.5E-4 1.0E+3 11 661 TMDPVOL (1.0E+5) 10.0 1.0E+6
+90.0
+10.0 1.5E-4 1.0E+3 11
- 663 TMDPVOL (1.0E+5) 10.0 1.0E+6
+90.0
+10.0 1.5E-4 1.0E+3 1.~
- Deactivated.
MJ
=
=
mm um aus TABLE 2.1.2 (Continued)
Summary of VY NSSS Fluid Volumes Component Area Length Volume Vert. Angle Elev. Change Rough.
Hyd. Diam.
Flags Number Type (ft )
(ft)
(ft )
(degrees)
(ft)
(ft)
(ft) 2 3
Reactor Core Isolation Cooling (RCIC). Deactivated
- 710 TMDPVOL (1.0E+5) 10.0 1.0E+6
+90.0
+10.0 1.5E-4 1.0E+3 11 Emergency Core Cooling Systems (ECCS)
HiRh Pressure Coolant Injection (HPCI) 700 TMDPVOL (1.0E+5) 10.0 1.0Et6
+90.0
+10.0 1.5E-4 1.0E+3 11 Low Pressure Core Spray (LPCS) d 720 TMDPVOL (1.0E+5) 10.0 1.0E+6
+90.0
+10.0 1.5E-4 1.0E+3 11 I
Low Pressure Coolant Injection - A Loop (LPCI-A) 1 740 TMDPVOL (1.0E+5) 10.0 1.0E+6
+90.0
+10.0 1.5E-4 1.0E+3 11 742 SNGLVOL 2.5360 25.0 (63.40) 0.0 0.0 1.5E-4 1.7968 00 1
Low Pressure Coolant Inlection - B Loop (LPCI-B) l 760 TMDPVOL (1.0E+5) 10.0 1.0E+6
+90.0
+10.0 1.5E-4 1.0E+3 11 1
762 SNGLVOL 2.5360 25.0 (63.40) 0.0 0.0 1.5E-4 1.7968 00 l
l l
l l
l l
M M
M M
M TABLE 2.1.3 Summary of VY N!'SS Junctions Component Connections Area Loss Coefficient Discharge Coefficient Flags Number Type From Vol.
To Vol.
(ft )
Forward Reverse Subcooled 2-Phase 2
Lower Plenum Region 003 SNGLJUN 00201 00400 (70.60) 0.169 0.184 1.0 1.0 00000 004-1 BRANCH 00400 02400 2.127E-3 1.500 1.500 1.0 1.0 00020 004-2 BRANCH 00401 00600 (119.89) 0.060 0.111 1.0 1.0 00000 007 SNGLJUN 00601 00800 (81.42) 0.221 0.238 1.0 1.0 00000 009 SNGLJUN 00801 01000 (81.42) 0.000 0.000 1.0 1.0 00000 011 SNGLJUN 01001 01200 (75.42) 0.022 0.005 1.0 1.0 00000 031 SNGLJUN 01201 10000 0.1877 1.492 1.494 1.0 1.0 00020 032 SNGLJUN 01201 12000 2.1195 2.312 2.798 1.0 1.0 00000 034 SNGLJUN 01201 14000 6.6786 2.203 2.865 1.0 1.0 00000 036 SNGLJUN 01201 16000 0.10772 2.203 2.865 1.0 1.0 00000 Control Rod Guide Tube Region (89 Combined) 023 SNGLJUN 02201 02400 (10.27) 0.624 0.316 1.0 1.0 00000 d,
024-1 PIPE 02401 02402 (48.88) 0.000 0.000 1.0 1.0 00000 T
024-2 PIPE 02402 02403 (48.88) 0.000 0.000 1.0 1.0 00000 024-3 PIPE 02403 02404 (48.88) 0.000 0.000 1.0 1.0 00000 024-4 PIPE 02404 02405 (48.88) 0.000 0.000 1.0 1.0 00000 037 SNGLJUN 02405 10000 2.000 1.414 1.403 1.0 1.0 00000 Core Bypass Region 100-1 BRANCH 10001 10200 60.15 0.011 0.035 1.0 1.0 00000 102-1 PIPE 10201 10202 (67.07) 0.000 0.000 1.0 1.0 00000 102-2 PIPE 10202 10203 (67.07) 0.000 0.000 1.0 1.0 00000 102-3 PIPE 10203 10204 (67.01) 0.000 0.000 1.0 1.0 00000 102-4 PIPE 10204 10205 (67.07) 0.000 0.000 1.0 1.0 00000 102-5 PIPE 10205 10206 (67.07) 0.000 0.000 1.0 1.0 00000 102-6 PIPE 10206 10207 (67.07) 0.000 0.000 1.0 1.0 00000 102-7 PIPE 10207 10208 (67.07) 0.000 0.000 1.0 1.0 00000 102-8 PIPE 10208 10209 (67.07) 0.000 0.000 1.0 1.0 00000 109 SNGLJUN 10209 20600 41.47 0.490 0.401 1.0 1.0 00000
M M
M M'
M M
M M'
M TABLE 2.1.3 (Continued)
Summary of VY NSSS Junctions Component Connections Area Loss Coefficient Discharge Coefficient Flags Number Type From Vol.
To Vol.
(ft )
Forward Reverse Subcooled 2-Phase 2
Core Fuel Assembly Region Peripheral Low Power (116 Assemblies) 120-1 BRANCH 12001 10001 0.3714 5.415 5.415 1.0 1.0 00020 120-2 BRANCH 12001 12200 8.0230 0.437 0.543 1.0 1.0 00000 122-1 PIPE 12201 12202 (12.748)
'1.085 1.085 1.0 1.0 00000 122-2 PIPE 12202 12203 (12.748) 1.085 1.085 1.0 1.0 00000 122-3 PIPE 12203 12204 (12.748) 1.085 1.085 1.0 1.0 00000 122-4 PIPE 12204 12205 (12.748) 1.085 1.085 1.0 1.0 00000 122-5 PIPE 12205 12206 (12.748) 1.085 1.085 1.0 1.0 00000 122-6 PIPE 12206 12207 (12.748) 1.085 1.085 1.0 1.0 00000 122-7 PIPE 12207 12208 (12.748) 1.085 1.085 1.0 1.0 00000 122-8 PIPE 12208 12209 (12.748) 1.085 1.085 1.0 1.0 00000 129 SNGLJUN 12209 20600 10.058 0.909 0.909 1.0 1.0 00000 Central Average Power (248 Assemblies) 140-1 BRANCH 14001 10001 0.7940 4.322 4.322 1.0 1.0 00020 1
140-2 BRANCH 14001 14200 17.1520 0.437 0.543 1.0 1.0 00000 Y 142-1 PIPE 14201 14202 (27,255) 1.085 1.085 1.0 1.0 00000 142-2 PIPE 14202 14203 (27.255) 1.085 1.085 1.0 1.0 00000 142-3 PIPE 14203 14204 (27.255) 1.085 1.085 1.0 1.0 00000 142-4 PIPE 14204 14205 (27.255) 1.085 1.085 1.0 1.0 00000 142-5 PIPE 14205 14206 (27.255) 1.085 1.085 1.0 1.0 00000 142-6 PIPE 14206 14207 (27.255) 1.085 1.085 1.0 1.0 00000 142-7 PIPE 14207 14208 (27.255) 1.085 1.085 1.0 1.0 00000 142-8 PIPE 14208 14209 (27.255) 1.085 1.085 1.0 1.0 00000 149 SNGLJUN 14209 20600 21.504 0.909 0.909 1.0 1.0 00000
m m
M M
M M
M M
M M
M M
M M
M M
M TABLE 2.1.3 (Continued)
Sumary of VY_ NSSS Junctions Compenent Connections Area Loss Coefficient Discharte Coefficient Flans _
Number Type From Vol.
To Vol.
(ft )
Forward Reverse Subcooled 2-Phase 2
Central High Power (4 Assemblies) 160-1 BRANCH 16001 10001 0.01281 4.247 4.247 1.0 1.0 00020 160-2 BRANCH 16001 16200 0.2766 0.437 0.543 1.0 1.0 00000 162-1 PIPE 16201 16202 (0.4396) 1.085 1.085 1.0 1.0 00000 162-2 PIPE 16202 16203 (0.4396) 1.085 1.005 1.0 1.0 00000 162-3 PIPE 16203 16204 (0.4396) 1.085 1.085 1.0 1.0 00000 162-4 PIPE 16204 16205 (0.4396) 1.085 1.085 1.0 1.0 00000 162-5 PIPE 16205 16206 (0.4396) 1.085 1.085 1.0 1.0 00000 162-6 PIPE 16206 16207 (0.4396) 1.085 1.085 1.0 1.0 00000 162-7 PIPE 1620' 16208 (0.4396) 1.085 1.085 1.0 1.0 00000 162-8 PIPE 16208 16209 (0.4396) 1.085 1.085 1.0 1.0 00000 169 SNGLJUh 16209 20600 0.3468 0.909 0.909 1.0 1.0 00000 Upper Plenum Region d
206-1 BRANCH 20601 20800 (151.58) 0.043 0.007 1.0 1.0 00000 7
208-1 BRANCH 20801 21000 25.681 0.415 0.688 1.0 1.0 00000 Standpipe and Separator Regions (129 Combined) 220-1 BRANCH 21001 22000 25.8809 1.216 1.216 1.0 1.0 00000 220-2 BRANCH 22001 22800 77.1200 0.000 0.000 1.0 1.0 00000 223 TRPVALVE 22801 22200 31.1143 1.9137 1.4091 1.0 1.0 00000 225 TMDPJUN 22400 23200 31.1143 NA NA NA NA NA 227 TMDPJUN 22600 26400 58.0618 NA NA NA NA NA 229 TRPVALVE 22801 23200 31.1143 1.9137 1.4091 1.0 1.0 00000 231 TRPVALVE 22800 26000 58.0618 1.6263 1.6263 1.0 1.0 00000
M M
M M
M M
M M
M M
M M
M
'M M
M M
M TABLE 2.1.3 (Continued)
Summary of VY NSSS Junctions Component Connections Area Loss Coefficient Discharge Coefficient Flags Number Type From Vol.
To Vol.
(ft )
Forward Reverse Subcooled 2-Phase 2
Intermediate Steam Region 232-1 BRANCH 23201 23400 108.19 1.196 1.1877 1.0 1.0 00000 232-2 BRANCH 23200 25600 149.76 0.123 0.0810 1.0 1.0 00000 Steam Dryer Region 234-1 BRANCH 23401 24000 37.33 0.9851 0.7264 1.0 1.0 00000 Steam Dome Region 240-1 BRANCH 24000 25000 (25.21) 0.345 0.792 1.0 1.0 00000 Downcomer Region 250-1 BRANCH 25001 25200 (13.90) 0.184 0.201 1.0 1.0 00000 252-1 BRANCH 25201 25400 (13.76) 0.000 0.000 1.0 1.0 00000 tj 254-1 BRANCH 25401 26200 (13.76) 0.004 0.001 1.0 1.0 00000 l
e 8 256-1 BRANCH 25601 26000 (95.45) 0.151 0.132 1.0 1.0 00000 l
260-1 BRANCH 26001 26400 (95.45) 0.2204 0.2347 1.0 1.0 00000 l
262-1 BRANCH 26201 26400 (13.76) 0.8529 0.4618 1.0 1.0 00000 i
264-1 BRANCH 26401 26600 (59.29) 0.265 0.450 1.0 1.0 00000 266-1 BRANCH 26601 26800 (59.29) 0.008 0.027 1.0 1.0 00000 270-1 BRANCH 26801 27000 (65.22) 0.001 0.004 1.0 1.0 00000 270-2 BRANCH 27001 27400 (64.76) 0.005 0.001 1.0 1.0 00000 275 SNGLJUN 27401 27600 (64.23) 0.000 0.000 1.0 1.0 00000 l
I 277 SNGLJUN 27601 27800 (61.12) 0.010 0.002 1.0 1.0 00000 280-1 BRANCH 27801 28000 (61.12) 0.011 0.036 1.0 1.0 00000 280-2 BRANCH 28001 29000 (57.57) 0.0782 0.0244 1.0 1.0 00000
V G
M M
M M
Ik V
M M
M M
P TABLE 2.1.3 (Continued)
Summary of VY WSSS Junctions Component Connections Area Loss Coefficient Discharge Coefficient Flags Number Type From Vol.
To Vol.
(ft )
Forward Reverse Subcooled 2-Phase 2
Recirculation Loop A (Broken Loop) 302-1 BRANCH 28001 30200 (3.6395) 0.159 0.664 1.0 1.0 00000 302-2 BRANCH 30201 30400 (3.6395) 0.168 0.168 1.0 1.0 00000 304-1 PIPE 30401 30402 (3.6395) 0.000 0.000 1.0 1.0 00000 305 SNGLJUN 30402 30800 2.670 0.264 0.264 1.0 1.0 00000 310-1 PUMP INLET 30801 (31000) 1.8045 0.240 0.240 1.0 1.0 00000 310-2 PUMP OUTLET (31001) 31200 (3.6395) 0.000 0.000 1.0 1.0 00000 313 MTRVALVE 31201 32000 2.047 0.264 0.264 1.0 1.0 00100 320-1 PIPE 32001 32002 (3.6395) 0.000 0.000 1.0 1.0 00000 326-1 BRANCH 32002 32600 1.6669 0.124 0.324 1.0 1.0 00000 327 TRPVALVE 32601 33000 3.6395 0.696 1.288 1.0 1.0 00000 335 SNGLJUN 33001 33600 (3.528) 0.810 0.855 1.0 1.0 00000 l
337 SNGLJUN 33601 33800 (3.528) 0.833 u.730 1.0 1.0 00000 l
Jet Pump Bank A (10 Combined) l 340-1 JETPMP-DR 33801 34000 0.5304 (a)
(b) 1.0 1.0 00000 d>
340-2 JETPMP-SU 27001 34000 2.4006 (a)
(b) 1.0 1.0 00000 7
340-3 JETPMP-TH 34001 34200 2.0462 (a)
(b) 1.0 1.0 00000 (a) FDK1 to FDK7:
0.04129 0.1174 0.8733 0.1268 0.6400 0.8998 0.0 (b) RDK1 to RDK7:
0.90000 0.2441 1.0000 0.4695 1.0000 0.0 0.0 342-1 PIPE 34201 34202 4.1539 0.1501 0.1065 1.0 1.0 00000 342-2 PIPE 34202 34203 11.0753 0.0335 0.0305 1.0 1.0 00000 343 SNCLJUN 34203 00601 11.0753 4.8000 0.3653 1.0 1.0 00000 l
M M
M M
M M
M M
M M
M M
M M
M M
M M
M TABLE 2.1.3 (Continued)
Summary of VY_N ES_ Junctions Component Connections Ares Loss Coefficient Discharge Coefficient Flats Number Type From Vol.
To Vol.
(ft )
Forward Reverse Subcooled 7-Phase 2
Recirculation Loop B (Intact Loop) 352-1 BRANCH 28001 35200 (3.6395) 0.159 0.664 1.0 1.0 00000 352-2 BRANCH 35201 35400 (3.6395) 0.168 0.168 1.0 1.0 00000 354-1 PIPE 35701 35402 (3.6395) 0.000 0.000 1.0 1.0 00000 355 SNGLJUN 35402 35800 2.670 0.264 0.264 1.0 1.0 00000 360-1 PUMP INLET 35801 36000 1.8045 0.240 0.240 1.0 1.0 00000 360-2 PUMP OUTLET 36001 36200 (3.6395) 0.000 0.000 1.0 1.0 00000 363 MTRVALVE 36201 3/000 2.047 0.264 0.264 1.0 1.0 00100 370-1 PIPE 37001 37002 (3.6395) 0.000 0.000 1.0 1.0 00000 376-1 BRANCH 37002 37600 1.6669 0.124 0.124 1.0 1.0 00000 377 TRPVALVE 37601 38000 3.6395 0.696 1.288 1.0 1.0 00000 385 SNGLJUN 38001 38600 (3.528) 0.810 0.855 1.0 1.0 00000 387 SNGLJUN 38601 38800 (3.528) 0.833 0.730 1.0 1.0 00000 Jet Pump Bank B (10 Conbined) 390-1 JETPNP-DR 38801 39000 0.5304 (a)
(b) 1.0 1.0 00000 d>
390-2 JETPMP-SU 27001 39000 2.4006 (a)
(b) 1.0 1.0 00000 7
390-3 JETPKP-TH 39001 39200 2.0462 (a)
(b) 1.0 1.0 00000 (a) FDK1 to FDK7:
0.04129 0.1174 0.8733 0.1268 0.6400 0.8998 0.0 (b) RDK1 to RDK7:
0.90000 0.2441 1.0000 0.4695 1.0000 0.0 0.0 392-1 PIPE 39201 39202 4.1539 0.1501 0.1065 1.0 1.0 00000 392-2 PIPE 39202 39203 11.0753 0.0335 0.0305 1.0 1.0 00000 393 SNGLJUN 39203 00601 11.0753 4.8000 0.3653 1.0 1.0 00000 Feedwater Lines (2 and 4 Combined) 401 TMDPJUN 40000 40200 2.044 NA NA NA NA NA 402-1 PIPE 40201 40202 2.044 0.666 0.730 1.0 1.0 00000 402-2 PIPE 40202 40203 2.074 0.308 0.308 1.0 1.0 00000 403 SNGLJUN 40203 26400 2.074 1.453 0.970 1.0 1.0 00000
7 n
W W
U n
v n
__ m _
W m_
m TABLE 2.1.3 (Continued)
Sununary of VY NSSS Junctions Component Connections Area Loss Coefficient Discherme Coefficient Flags Number Type From Vol.
To Vol.
(ft )
Forward Reverse Subcooled 2-Phase 2
Main Steam Lines (4 Combined)_
541 SNGLJUN 25001 54200 5.672 0.2944 1.0944 1.0 1.0 00000 543 SNGLJUN 54201 54400 (5.672) 0.3109 0.3109 1.0 1.0 00000 545 SNGLJUN 54401 54600 (5.672) 0.1623 0.1623 1.0 1.0 00000 546-1 PIPE 54601 54602 1.665 0.1708 0.4488 1.0 1.0 00000 547 MTRVALVE 54602 54800 5.672 1.5436 1.5436 1.0 1.0 00100 539 SNGLJUN 54801 54000 5.672 0.0000 0.0000 1.0 1.0 00000 549 MTRVALVE 54001 55000 5.672 0.6187 0.6187 1.0 1.0 00100 Steam Relief and Supply 551 MTRVALVE 54401 65100 0.09945 0.0 0.0 1.0 0.848 00100 553 MTRVALVE 54401 65300 0.19890 0.0 0.0 1.0 0.848 00100 l
555 MTRVALVE 54401 65500 0.09945 0.0 0.0 1.0 0.848 00100 557 MTRVALVE 55401 65700 0.20360 0.0 0.0 1.0 0.848 00100 559 TRPVALVE 54401 65900 0.39780 0.0 0.0 1.0 0.848 00100 561 TMDPJUN 54401 66100 0.49870 NA NA NA NA NA
- 563 TMDPJUN 54401 66300 0.49870 NA NA NA NA NA 571 MTRVALVE 54001 57000 1.01000 0.0 0.0 1.0 1.000 00100 Reactor Core Isolation Cooling (RCIC). Deactivated
- 711 TMDPJUN 71000 40200 0.7610 NA NA NA NA NA
- Deactivated.
TABLE 2.1.3 (Continued)
Summary of VY NSSS Junctions Component Connections Area Loss Coefficient Discharge Coefficient Flags __
Number Type From Vol.
To Vol.
(ft )
Forward Reverse Subcooled 2-Phase 2
Emergency Cor Coolinn Systems (ECCS)
High Pressure Coolant Injection (HPCI) 701 TMDPJUN 70100 40200 0.7610 NA NA NA NA NA Low Pressure Core Spray (LPCS). (2 Combined) 721 TMDPJUN 72000 20600 31.76 NA NA NA NA NA Low Pressure Coolant Injection - A Loop (LPCI-A) 741 TMDPJUN 74000 74200 2.536 NA NA NA NA NA 743 MTRVALVE 74201 32600 2.536 0.0 0.0 1.0 1.0 00100 Low Pressure Coolant Injection - B Loop (LPCI-B) 1 761 TMDPJUN 76000 76200 2.536 NA NA NA NA NA
- i" 763 MTRVALVE 76201 37600 2.536 0.0 0.0 1.0 1.0 00100 1
Pipe Breaks 801 TRPVALVE 32601 61400 3.6395 0.0 0.0 1.0 1.0 10100 802 TRPVALVE 33000 61600 3.6395 0.0 0.0 1.0 1.0 10100
m W
W W
M M
M M
M M
M M
M M
M M
M M
M TABLE 2.1.4 Summary of VY NSSS Model Valves valve Initial Trips Stroke Number Type Position Open Close Time (see)
Description Steam Separator Region 223 Trip Opened 501 0.0 Steady-State Separator Exit 229 Trip Closed 502 0.0 Transient Separator Top Path 231 Trip Closed 502 0.0 Transient Separator Skirt Path Recirculation Loops 313 Motor Opened 501 647*
33.0 Recirculation Loop A Discharge Valve 327 Trip opened 501 0.0 DEC Pipe Severance 801 Trip Closed 502 0.0 Upstream Pipe Break 802 Trip Closed 502 0.0 Downstream Pipe Break 363 Motor Opened 501 647 33.0 Recirculation Loop B Discharge Valve Main Steam Lines 1 547 Motor Opened 501 531 10.0 4 Main Steam Isolation Valves T 549 Motor Opened 501 517 0.1 4 Turbine Stop Valves Steam Relief 551 Motor closed 660 661 0.03 1 Safety / Relief Valve 553 Motor Closed 662 663 0.03 2 Safety / Relief Valves 555 Motor Closed 664 665 0.03 1 Safety / Relief Valve 557 Motor closed 566 567 0.20 2 Safety Valves 559 Trip Closed 655 0.0 Auto Depressurization System 571 Motor Closed 518 503 0.60 10 Turbine Bypass Valves ECCS 743 Motor Closed 547*
503 24.0 LPCI-A Injection Valve 763 Motor Closed 547**
503 24.0 LPCI-B Injection Valve
- Deactivated for DEG discharge pipe breaks.
- Single failure for large recirculation pipe breaks.
m M
M M
M M
M M
M M
M M
M M
M M
M M
W TABLE 2.1.5 Summary of VY NSSS Heat Structures SS-CHF Component Geom.
Mesh Number Type Points Type Volume (ft )
Left Right Left Right Flag Lower Plenum Region 0021 SPlf.SEG.
2 C-Steel 105.643 8.5417 9.0486 002 0
15 0041 CYL 2
C-Steel 77.966 6.1775 6.6800 004 0
15 0061 CYL 2
C-Steel 74.993 7.1140 7.6165 006 0
15 0081 CYL 2
S-Steel 16.306 6.6563 6.7954 008 0
15 0101 CYL 2
S-Steel 16.297 6.6563 6.7953 010 0
15 0121 CYL 2
S-Steel 46.882 6.6563 7.0491 012 0
15 Control Rod Guide Tube Region (89 Combined) 0221 CYL 2
S-Steel 173.770 0.21230 0.29610 022-1 0
15 0241-1 CYL 2
S-Steel 17.368 0.43345 0.46335 024-1 004 15 g
0241-2 CYL 2
S-Steel 16.704 0.43345 0.46335 024-2 006 15 s
0241-3 CYL 2
S-Steel 20.773 0.43345 0.46335 024-3 008 15 0241-4 CYL 2
S-Steel 20.773 0.43345 0.46333 024-4 010 15 0242 CYL 2
S-Steel 28.703 0.43345 0.47420 024-5 012 15 Core Bypass Region (Shroud. 89 Control Rods. 30 Instrument Tubes) 1001 CYL 2
S-Steel 8.020 6.8125 7.0151 100 0
15 1021-1 CYL 2
S-Steel 13.060 6.8125 7.0130 102-1 0
15 1021-2 CYL 2
S-Steel 13.060 6.8125 7.0130 102-2 0
15 1021-3 CYL 2
S-Steel 13.060 6.8125 7.0130 102-3 0
15 1021-4 CYL 2
S-Steel 13.060 6.8125 7.0130 102-4 0
15 1021-5 CYL 2
S-Steel 13.060 6.8125 7.0130 102-5 0
15 1021-6 CYL 2
S-Steel 13.060 6.8125 7.0130 102-6 0
15 1021-7 CYL 2
S-Steel 13.060 6.8125 7.0130 102-7 0
15 1021-8 CYL 2
S-Steel 13.060 6.8125 7.0130 102-8 0
15 1021-9 CYL 2
S-Steel 31.312 6.8125 7.0130 102-9 0
15
m aema ea maeg a
g g
TABLE 2.1.5 (Continued)
Summary of VY NSSS Heat Structures Material Coordinates (ft)
Boundary Volumes Component Geom.
Mesh 3
Number Type Points Type Volume (ft )
Left Right Left RlRht Flan I
l Core Fuel Assembly Regions i
Peripheral Low Power Fuel Rods (116 Bundles x 62 Rods Each = 7.192 Fuel Rodsl 1221-1 CYL 7
(Note a) 13.7135 0.0 0.02011539 0 12201 15 1221-2 CYL 7
(Note a) 13.7135 0.0 0.02011539 0 12202 15 1221-3 CYL 7
(Note a) 13.7135 0.0 0.02011539 0 12203 15 1221-4 CYL 7
(Note a) 13.7135 0.0 0.02011539 0 12204 15 1221-5 CYL 7
(Note a) 13.7135 0.0 0.02011539 0 12205 15 1221-6 CYL 7
(Note a) 13.7135 0.0 0.02011539 0 12206 15 1221-7 CYL 7
(Note a) 13.7135 0.0 0.02011539 0 12207 15 g
1221-8 CYL 7
(Note a) 13.7135 0.0 0.02011539 0 12208 15 8
1221-9 CYL 7
(Note a) 4.5712 0.0 0.02011539 0 12209 15 4 Intervals UO, 1 Interval Fission Gas, 1 Interval Zircaloy Cladding Note a:
2 Peripheral Low Power Hardware (116 Assemblies) 1201 CYL 2
S-Steel 7.506 0.17855 0.23340 120 100 15 1222-1 CYL 2
Zircaloy 2.130 0.28888 0.29555 122-1 102-1 15 1222-2 CYL 2
Zircaloy 2.130 0.28888 0.29555 122-2 102-2 15 1222-3 CYL 2
Zircaloy 2.130 0.28888 0.29555 122-3 102-3 15 1222-4 CYL 2
Zircaloy 2.130 0.28888 0.29555 122-4 102-4 15 1222-5 CYL 2
Zircaloy 2.130 0.28888 0.29555 122-5 102-5 15 1222-6 CYL 2
Zircaloy 2.130 0.28888 0.29555 122-6 102-6 15 1222-7 CYL 2
Zircaloy 2.130 0.28888 0.29555 122-7 102-7 15 1222-8 CYL 2
Zircaloy 2.130 0.28888 0.29555 122-8 102-8 15 1223 CYL 2
S-Steel 8.360 1.05644 1.06311 122-9 102-9 15
m M
M M
M M
M M
M M
M M
M M
M M
M M
M TABLE 2.1.5 (Continued)
Summary of VY NSSS tleat Structures SS-CHF Component Geom.
Mesh 3
Number Type Points Type Volume (ft )
Left Right Left Right Flat Central Average Power Fuel Rods (248 Bundles x 62 Rods Each = 15.376 Fuel Rods) 1421-1 CYL 7
(Note a) 29.3176 0.0 0.02011510 0 14201 15 1421-2 CYL 7
(Note a) 29.3176 0.0 0.02011510 0 14202 15 1421-3 CYL 7
(Note a) 29.3176 0.0 0.02011510 0 14203 15 1421-4 CYL 7
(Note a) 29.3176 0.0 0.02011510 0 14204 15 1421-5 CYL 7
(Note a) 29.3176 0.0 0.02011510 0 14205 15 1421-6 CYL 7
(Note a) 29.3176 0.0 0.02011510 0 14206 15 1421-7 CYL 7
(Note a) 29.3176 0.0 0.02011510 0 14207 15 1421-8 CYL 7
(Note a) 29.3176 0.0 0.02011510 0 14208 15 1421-9 CYL 7
(Note a) 9.7725 0.0 0.02011510 0 14209 15 4 Intervals UO, 1 Interval Fission Gas, 1 Interval Zircaloy Cladding Note a:
2 Central AveraRe Power Hardware (248 Assemblies) 1401 CYL 2
S-Steel 16.048 0.17855 0.23340 140 100 15
$ 1422-1 CYL 2
Zircaloy 4.655 0.28888 0.29555 142-1 102-1 15 1422-2 CYL 2
Zircaloy 4.655 0.28888 0.29555 142-2 102-2 15 1422-3 CYL 2
Zircaloy 4.655 0.28888 0.29555 142-3 102-3 15 1422-4 CYL 2
Zircaloy 4.655 0.28888 0.29555 142-4 102-4 15 1422-5 CYL 2
Zircaloy 4.655 0.28888 0.29555 142-5 102-5 15 1422-6 CYL 2
Zircaloy 4.655 0.28888 0.29555 142-6 102-6 15 1422-7 CYL 2
Zircaloy 4.655 0.28888 0.29555 142-7 102-7 15 1422-8 CYL 2
Zircaloy 4.655 0.28888 0.29555 142-8 102-8 15 1423 CYL 2
S-Steel 17.984 1.05644 1.06311 142-9 102-9 15
m m
em W
W W
W W
mem a
mm W
W W
W TABLE 2.1.5 (Continued)
Summary of VY NSSS Heat StrW':Urm Material Coordinates (ft)
Boundary Volumes i
Component Geom.
Mesh 3
Number Type Points Type Volume (ft )
Left Right Left Right Flag l
Central High Power Fuel Rods (4 Bundles x 62 Rods Each = 248 Fuel Rods) 1621-1 CYL 7
(Note a) 0.4721 0.0 0.02009934 0 16201 15 1621-2 CYL 7
(Note a) 0.4721 0.0 0.02009934 0 16202 15 l
1621-3 CYL 7
(Note a) 0.4721 0.0 0.02009934 0 16203 15 1621-4 CYL 7
(Note a) 0.4721 0.0 0.02009934 0 16204 15 1621-5 CYL 7
(Note a) 0.4721 0.0 0.02009934 0 16205 15 1621-6 CYL 7
(Note a) 0.4721 0.0 0.02009934 0 16206 15 1621-7 CYL 7
(Note a) 0.4721 0.0 0.02009934 0 16207 15 1621-8 CYL 7
(Note a) 0.4721 0.0 0.02009934 0 16208 15 1621-9 CYL 7
(Note a) 0.1574 0.0 0.02009934 0 16209 15 Note a: 4 Intervals UO2, 1 Interval Fission Gas, 1 Interval Zirealoy Cladding Central High Power Hardware (4 Assembliesl f
I 1601 CYL 2
S-Steel 0.25900 0.17855 0.23340 160 100 15 CYL 2
Zircaloy 0.07344 0.28888 0.29555 162.1 102-1 15 f
1622-1 1622-2 CYL 2
Zircaloy 0.07344 0.28888 0.29555 162-2 102-2 15 1622-3 CYL 2
Zircaloy 0.07344 0.28888 0.29555 162-3 102-3 15 l
1622-4 CYL 2
Zircaloy 0.07344 0.28888 0.29555 162-4 102-4 15 1622-5 CYL 2
Zircaloy 0.07344 0.28888 0.29555 162-5 102-5 15 1622-6 CYL 2
Zircaloy 0.07344 0.28888 0.29555 162-6 102-6 15 1622-7 CYL 2
Zircaloy 0.07344 0.28888 0.29555 162-7 102-7 15 1622-8 CYL 2
Zircaloy 0.07344 0.28888 0.29555 162-8 102-8 15 1623 CYL 2
S-Steel 0.28848 1.05644 1.06311 162-9 102-9 15 Upper Plenum Region 2061 CYL 2
S-Steel 12.778 7.2500 7.4013 206 0
15 2081 CYL 2
S-Steel 26.560 7.2500 7.5579 208 0
15
m M
M M
M M
M M
M M
M M
M M
M M
M M
M TABLE 2.1.5 (Continued)
Summary of VY NSSS Heat Structureq SS-CHF Component Ceom.
Mesh 3
Number Type Points Type Volume (ft )
Left Right Left Right Flag Standpipe and Separator Regions (129 Combined) 2561 CYL 2
S-Steel 49.455' O 3270 0.3700 0
256 15 2601 CYL 2
S-Steel 49.455 0.5013 0.5313 0
260 15 2642 CYL 2
S-Steel 39.390 0.2147 0.2761 0
264 15 Steam Dryer Region 2341 CYL 2
S-Steel 78.930 0.2590 0.2774 234 0
15 Steam Dome Re?.lon 2401 SPH.SEG.
2 C-Steel 270.758 8.$417 9.0441 240 0
15 Downcomer Region 2501 CYL 2
C-Steel 143.425 8.5417 9.0441 250 0
15 C
2521 CYL 2
C-St. eel 72.586 8.5417 9.1128 252 0
15 2541 CYL 2
C-Steel 105.796 8.5417 9.0633 254 0
15 2621 CYL 2
C-Steel 104.706 8.5417 9.0581 262 0
15 2641 CYL 2
C-Steel 117.400 8.5417 9.0441 264 0
15 2661 CYL 2
C-Steel 102.407 8.5417 9.0441 266 0
15 2681 CYL 2
C-Steel 80.758 8.5417 9.0441 268 0
15 2701 CYL 2
C-Steel 80.787 8.5417 9.0793 270 0
15 2741 CYL 2
C-Steel 112.981 8.5417 9.0793 274 0
15 2761 CYL 2
C-Steel 108.392 8.5417 9.0793 276 0
15 2781 CYL 2
C-Steel 114.721 8.5417 9.0793 278 0
15 2801 CYL 2
C-Steel 76.800 8.5417 9.0793 280 0
15 2901 CYL 2
C-Steel 99.943 8.5417 9.0793 290 0
15
)
TAQLE 2.1.5 (Continued)
Summary of VY NSSS Heat Structures Material Coordinates (ft)
Boundary Volumes 8-"
Component Geom.
Mesh 3
Number Type Points Type Volume (ft )
Left Right Left RIKht Flat Recirculation Loop A (Broken Loop) 3021 CYL 2
S-Steel 3.872 1.0778 1.1753 302 0
15 3041-1 CYL 2
S-Steel 13.888 1.0778 1.1753 304-1 0
15 3041-2 CYL 2
S-Steel 13.888 1.0778 1.1753 304-2 0
15 3081 CYL 2
S-Steel 6.885 1.0778 1.1753 308 0
15 3101 CYL 2
S-Steel 25.332 1.0763 1.3278 310 0
15 3121 CYL 2
S-Steel 6.753 1.0778 1.1807 312 0
15 3201-1 CYL 2
S-Steel 6.328 1.0778 1.1807 320-1 0
15 3201-2 CYL 2
S-Steel 6.328 1.0778 1.1807 320-2 0
15 3261 CYL 2
S-Steel 6.328 1.0778 1.1807 326 0
15 3301 CYL 2
S-Steel 13.435 0.8303 0.9637 330 0
15 3361 CYL 2
S-Steel 12.826 0.4740 0.7156 336 0
15 3381 CYL 2
S-Steel 3.969 0.4175 0.5531 338 0
15 Jet Pump Bank A (10 Combined) 3401 CYL 2
S-Steel 1.539 0.25521 0.27604 340 274 15 i
3421 CYL 2
S-Steel 1.378 0.28440 0.30523 342-1 276 15 3422 CYL 2
S-Steel 2.995 0.48238 0.50411 342-2 278 15 3423 CYL 2
S-Steel 5.112 0.59375 0.62287 342-3 290 15 Recirculation Loop B (Intact Loop) 3521 CYL 2
S-Steel 3.872 1.0778 1.1753 352 0
15 3541-1 CYL 2
S-Steel 13.888 1.0778 1.1753 354-1 0
15 3541-2 CYL 2
S-Steel 13.888 1.0778 1.1753 354-2 0
15 3581 CYL 2
S-Steel 6.885 1.0778 1.1753 358 0
15 3601 CYL 2
S-Steel 25.332 1.0763 1.3278 360 0
15 3621 CYL 2
S-Steel 6.753 1.0778 1.1807 362 0
15 I
3701-1 CYL 2
S-Steel 6.328 1.0778 1.1807 370-1 0
15 3701-2 CYL 2
S-Steel 6.328 1.0778 1.1807 370-2 0
15 3761 CYL 2
S-Steel 6.328 1.0778 1.1807 376 0
15 3801 CYL 2
S-Steel 13.435 0.8303 0.9637 380 0
15 3861 CYL 2
S-Steci 12.826 0.4740 0.7156 386 0
15 3881 CYL 7
S Steel 3.969 0.4175 0.5%31 383 0
15
R R
V J1 T-U ~ U R
_f l f7 F1 R
T1 R
M G
n n
n TABLE 2.1.5 (Continued)
Sumnary of VY NSSS Heat Structures Material Coordinates (ft)
Boundary Volumes SS-N Component Gaom.
Mesh 3
Number Type Points Type Volume (ft )
Left Right Left Right Flat Jet Pump Bank B (10 Combined)_
3901 CYL 2
S-Steel 1.539 0.25521 0.27604 390 274 15 3921 CYL 2
S-Steel 1.378 0.28440 0.30523 392-1 276 15 3922 CYL 2
S-Steel 2.995 0.48238 0.50411 392-2 280 15 3923 CYL 2
S-Steel 5.112 0.59375 0.62287 392-3 290 15 Main Steam Lines (4 Combined) 5421 CYL 2
S-Steel 68.088 0.6718 0.7500 542 0
15 5441 CYL 2
S-Steel 35.597 0.6718 0.7500 544 0
15 5461-1 CYL 2
S-Steel 32.043 0.6718 0.7500 546-1 0
15 5461-2 CYL 2
S-Steel 32.043 0.6718 0.7500 546-2 0
15 5481 CYL 2
S-Steel 303.041 0.6718 0.7500 548 0
15
TABLE 2.401 HPCI Steam Turbine Flow Rates 4
Steam Line Pressure (psia)
Vapor Flow Rate (Ibm /sec) 90 0.0 90 18.4 165 19.6 f
450 24.7 L
550 27.1 650 29.9 750 33.2 850 37.2 1,100 47.2 L
1,135 48.1 1,135 0.0 E
eL Eu H
[
E H L
TABLE 2.4.2 LPCS Injection Velocities and Flow Rates CV721 (esid)
Liauld Velocity (ft/see)
Volumetric Flow (nom)
-10.01 0.0000 0
-10.00 0.6174 8,800 f
L 46.58 0.5612 8,000 100,13 0.4910 7,000 120.00 0.4608 6,570 145.97 0.4210 6,000
(
184.11 0.3508 5,000 r'
L 216.66 0.2806 4,000 239.39 0.2104 3,000 r
l 256.53 0.1404 2,000 I
265.97 0.0702 1,000 I
L 274.05 0.0000 0
[
E E
[
E-E -
E
TABLE 2.4.3 LPCI In_iection Velocities and Flow Rates CV741 (psid)
Liquid Velocity (ft/see)
Volumetric Flow (Rpm)
-10.01 0.000 0
-10.00 12.188 13.873 69.98 10.543 12,000 134.50 8.785 10,000 185.18 7.028 8,000 224.12 5.271 6,000 251.34 3.514 4,000 266.82 1.757 2,000 279.03 0.000 0
I I
I I
I I
I I
I I I
I I
3.0 HOT CHANNEL MODEL I
3.1 Objective and Description of the Model l
I The Vermont Yankee Hot Channel (VY HC) model is a subset of the Vermont Yankee Nuclear Steam Supply System (VY NSSS) model. The Vermont Yankee HC I
model utilizes appropriate hydraulic boundary conditions at the inlet and outlet of the high power bundle region from a Vermont Yankee NSSS calculation with the same accident assumptions.
The objective of the Vermont Yankee HC model is to allow greater detail in the investigation of the impact of fuel exposure (Burnup Study) on the outcome of postulated Loss-of-Coolant Accidents (LOCA). These advantages I
result from the comparatively low computational time associated with the Vermont Yankee HC model. This time is approximately one-third of the Vermont Yankee NSSS computational time for large breaks. The increase in detail comes in the fom of higher numbers of burnup points that can be investigated using a combination of both the Vermont Yankee NSSS and Vermont Yankee HC models than would be feasible to investigate with the Vermont Yankee NSSS model alone.
I Figure 3.1-1 shows the heat structures (stippled) and fluid volumes (clear) represented in the Vermont Yankee HC model.
I The hydrodynamic characteristics of the hot channel are obtained from 3
two parallel channels: one representing one-eighth of the hot assembly, the other representing the associated bypass region. The bypass region is modeled so that the heat transfer across the channel wall can be taken into account.
The outlets to both the bypass and the hot assembly region are represented by time-dependent volumes and junctions as shown in Figure 3.1-1.
These two volumes (TV207 and TV206) are represented by only one volume (V206) in the Vermont Yankee NSSS model. The Vermont Yankee HC model requires two volumes
,g W
because of a code constraint. The constraint requires that a time-dependent volume can be connected to no more than one junction. Therefore, the boundary conditions fed into TV206 and TV207 are identical.
j lI l
tI I
The Vermont Ycnkee HC model represents one-eighth of the hot assembly and its corresponding fraction of the bypass region. Thus, the Vermont Yankee HC model represents the ten rods (six full-size and four half-size rods) shown in Figure 3.1-2(a).
However, the Vermont Yankee HC model does not use ten heat structure geometries, since this would require excessive computation time. Instead, Rods 1 to 9 of Figure 3.1-2(a) are lumped into an average rod and Rod 10 (half-size rod) is kept as the hot rod. The arrangement is shown L
in Figure 3.1-2(b).
Hence, the vermont Yankee HC model embodies three heat structure geometries:
(A) channel (box) wall, (B) average rod, and (C) hot rod. This compromise allows radiation heat transfer (to be used in a conservative manner) while retaining sufficient radial power detail to ensure that the hottest rod is accounted for.
As in the Vermont Yankee NSSS model, the corresponding heat structure geometries in the Vermont Yankee HC model have nine axial nodes each as shown in Figure 3.1-1.
The numbering scheme is similar to that used in the Vermont L
Yankee NSSS model which has been given in Tables 2.1.1 to 2.1.5.
The only exceptions are the heat structures and hydrodynamic components shown in Table 3.1.1.
L E
The input parameters for the Vermont Yankee HC model are obtained (with 1
the exception of the radiation heat transfer input) from the Vermont Yankee NSSS input deck in the manner specified below:
s a.
The hot assembly volume flow areas are divided by 32 since Vermont L
Yankee NSSS represents four assemblies while Vermont Yankee HC represents one-eighth of one assembly. The bypass flow areas are divided by 368 (assemblies in the core) and then by eight to account for the symmetry.
b.
The length, hydraulic diameter, and roughnesses are set at the same value in the Vermont Yankee HC as in the Vermont Yarkee NSSS.
The junction flow areas are divided by 32 and 2944 as described in c.
(a) above.
E EL
I d.
Tha junction less co0fficicnts cro unehmssd.
THe Vermont Yankee HC nodal power factors for the average Rod B e.
(Figure 3.1-2(b)) are obtained from the Vermont Yankee NSSS high I
power region rod divi'ded by 62 (rods per assembly), then divided by 1
4 (assemblies in high power region), then multiplied by 7.5 (rods I
represented by average Rod (B)).
f.
Similarly, the Vermont Yankee HC nodal power factors for the hot Rod C (Figure 3.1-2(b)) are the Vet 1nont Yankee NSSS value for the i
high power region divided by 62, then by 4, then by 2 (the hot rod is half-size, Figure 3.1-2(b)).
I The radiation view factors are obtained in a two-step process:
(1) The view factors for the actual physical rod layout shown in Figure 3.1-2(a) are obtained, and then (2) these view factors are " collapsed" into the geometry implemented in the Vermont Yankee HC model which is shown in Figure 3.1-2(b).
In Step One, the planar view factors, i.e., the various Fg, where i=1 to 11 and j=1 to 11, where 11 represents the channel wall (Figure 3.1-2(a)), are computed for a given elevation utilizing the PWR Licensing Code HUXY (Reference 3.1) and the methodology described in that report. View factors between different elevations are set to zero for the sake of simplicity. By I
neglecting interplanar radiation, the heat removal from the hottest heat structure is slightly less than if such mechanism had been considered.
In Step Two, the radiation view factors previously computed for Heat Structures 1 to 11, Figure 3.1-2(a), are used to obtain the view factors for Heat Structures (A), (B), and (C), shown in Figure 3.1-2(b).
This three heat structure arrangement has been chosen because it is the minimum number of heat structures which allows for some radiation heat transfer without having the hot rod radiate directly to the channel wall. The method used to collapse the view factors is a direct application of continuity and reciprocity relations.
i In the equations that follow, the letter "F" stands for the view factor and "A" for the radiation heat transfer area. The subscripts, as in F
, are either the letters A, B, C, or numbers 1 to 11, as in Figures 3.1-2(b) and (a), respectively.
I I J
I Ths ksy rolcticns for ths these hast structura modal tre th2rofore:
1.
=AF Ag B BA 2.
=AF d
B CB 4.
Fg+Fg+FAC 5.
F BB + BC "
BA 6.
FCA + FCB CC "
Similarly, for the 11-heat structure model:
9
.0 7.
F1010 + 1011 + j =1 10j =
I 8.
F1110 +
lij
- lill "
y Now F olo, F oli, F1110, and F1111 are known from Step One (HUXY).
i i
3 By a simple comparison of Figures 3.1-2(a) and (b), it is evident that:
E FAA = F1111 FCA = F o11 l
FCB and FAB are obtained by comparing Equations (6) and (7) and I
Equations (4) and (8), respectively.
Finally, Equations (1) and (2) are used to obtain FBA and FBC-l I I I l
3.2 Boundary conditiocs and Validation
{
The hydrodynamic boundary conditions used in the Vermont Yankee HC calculation are the inlet phasic velocities and outlet pressures as indicated by the use of time-dependent junctions (TJ) and time-dependent volumes (TV) in Figure 3.1-1.
Furthermore, for the purpose of property donoring, the pressure, internal energy, and quality are specified at the bottom and top of L
the solution domain (volumes 100, 160, 206, and 207 shown in rigure 3.1-1).
These boundary conditions are read off tape from a previous Vermont Yankee NSSS calculation and utilized in the Vermont Yankee HC calculation.
[
In order to verify the methodology and the implementation of the Vermont Yankee HC model, the three Vermont Yankee NSSS calculations have been performed with the Vermont Yankee HC model, as described in this section.
In all cases, excellent agreement was observed between Vermont Yankee HC and Vermont Yankee NSSS results, for clad temperatures, qualities and heat I
transfer coefficients for the high power region. This is shown in L
Figures 4.1-15 to 4.1-26, 4.2-15 to 4.2-26, and 5.0-17 to 5.0-27.
A comparison of Vermont Yankee HC hot and average clad temperatures in Figures 4.1-21 to 22, 4.2-21 to 22, and 5.0-23 to 24 showed a very small calculated impact of radiation heat transfer on rod temperatures. Both hot and average rods utilized the same nodal power factors (those of the NSSS high y
k power region rod) in these calculations. The only difference was in the radiation view factors. The hot rods "saw" only the average rods, while the average rods "saw" both the hot rod and the channel wall. The Vermont Yankee HC hot rod temperatures matched the corresponding Vermont Yankee NSSS rod temperatures, and the average Vermont Yankee HC rods were slightly cooler, although the difference is insignificant. This was expected since Vermont Yankee NSSS rods did not account for radiation heat transfer, while the
~
Vermont Yankee HC hot rod minimized it by radiating only to the average rod and not to the channel wall.
7L I
UPPER PLENUM I
TV 207 TV 206 103 163 104 164 109 169 1_
n
-09 01
-09
-09
~
-08
'Of
-08
-08
-07
- 02
-07
-07 l
-06 OE
-06
-06 m
8' m
1 o
S
-05 h
j
'.05
-05 g
-05 ga H
s 8
I o
-04
.L -04 5
-04
?
-04 i
Ao i
a n
n N
N
~
-03
-03
-03
-03
-02
-02
-02
-02
-01
-01
-01
-01 TJ 101 TJ 161 BYPASS H-CHANNEL INLET INLET TV 100 TV 160 I
Figure 3.1-1: VY HC Axial Nodalization I -_-
m W
W W
m m
m m
m m
m m
M M
M M
m m
M f'A-CHANNELi[(Box)[WALN; tilyCHANNEL-(Box).;VALL4 g
j g
e
[i Okll 3N 45 SYMMETRY
$d}
/
i
- 9?
d, i 10:
C-i e
i s
_S j
(a)
\\
(b)
\\
- i Figure 3.1-2:
VY Rod Lay Out:
(a) Number and Location of the Rods Represented, and (b) As Modeled i
1 T_AB_L_E 3.1.1 Differences Between W NSSS and W HC Numbering Scheme for Heat Structures and Hydrodynamic Components Region Description (a) Heat Structure Number in W NSSS W HC Channel Wall
- Elevation 01-08 1622 (001-008) 1631 (001-008)
- Elevation 09 1623 (001) 1632 (001)
Average Rod in 1621 (001-009) 1621 (001-009)
Hot Assembly I
1622 (001-009)
Hot Rod in Hot Assembly Region Description (b) Hydrodynamic Component Number in VY NSSS VY HC I
Bott'un of Bypass 100010000 101000000 Bottom of Hot Assembly 160020000 161000000 Upper Plenum
- Above Bypass 206010000 104000000 I
207000000
- Above Hot Assembly 206010000 164000000 206010000 I
I I
I
I 4.0 SAMPLE PROBLEMS 1 AND 2: LARGE RECIRCULATION LOOP BREAKS This section presents licensing analysis results for a Double-Ended Guillotine (DEG) break in the discharge pipe of one recirculation loop. The accident assumptions are summarized in Table 4.0.1.
The method used complies with 10CFR50.46 and Appendix K thereto. The major assumptions are that a DEG break occurs at time 4.0E-6 seconds with a coincident loss of normal auxiliary LPCI injection into the broken A loop is lost due to the proximity of power.
the break location. LPCI injection into the B loop is lost due to the assumed failure of the LPCI-B injection valve to open on demand (Single Failure Criterion). Thus, HPCI and two LPCS Systems are available to mitigate the consequences of this accident.
I Two cases are presented. Section 4.1 describes Base Case EA wherein decreasing power from the motor-generator sets that are coasting down is provided to the recirculation pump motors until pump trip signals arrive.
Section 4.2 describes Sensitivity Case EB wherein the power to each recirculation pump motor terminates at the accident initiation time due to an additional assumed failure in each electrical system.
I I
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TABLE 4.0.1 Summary of Vermont Yankee Large Break Accident Assumptions 2
1.
DEG recirculation discharge break (2*3.624 ft ) occurs at 4.0 E-6 seconds.
2.
Loss of auxiliary power occurs at 4.0 E-6 seconds.
3.
Reactor scrams after a 0.5-second delay from first RPS signal. Scram Curve 67B-EOC is used.
I l
4.
Feedwater coasts down to 0.0 lbm/see at 5.0 seconds, i
S.
MSIVs close 10.0 seconds after isolation signal plus 0.5 second delay.
Recirculation Loop A discharge valve fails open due to proximity of break location.
- h E
6.
Recirculation pumps in A and B loops coast down with decreasing power from loss of MG sets.
7.
ADS may actuate if appropriate signals exist. Thereafter, ADS cycles open/close at 12 psid between steam line and drywell anytime ADS criteria J
are currently met.
8.
HPCI injects upon demand, and terminates on high RPV level or low steam line pressure (<90 psia).
9.
No credit for RCIC operation.
10.
Two LFCS Systems inject on demand.
11.
LPCI-A pipe severs due to proximity of break location.
12.
LPCI-B injection valve fails to open upon demand (single failure).
13.
Drywell pressure and temperature are assumed constant at 16.4 psia and 2120F for fluid sink conditions. High drywell pressure is calculated i
to occur at 0.04 seconds for this case by containment model.
14.
Wetwell pressure and temperature are assumed constant at 14.7 psia and 1650F for fluid source and sink conditions.
15.
EM point reactor kinetics initially at 1664 MWth.
16.
EM core heat transfer with XMNB = 0.5.
17.
Passive heat structures are included.
18.
Moody two-phase critical flow model is used at each break location.
19.
1971 ANS Decay Heat Standard plus 20%.
4.1 Sample Problem 1: Large Break Case EA
(
Table 4.1.1 summarizes the timing of significant events for this accident case. This table provides an aid to review the following figures that contain results. Figures 4.1-1 through 4.1-12 present Vermont Yankee
{
NSSS results for the system response. Figures 4.1-13 through 4.1-20 present Vermont Yankee NSSS results for the core response. Figures 4.1-21 through 4.1-26 present Vermont Yankee hot channel results that show this model replicates the high power bundle results obtained from the Vermont Yankee NSSS
[
model. A description of these results follows.
I Figures 4.1-1 and 4.1-2 show the reactor power and net reactivity following the accident. The reactor power begins to decrease due to negative reactivity from core voiding. Reactor scram occurs on a high drywell pressure signal at 0.55 seconds, and the control rods are fully inserted by 4.22 seconds. Beyond this time, core power follows the Appendix K decay heat values (1.2 times 1971 ANS Decay Heat Standard).
F Figure 4.1-3 shows the early feedwater and main steam line flow histories. The loss of auxiliary power to the condensate and feedwater pumps causes the feedwater flow to coastdown in 5.0 seconds. The steam line flow initially drops due to decreases in the feedwater flow, core power, and steam dome pressure. However, the high drywell pressure signal, sensed on the LPCS instruments, causes a generator and turbine trip followed by opening of the turbine bypass valves at 1.16 seconds. The MSIVs begin to close on an RPS MG E
set underfrequency condition at 3.52 seconds and are fully closed by 13.52 seconds (the slowest achievable closure rate). This terminates all main steam line flow to the turbine bypass valves.
Figure 4.1-4 shows the narrow-range water level measured with respect to the top of the enriched fuel (351.5 inches above the RPV invert). The 1evel generally declines and reaches the low-low level (82.5 inches) setpoint
{
at 13.3 seconds. Three short-term rises occur. The first level swell occurs shortly after the turbine bypass valves open at 1.16 seconds. The second level swell starts at the onset of lower plenum flashing at about
{ f
5.0 seconds. The third mild swell between 9 and 13 seconds occurs as the vessel fluid redistributes when the MSIVs are reaching full closure. The narrow-range level remains off-scale low thereafter.
I Figure 4.1-5 shows the early upstream (Junction 801) and downstream L
(Junction 802) break flow rates. These flows rapidly accelerate to two-phase
[
critical flow (Moody model) in less than 0.1 second. Rapid depressurization, k
flashing, and upstream internal choking limit the initial peak flow rates.
The downstream peak flow rate is larger than the upstream peak since it is fed i
2 h
by a larger area (Junction 335, 3.528 ft versus Junction 326, 1.667 ft ).
The peak flow rates quickly decrease as their respective F(
upstream volumes rapidly void (95% at 0.6 second). The downstream break subsequently decreases to a lower flow rate when the broken loop jet pump 2
drive nozzles (Junction 340-1, 0.53 ft ) choke. Figure 4.1-6 shows the long-term break flow histories. These continue to decline to low values as the vessel pressure declines. Beyond 206 seconds, occasional oscillations r
b occur in the downstream break flow rate when the system has reached lower This anomalous behavior was caused by the Moody model occasionally pressures.
dropping the break volume pressure below the containment backpressure and an improper specification of the containment conditions (16.4 psia, 212.0 F).
Thus, the break flow would occasionally reverse and draw subcooled water rather than steam from the containment TMDPVOL. This will be eliminated in the future by specifying saturated steam in the TMDPVOL. However, this anomaly had no significant impact since the core was well cooled beyond 190 seconds.
Figure 4.1-7 shows the steam dome pressure history throughout this L
accident. The initia117 rapid depressurization is temporarily arrested at 13.5 seconds when the MSIVs close. The vessel subsequently depressurizes to 315 psia (low pressure permissive for ECC injection valves to open) at 59.6 seconds, and to 90 psia (HPCI shutoff) at 95.2 seconds. The vessel
['~
pressure is 33.0 psia and decreasing very slowly at 300 seconds.
L Figure 4.1-8 shows the ECCS flow rates. The HPCI System begins injecting coolant at 20.4 seconds, and terminates due to low turbine pressure at 95.2 seconds. The two low pressure core spray systems commence injecting coolant at 68.2 seconds about 8 seconds after the low pressure permissive _____-____-_ _ _
I signal occurs to open the injection valve. The occasional zero ficw spikes appear to result from very short pressure spikes, calculated in the upper plenum pressure, that temporarily halt LPCS flow. However, these pressure spikes are less than a 0.2-second duretion since they do not appear in the plot file.
I Figure 4.1-9 shows the net flow rate into the NSSS. Large negative values prior to 0.6 second have been suppressed to enhance the ordinate scale I
for this accident. Note that the net inflow becomes positive shortly after LPCS injection begins at 68.2 seconds. The oscillations after 206 seconds are caused by the anomalous break flow rate.
Figure 4.1-10 shows the fluid mass within the NSSS. The minimum value occurs shortly after LPCS injection begins at 68.2 seconds. The anomalous break flow rates beyond 206 seconds have an insignificant impact on the NSSS fluid inventory. The curve shows that by 300 seconds the primary system has regained 142,769 lbm from the minimum I
inventory of 90,213 lbm.
Figures 4.1-11 and 4.1-12 show the fluid mass histories for the upper plenum, bypass, lower plenum, and Control Rod Guide Tube (CRGT) regions that surround the core.
Flashing in the lower plenum and CRGT regions at five seconds expels fluid through the broken loop jet pumps and core. This expulsion through the core causes a temporary increase in the upper plenum mass until 13.2 seconds. The liquid subsequently drains bac1r through the bypass and core to the lower vessel regions. LPCS injection into the upper I
plenum begins at 68.2 seconds. The emergency coolant preferentially drains through the bypass and peripheral low power regions against updrafting vapor.
I The drainage behavior is quite similar to results from the CCFL/ Refill System Effects Tests conducted in the large 30 degree sector Lynn Facility (Section 2.3.3, Reference 4-1).
By 97.5 seconds, sufficient hydrostatic head has developed in the bypass region to increase the drainage rate to the lower plenum and CRGT regions. Beyond 126 seconds, the bypass and upper plenum regions refill in an oscillatory manner, i.e., cycles of liquid accumulation and drainage, due to the interaction of relatively cold ECC with hot steam.
II Likewise, the lower plenum and CRGT regions begin refilling beyond 96 seconds.
'I I -
Figures 4.1-13 and 4.1-14 show the fluid mass histories for the three core fuel assembly regions. The early peaks between 4 and 37 seconds are due to fluid flashing in the lower core and vessel regions, and drainage from the upper plenum (through the upper tie plates) and bypass region (through the drilled holes in the fuel supports). These fluid masses drain to the lower plenum until af ter LpCS injection begins at 68.2 seconds. Subsequently, the peripheral low power region refills more readily than the central average and high power regions. All three regions have regained significant fluid inventories by 180 seconds.
l Figures 4.1-15 and 4.1-16 show the outer clad surface temperatures and static qualities at four elevations in the high power assemblies. The four high power assemblies remain well cooled during the first 13 seconds due to positive core flow induced by rundown of the intact loop recirculation punp and lower plenum fluid flashing. CHF occurs at 13.1 seconds and a rapid heatup occurs as the bundles dry out. The early heatup is temporarily 1
arrested between 22 and 37 seconds as a two-phase mixture flows back into these assemblies from the bypass region. The conservative heat transfer I
assumptions per Appendix K did not allow the rods to quench as might be expected from TLTA Test 6424 Run 1 (Pages K-78 through K-88, Reference 4-2).
l Subsequently, these assemblies dry out and heat up to a maximum of 1065.2 F (1066.6 F inner peak clad temperature) at the 81-inch elevation at 82.6 seconds. Liquid enters from the bypass region through the drilled holes in l
the fuel support piece to arrest the second heatup. Thereafter, these assemblies gradually cool down and are essentially quenched by 190 seconds.
i I
The last elevation to quench is located at 89 inches above the bottom of the I
fuel.
Iu Figures 4.1-17 and 4.1-18 show the corresponding heat transfer coefficients for the high power assemblies. These figures show the transition I
to low post-CHF heat transfer at 13 seconds down to values as low as 0.0028 j
Btu /see ft F (10 Btu /hr ft F) at 82 seconds and generally rising O
values thereafter. These post-CHF heat transfer values form a lower bound on those determined from TLTA Test 6423 Run 3 (Peak Power, Low Rate /High Temperature ECC) given in Figure 3-86 of Reference 4-3.
t I l
-4
-=
f
.(
I Figures 4.1-19 and 4.1-20 show outer clad surface temperature and hsaf.
transfer coef ficients calculated for each core reg' on at the 81-inch i
elevation. The low power assetnblies encounter CHF later at 38 seccnds compared to the central assemblies that reach CHF at 13 seconds. Otherwise, the temperatures are simil6r afid reflect the relative powar in each type of assembly.
Figures 4.1-21 thecugh 4.1-26 show results from the cor respondins j
I Vennont Yankee hot channel model for this accident case. These results show that the Vermont Yankee HC raodel replicates the Vermont Yankee NS3S results for the high power assemblies as discussed in Section 3.2.
I I
I I
I I
I I
I I
I I I
TABLE 4.1.1 Sequence of Events for Large Break Case EA Event Time (seconds) 1.
Break opens 4.0E-6 2.
Loss of normal auxiliary power 4.0E-6 l
3.
High drywell pressure (p >2.5 psis) 0.048 4.
Turbine stop valve starts to close (and 0.45 L
is completely closed in 0.1 secon$)
5.
Reactor scram on high drywell pressure 0.55 l
6.
Initiate turbine bypass valve opening 0.56 f
7.
Turbine bypass valve completely open 1.16 8.
Low level signal (127 inches) occurs 2.96 9.
RPS MG set underfrequency (57 Hz) 3.01 condition occurs
[
10.
MSIVs begin to close on RPS underfrequency 3.52 11.
Control rods are fully inserted 4.22 12.
Feedwater flow coasts down to zero 5.0
'13. Lower plenum flashing begins 5.0 14.
Earliest nodal critical heat flux 13.10 15.
Low low level signal (82.5 inches) occurs 13.38 16.
MSIVs are completely closed 13.52 17.
Recirculation pumps trip on 17.00 underfrequency due to loss of auxiliary power t
18.
HPCI injection begins 20.40 19.
Recirculation Loop B (intact) discharge 59.62 L
valve begins to close a
20.
LPCS injection begins 68.20 f _ __
I TABLE 4.1.1 (Continued)
Sequence of Events for Large Break Case EA I
Event Time (seconds)
I 21.
Minimum Primary System inventory 74.60 (90,213 lb) occurs 22.
Peak clad temperature occurs 82.60 I
(inner 1066.60F, outer 1065.80F) 23.
Recirculation Loop B discharge valve 92.61 I
closed 24.
HPCI flow terminated on low 95.20 pressure signal 25.
Core, including high power bundle, 190.00 is well cooled I
I I
I I
I
'I I
I I
I,
E 1
VER*10NT YFMCC NS$3 LICCNSING MODEL CASE CA: LARGC M LCCR MPPOOIX K RCSLLTS DES RECIRO LOOP O!MGC PIPC BREAK E2 X 3.64 FT21 bq iR I
O4 n
l h0-9e I
7 I
i O
4-O I
d 0.0 5'. 0 10.0 15.0 20.0 25.0 TIFE (SEC)
Figure 4.1-1:
Reactor Power History (LBLOCA-EA)
I Od I
G ga 8-I e
e-. @
s I
r:eWe j
Y l
O g
\\
g e,
O I
0.0 5.0 10.0 15.0 20.0 25.0 i
TIME ISEC)
Figure 4.1-2:
Net Reactivity (LBLOCA-EA) I
VDMONT If#EEE NS$$ LICDi$!NG MODEL CR$C ER: 1ARGE ORDW LOCR frPDO!X K PESLA.TS DEG RECIRO LOOP 0!$CHNtE PIPE BRDE (2 x 3.64 FT21 f.
j;/;\\ r- ----e~..,'
8 Qog. --.9. '.'
's, n.
O I.
's*
o
- g..; !
s l
I"
\\
e-eFECOWflTER I
e--eMRIN STEf)MLINE F
M
~i
~ H-i
(
s
$~
if
\\.
k o
I'
$ f- ~'fh H
E
\\
k bo g
28
[
E9 R
O.0 5'. 0 1d.0 1$.0 20.0 25.0
[
TIME (SEC)
Figure 4.1-3:
Feed and Main Steam Flows (LBLOCA-EA) o
[
o y.\\
4 E
o
~-
N$
t-O~
(
d>o y
dN g
h
~
"9 i
38 2
O.
8-x o
N 0.0 5.0 10.0 1$.0 20.0 25.0 TIME (SEC)
Figure 4.1-4:
Vessel Water Level (LBLOCA-EA) -
I VDtrOiT YMP4(II NS$$ LICDfSING MOOCL Cast cne vwGc Brow Lom nerootr K ncsutt$
OEG RCCIRC LOOP OISCHFFGE PIPC BRDw (2 x 3.64 FT21 I
bo TR 8
So t
mW r
3 I,
I 89 d
4-n h
lI o-eUPSTREAM BREAK I
!I o--oDHNSTREAM BREAK odo l \\
C' l l
\\
I l
O 9
u e
&ag. _.1 a ""
i I
8 m
+
... >j o
,,a i
E
.., _, ' * ------L_______-
o d:
I 6.0 8.0 10.0 O. 0 2.0
- 1. 0 TIME (SEC)
Figure,4.1-5: Early Break Flowrates (LBLOCA-EA)
I O
Go M g-i a
o 1
I 3
'I h I
tl
- e Og-45 l.
s x
Co
.o.,
g
'! l i.
=
u9 l
l ll
)
~
W e - oUPSTREAM BREAK
- l. :
e--o DWNSTREAM BREAK
- I j I
ms-O 0.0 60.0 130.0 180.0 240.0 3CD.0 TIME (SEC)
Figure 4.1-6:
Long Term Break Flowrates (LBLOCA-EA).
I
E VDtNOPff YFWWEE NSSS LITN$1NG ftXEL CASE FA tivP:t BRffK LOOR frNX>IX X RCSULTS KG RECTRO LOT
- 0*$0feGE PIPE BR W 12 X 1 64 @
[
i 3
l a
"O
~
I M
(n L
lE 0 I
W l
8=
r-s -
6*
\\
I n
o I
8 g
c D
00 60.0 120.0 180.0 20.0 300.0 TIME ISEC) l Figure 4.1-7:
Vessel Pressure History (LBLOCA-EA)
I o
k s -cHPCI TURBINE
.2_
- 2 2 -
~
I o--o HPCI PUMP
[W
- -* CORE SPRflY (2) q g_
j
~
l 59 t
R8-5 l
O 4ME
-l ~-Y I
o I
8d i
8 I
l
'~
i
.d e.g,= mere:_ =,,
,,, y TIME (SEC)
Figure 4.1-8:
ECCS Flowrates (LBLOCA-EA)
I
I VDMONT YPNGI NSSS LITNSING MODCL CRSC CA: LARCC BRCMK LOCR PPPD0!X K RESULTS q
bo 006 RCCIRO LOOP DISCHPPE PIPC BRCRK (2 X 3.64 FT21 Td 9-
-8 m
I r-r
-r---
}
Am I'
l.
c I
s n*
OT E
d H
N9 i
l O
N i
I 18'.0 240.0 300.0
' O.0 65.0 llo.O 0
TINC (SEC)
Figure 4.1-9:
Net Flowrate into NSSS (LBLOCA-EA) 8 2,
xI 1
\\
-a i
b.
t o
1 On 3
I 1
r
/
I l
/
Q,R.
I O.
I Q
0.0 65.0 120.0 18'0.0 240.0 300.0 TIME (SEC)
Figure 4.1-10:
NSSS Fluid Mass Inventory (LBLOCA-EA) ~
I
VDtrOff YftpeqI NS$$ t.lCDfSING fGEl.
CRSC Dh tARQC 8RfA m flPPD0!X K ftCSLLTS 00G RCCIRC LOOP OlsofutGC FIPC BRERK (2 X 3.64 FT2)
- o,
d x
e-e-o I
(
o dI m-eBYPRSS REGION o
e-oVPPER PLENUM E3 S
f-ii l l 1 m
@o l
F L
rp '5 0.
i i
d k'
p l
i N.
o A
, k.f h
F o
)
j.
["i o
L.i.-
..I; ]
.1.
- 9. U.
_Q
-;.g
.i..
n l
c j
ll g
~ p g
- y r
L o,
Q.%%_
t,,.
k r.0 so.O tio.O 180.0 2c.O 300.0 TIME (SEC) s l
Figure 4.1-11:
Bypass and Upper Plenum Fluid Mass (LBLOCA-EA)
J
~S$
I -*
e-eLOWER PLENUM o
o-oCRGT REGION I
w 7 -,a o
8.
-s-y E
f "j
6
"~
G "*.
/
l
'b*g
, _,., r o,#,f.V#'
.a
=%.
W oa 0.0 60.0 12.0 100.0 240.0 300.0 TIME (SEC) l Figure 4.1-12:
CRGT and Lower Plenum Fluid Mass (LBLOCA-EA) I l
E VERMONT YNSIE NS$$ t,1(ENSING f(E1.
CRSE Em LARGE BRDE t,00I frPDO!X K ftESLLTS DEC RECIRO t.00P O!SOfmE P!PC DRDW (2 X 3.61 FT21 g,
Yd l
l 8
r si a--eLOW POW REGION
- -e CEN f!VG REGION
- h," 4 l
k i
i n
I j'}"}"
Q!j g
wow- - - -
t i
0.0 65.0 IN.0 IN.0 240.0 300.0 TIME (SEC)
Figure 4.1-13: Outer and Central Core Fluid Mass (LBLOCA-EA) i 9
5 8
e, 35 I
~
E U
ha n
E
$e n\\
l 50-s-
ll k 9
I I
{
L I
0.0 6b.0 120.0 180.0 240.0 300.0 j
TIME (SEC1 Figure 4.1-14: High Power Assembly Fluid Mass (LBLOCA-EA) __
l VDMONT YAM I NS$$ t.ICDiSING 70001.
W CRSE % (ARGC ORDE LDCR APPDOIY K RESULTS 9
DEG RECIRO t.00P OISCHARGE PIPC BREN (2 X 3.64 rT2)
I R
c-
.I Bj A 4%
b I
B
.r, ga
/
,4, 4 m - e63 INCH ELEV i
O N
o
- - o81 INCH ELEV c'
g0
,t-j
- --* 99 INCH ELEV I
g
,t'
'q K
+--+ 117 INCH CLEV u
l ', %+[
)
F j., b 9
lI' it; f{.
T r !* 1 i
b 9
s 5
!M#5AuAL3 I
I IIE b@ag g
6b.0 120.0 18'0.0 240.0 300.0 0.0 JI TIME (SEC)
Figure 4.1-15: High Power Bundle Clad Temperatures (LBLOCA-EA)
I
~
a Tif 1
.E g
(
U.,
1
!5 a-c63 INCH ELEV D-h e-o 81 INCH ELEV
'l id d-j
- -* 99 INCH ELEV 8
s
- I
+--+ 117 INCH ELEV g
P L
lA o
4 g:.
l i n 3
m N
l, (,
{ j.
q l
(
i 8
I c
t 3
iL ei g.,
g g
a f
hk.
o E
2E).o 300.0 s5.0 120.0 180.0 ig TIME (SEC) o.O g
Figure 4.1-16: High Power Bundle Qualities (LBLOCA-EA) I
VDMOPC YR$IE NS$$ t.1TNSING N000.
CASE EPs UWtGC BREN LOCA pytootX K RESLLTS Oc: RcetR urr DISC *GC PIrc BRcm <2 x 3.69 rT21 I
"d c
n L
I H
e-e63 INCH ELEV iM o-o81 INCH ELEV g*
- --a 99 INCH ELEV g
+--+ 117 INCH ELEY g
t E
Lb 83-i u
N E
9 o
d
--'-~,
. j l
o I
0.0 d.0 130.0 180.0 2W'.0-300.0 TIME (SEC)
Figure 4.1-17: Long Term Heat Transfer Coefficients (LBLOCA-EA) i 2.
d' a
b 8 n
C E
80 38 e-c63 INCH ELEY 5
9 d' o-o81 INCH ELEY
- -a99 INCH ELEV h
+--+ 117 INCH ELEV g
9o 8
A s.
\\
e-
)_
g lg m, A,
_[bMP 0.0 3b.0 ed.0 W.0 12.0 M.0 TIME (SEC)
Figure 4.1-18: Degraded Heat Transfer Coefficients (LBLOCA-EA)
I' I
VDtMONT YR*CE NS$$ t.!CDe$ LNG RODCt.
CASC CA: LARGE BREAK LO@ RPPDCIX K RCSULTS DEG RCCTRO t.00P DISCHr*GC PIPC BRD* (2 X 3.64 FT21
'g o
5 R
I t
I C
M.
N>
81 Inch Elev gE
/
',4.N 7
i W AVG POWER BUNDLE E
g, 5
'a
- --* HI POWER BUNDLE g
U Vy g
g, D,,
's I
'g
~O D
Eg.
']
N E,
/
N hC,@_j I
T T
I 0.0 65.0 L20.0 180.0 240.0 L. 0 TIME (SEC)
Figure 4.1-19: Maximum Bundle Clad Temperatures (LBLOCA-EA)
I e
NU, I
8 81 Inch Elev R3 e--eLOW POWER BUNDLE S
e-oRVG POWER BUNDLE 9
- - *HI POWER BUNDLE
$3 E
I 3
E y
5 b
is h
!i I
i:
l f
l l
= = R-1 M_ j li? v4 i
'3 5
4:' -=======
0.0 60.0 120.0 180.0 240.0 300.0 g
TIME (SEC)
Figure 4.1-20:
Bundle Heat Transfer Coefficients (LBLOCA-EA) I
VDtMONT WWOTE HC LICcNSING 6 c a 6, tenac encm toca e li x ncsutts 2 x 3.64 rT21 Ocs accine tour 01m rirt enue:
b C
b
$ag.
R$
~ k f.
i e-e63 INCH ELEV
,8 ' \\,.##g e--o 81 INCH ELEV C*
D $
7 e
- - *99 INCH ELEV
,/
~%, p
+--+ 117 INCH ELEY i
b
- g H
y, m.)
i,r%
L D
ii r
- r.
hI\\
4 l
I I
9
/
A J
e AT L
J R
6b.0 1$.0 lb.0 240.0
- 2. 0 0.0 TIME (SEC) l Figure 4.1-21: VY-HC Avg Rod Clad Teuperatures (LBLOCA-EA) a N_
1 09
- Pf\\%
'?Q g
.n j' '. af#
o -eB3 INCH ELEY s
Eo
!/
+
e-o 81 INCH ELEV N
(
wl
- - *99 INCH ELEV 7
h
?,/
', ?(h H
+--+ 117 INCH ELEV n
e J
y, m/
c i
w-Mal(N l y, 3
iin
/
e
.q-g s
f/hM h
'8^T r
n 0.0 80.0 130.0 180.0 240.0 E.0 TIME (SEC)
Figure 4.1-22: VY-HC Hot Rod Clad Teciperatures (LBLOCA-EA) I
E VDtMONT YRPECC HC LICCN$1NG N00Q.
CRSC CR LMtGC BRCRK LOCR frPD0!X K RCSULTS OCG RECIRC LOOP 0150fftGC PIPC BRCN 82 X 3,64 FT21 I
U C
G E
lI l
I
[
m--e63 INCH ELEV u
e - o81 INCH ELEV sEo
- --* 99 INCH ELEY 4
S
+--+ 117 INCH ELEV 1
w 4,
l o
E U-k']
5 th l
e I
s t
I l
e
.f I
$ 3 59 l
0.0 e0.0 120.0 180.0 240.0 300.0 TIME (SEC)
Figure 4.1-23: VY-HC Avg. Rod Heat Transfer Coefs (LBLOCA-EA) l I
C I l n~
1 s
ga e-e63 INCH ELEV s4
' e--o 81 INCH ELEV 9
- --4 99 INCH ELEY
'E
+--+ 117 INCH ELEV y,
e i
Esq l
'I m
m u
,I i
s j
j e
L!
! i fi k Ea
==
,3 i
=~m
=
0.0 80.0 12).O 180.0 240.0 300.0 TIME (SEC)
Figure 4.1-24: VY-HC Hot Rod Heat Transfer Coefs (LBLOCA-EA) lI '
VERMONT Yf90Tr HC LICDtS116 fGE2.
cast ER umet Sacar tocn neeocix x Resutis DCG RECIRC LOOP OfsCHRRGC PIPC BRCnK (2 X 3.64 772)
I 2
C l
cs en I
B
- -e63 INCH ELEY I
F"
- - o81 INCH ELEY k
m- *99 INCH ELEV
+--+ 117 INCH ELEY I
2i8 I
- h. O I
s 9
'g W--
a l
I l,
13w I
m-
.-__7 0.0 3$.0 60.0 95.0 130.0 150.0 TIME (SEC)
Figure 4.1-25: VY-HC Degraded Heat Transfer Coefs (LBLOCA-EA) l 8
0 n.,
e-e63 INCH ELEV I.' l '
i e-o81 INCH ELEV I
I
- -*99 INCH ELEV I
U.
I', j,
{
~
i
+--+ 117 INCH ELEY l
d- --g%
')
l
]
I e
i l'
$ ". 1 g
ll Y,)
Y M
n 1
I o-n p
h I
r
}.
2j pj h
" oINb l
E =.l ik
'9 i
(
0.0 80'.0 131.0 180.0 240.0 300.0 i
}
TIME (SEC)
Figure 4.1-26: VY-HC High Power Bundle Qualities (LBLCCA-EA) I
4.2 Sample Problem 2: Large Break Case EB Sensitivity Case EB is identical to Base Case EA except for two changes. For this case, we have assumed that electrical power to each recirculation loop pump motor terminates at the accident initiation time.
This assumes that an additional failure has occurred in the Electrical-Mechanical System associated with each Recirculation Pump System.
Certain high-grade components in these systems are not classified as safety-related nor environmentally-qualified at the present time. A 4
preliminary engineering review indicates that these components are either qualifiable, similar to existing qualified equipment, or not likely to be adversely affected during the early stages of a LOCA prior to the arrival of pump trip signals (Reference 4-4).
Thus, we believe their failure is highly unlikely. However, given their current classification, Case EB examines l
whether earlier freewheeling pump conditions in the intact and broken loops would substantially alter the core flow coastdown induced by the intact loop recirculation pump and/or the flow rates out the broken recirculation loop.
The results for Case EB show very minor changes compared to Case EA.
These are discussed below.
Table 4.2.1 summarizes the timing of significant ovents for Case EB.
A comparison of these event times to those for Case EA shows the following I
differences:
b The low low level signal and mid-core CHF occurred about 1 second a.
p k
earlier for Case EB.
F k
b.
LpCS injection and recirculation Loop B discharge valve closure occur about 3 seconds earlier due to a slightly faster F
depressurization of the vessel.
c.
Minimum vessel inventory occurs 4.& seconds earlier and is 927 lbm L
lower.
d.
Maximum clad outer surface temperature occurs 4.6 seconds earlier I
and is 3.2 F lower. The peak clad temperature on the inner surface is 0.8 F lower (1,065.8 F. Case EB, 1,0f,6.6 F.
Case EA).
HPCl termination and core quenching occur 4 seconds earlier due to t.
'he slightly faster vessel depressurization.
I Figures 4.2-1 through 4.2-12 present Vermont Yankee NSSS results for the system response. Figures 4.2-13 through 4.2-20 present Vermont Yankee NSSS results for the core response. Figures 4.2-21 through 4.2-26 present Vermont Yankee hot channel results that show this model replicates the high Since the power bundle results obtained from the Vermont Yankee NSSS model.
results in these figures are very similar to those for Case EA, only differences will be described. Finally, Figure 4.2-27 presents a comparison of the maximum clad outer surface temperature histories for Cases EA und EB.
I A comparison of the break flow rates in Figures 4.1-5 and 4.1-6 to those in Figures 4.2-5 and 4.2-6, and a review of the output data for each case shows the following:
(a) The downstream break flow rates are virtually identical. After 206 seconds, the occasional spikes occur at different times, but the mean flowrates are nearly the same.
(b) The upstream break flow rates are virtually identical during the I
first 0.6 seconds.
(c) The break flowrates for Case EB become approximately 25% larger than for Case EA between 0.6 and 17 seconds (recirculation pump trip time for Case EA).
During this period, Recirculation Pump A began to overspeed for Case EB whereas the declining electrical power caused a rundown for Case EA.
The overspeed causes less hydraulic resistance and flashing through the pump than the rundown does. Thus, the quality in the discharge pipe at the choking I
locations (venturi and upstream pipe break) was lower for Case EB, resulting in higher break flow rates. I
I (d) After 17.0 eccends, Racirculstien Pump A bigins to evarapssd.
After 21 seconds, the upstream break flow rates are essentially the i
same for both cases.
I A comparison of Figures 4.1-7 to 4.2-7 shows that Case EB has a slightly faster depressurization rate than Case EA.
This accounts for several I
of the earlier event times discussed before. At 120 seconds, the two pressure histories merge and remain essentially the same thereafter.
A comparison of Figures 4.1-14 to 4.2-14 shows similar behavior for the high power bundle mass history except during the period from 91 to 132 seconds I
I when more two-phase fluid entered the bundle for Case EB.
This occurs after the peak cladding temperature has been reached in both cases and appears to result from a shifL of that the hottest temperature about 18 inches upward for Case EB.
This apparently allows fluid to enter the bundle more readily through the drilled holes in the fuel support piece at the bottom.
l A comparison of Figures 4.1-15 and 4.1-16 to Figures 4.2-15 and 4.1-16 shows generally similar behavior for the two cases, but the following differences are noted:
a.
Case EA suppresses CKF at all elevations until 13.1 seconds.
Case EB allowed the upper elevations (99 and 117 inches) to reach l
CHF at 2.t seconds and the mid-elevations (63 and 81 inches) to I
reach CHF at 12.3 seconds due to less mass in the high power bundle early in time, b.
The subsequent bundle heatup for Case EA shows the 81-inch elevation has the highest PCT whereas the 63 and 99-inch elevations are slightly lower. For Case EB, the 99-inch elevation has the highest PCT and essentially the same value as the 81-inch value for Case EA.
Now, the 63-inch and 81-inch elevations are cooler than for Case EA.
I lI lI l I
I It appears that the lower temperatures at the mid-core elevations c.
I allow the more pronounced two-phase fluid insurge in Case EB between 91 and 132 seconds as seen in Figures 4.1-16 and 4.2-16.
I The high power bundle cools down more readily in Case EB than in Case EA as seen in Figure 4.2-15.
The major edit data for both cases show that Node 5 (72 to 90 inches above BOHL) is the last node to quench in the high power bundle. Figure 4.2-27 shows a comparison of the peak node (5)
I temperature for Case EA and the peak node (6) temperature for Case EB.
I I
I
'I I
I I
I I
I I
lI I I
TABLE 4.2.1 Sequence of Events for Large Break Case EB Event Titne (seconds) 4.0E-6 1.
Break opens 2.
Loss of auxiliary power and power to the 4.0E-6 recirculation pump motors 3.
High drywell pressure (p >2.5 psig) 0.048 4.
Turbine stop valve starts to close (and 0.45 is completely closed in 0.1 second) 5.
Reactor scrams on high drywell prassure 0.55 6.
Initiate turbine bypass valve opening 0.56 7.
Turbine bypass valve completely open 1.16 8.
Low level signal (127 inches) occurs 3.04 9.
RPS MG set underfrequency (57 Hz) 3.01 I
condition occurs 10.
MSIVs begin to close on RPS 3.52 underfrequency 11.
Control rods are fully inserted 4.22 I
12.
Feedwater flow coasts down to zero 5.00 13.
Lower plenum flashing begins 5.00 14.
Midcore critical heat flux occurs 12.30 15.
Low low level signal (82.5 inches) occurs 12.30 16.
MSIVs are completely closed 13.52 17.
HPCI injection begins 20.40 18.
Recirculation Loop B (intact) discharge 56.50 valve begins to close 19.
LPCS injecticn begins b5.00 20.
Minimum Primary System inventory 70.00 (89,286 lb) occurs 21.
Peak clad temperature occurs 78.00 (inner 1,065.80F; outer 1,062.00F) llW 'I
TABLE 4.2.1 (Continued)
{
Sequence of Events for Lerne Break Case EB Event Time (seconds) 22.
Recirculation Loop B discharge valve 89.50 closed L
23.
HPCI flow terminated on low pressure 91.20 signal F
24.
Core, including high power bundle, is 194.00 well cooled
[
E E
E
[
E E
r D
[ I L
VDtMONT YMPeTE NSSS LICDfSING N00CL CASC CB: LNt C SRCN LOCA APPD0!X K RESULTS I
bo DCG RCCIRO LOOP DISCtfutGC P!PC BREM (2 X 3.61 FT2) 5k i
?
E vi M
ia.
I bb b
I l
I I
0.0 5.0 10.0 1$.0 20.0 25.0 TIMC (SCC)
Figure 4.2-1:
Reactor Power History (LBLOCA-EB)
I o
N e,
.s ?
I I
v fx g
e o
' I 0.0 5.0 10.0 15.0 20.0 25.0 j
TINC (SCC)
Figure 4.2-2:
Net Reactivity (LBLOCA-EB) ~
I
vomawr mort NSSS t.1CD6!M MEE.
CRSC EBi t>WtGC BREPIk 10CR FrPDOIX K RESLLTS DEG RECTRC LACP OISCHutwc PIPE stErlK 12 x 3.64 FT21 c
l Y
l 8
Ro j
,I,--.... '....., -
I r
4..
0 l
K ho I
I N-L i
1 e-eFEE0 HATER e--eNAIN STEAMLINE I
I 99 It
\\
~ a-ii f-I e
i
-g 2
E
?
t l
\\
=
2:-
2:
}
I o1 0.0 5.0 10.0 15.0 20.0 25.0 I
TIME (SEC)
Figure 4.2-3:
Feed and Main Steam Flows (LBLOCA-EB) o N
l o
9 f.
\\
l
~
o l
9d
=-
h d a N
\\f
!e 1I k'
1 S-o N
I 0.0 5.0 10.0 15.0 20.0 M
TIME (SEC)
Figure 4.2-4:
Vessel Water Level (LBLOCA-EB),
i VDtron YfgeGI teSSS LICDISING MODEL
(
cyisE Es, usteE sRum tm frPDetx x REsLLTS DEG RECtRC LOOP DISCWWteE PIPC ptDE 12 X 3.64 N2)
'b o
{
li' fi l
h b
g t
5
~
- \\
~
h
- i e-eUPSTREAM BREAK j {
e..e DWNSTREft1 BREAK
[
. i
- l 5"'
\\
s
\\
b M =-
i
.o c **
[
8
\\'......,
-N a
iri-
~~ ' "--........,,,
{
o
' + ---...................................
o o-0.0 I.0
- t. 0
- 8. 0 8.0 10.0
{
TIME (SEC)
Figure 4.2-5:
Early Break Flowrates (LBLOCA-EB) o I
U*
[
N g-h t
J l
t 1
3 t
l 8_..,kB al s
- ~~'
i :.
i:! ::::
f
[
w s.==._ey_ m_ __m. 9 a
U.
_ L u li n g
,,..n,,-
M i
[
E I
nil i
M" I
L.!
EN i
lli u'
m-eUPSTREAM BREAK l;
t
[
l i
t ho o--eDWNSTREAM BREAK j
i.
[
li0.0 65.0 120.0 l$.0 2W.0 M.0
[
TIME (SEC)
Figure 4.2-6:
Long Term Break Flowrates (LBLOCA-EB) -
i VDtPONT YfMGI NSSS LICDtS!W ftIII, UtSC EB LNtec BRUM L:XM FPPEN0!X K RESLLTS I
DEG RECIRC LOOP DISOflRK PIPC BRDE 12 X 3.64 FT2) o w
I o
l_
L Eb9
"-l en i
o I
88 s
ia
\\
\\
I i
od 0.0 80.0 120.0 180.0 240.0 300.0 TIME (SEC)
Figure 4.2-7:
Vessel Pressure History (LBLOCA-EB) o N
I I
e--oHPCI TURBINS 7 -- - - -- : -
~
o--e HPCI PUMP y#TF~-
- -* CORE SPRRY (2) y e
o t_
g.
f I
~
bo 8_
pa
- p. h-A'5BO'5:';4-T
~
M3 E
e
- r e
I o
l 3
o
.}
i i
I k
,.h __
d eseo.:
0.0 60.0 120.0 180.0 240.0 300.0 TIME (SEC)
Figure 4.2-8:
ECCS Flowrates (LBLOCA-EB) -
I N YM N$$$ LIENSING N(CEL N G' LNZ BREAK LOCR RPPDO!X K RESULT 3 I
OEG REC!RC LOOP DIStNetGC PIPC B6 (2 X 3.61 FT2)
Td I
--^
f P
h I
to E
I
?
e I
k!
T e
I g
o I
0.0 e0,0 12h.0 180.0 240.0 300,0 TIME (SEC)
Figure 4.2-9:
Net Flowrate into NSSS (LBLOCA-EB)
NR z5 I
-kI o
84 a
I m
/
3 B:
j od 0.0 ao.O 120.0 too.0 3,o.o soo,o TIME (SEC)
Figure 4.2-10: NSSS Fluid Mass Inventory (LBLOCA-EB) -
VDttOff YfNGI NSSS LI(ENSIIG IGI2.
CASE EBe UUNIE BREN IJ)CR FPPD0!X K RESLLTS I
000 ftEclRC LOOP OISOvetGE PIPE BRERK (2 X 3.64 fT21
'o o 14
/fI
, do a
RP e-eBYPASS REGION i
I e-oUPPER PLENUM l
i 11
=
l V
d 1
?
I
\\
E:i 4l
=
_.L I
l
),h o
i'
)w g
B-a 4,i j.
I d
^
@ o-4
. *Je-e.
5 0.0 60.0 120.0 18'0.0 2t'O.0 300.O TIME (SEC) l Figure 4.2-11: Bypass and Upper Plenum Fluid Mass (LBLOCA-EB) l e:
x-
[
0 - cLOWER PLENUM I
od e--oCRGT REGION O
l h
[
is c.
I y"';
.r !\\
7
t 9 *,
(.
p p' w'.,^'y V
l 3
3 p 'c' o
a 0.0 e0.0 t20.0 n a'0.0 2e0.0 300.0 TIME (SEC)
Figure 4.2-12: CRGT and Lower Plenum Fluid Mass (LBLOCA-EB)
vDWONTYrwEENSSSt.f(INSIE~000, M
cast Ege LARGE GRE N LOCR RPPD OjX K RCSULTS OCS RECIRC Loor O!SC mRGC PIPC BREN (2 1 3.64 PT2)
I D0 E **
@I ?
I I
o l
j g, EU e--cLOW POW REGION e -eCEN RVG REGION I
u>
$$J L
4.-
l f!j
- l.
- g' h,
l Ij l
y f
I 9.
' ',. 4,
\\
g,8 ji[
d f
- l h,s Y
y u
,,::;.p.' : ::=:
=
l o.o e.o is..
ie.o 2...
=. o TIMC (SEC)
Figure 4.2-13: Outer and Central Core Fluid Mass (LBLOCA-EB)
I e
i l
l E"
m h
I e
5 h-q g
[i kjU I
W f
j l
o.o e.o tao.o te.o ae.o n.o TIMC (SEC) i Figure 4.2-14: High Power Assecibly Fluid Mass (LBLOCA-EB).
I
B E!mgNT Ytws_mj;s upsis na0q.
cex cs, ve0t arou um em>otx x a:sults I
gc,s f!g:.1!tg,yjg_e_g12cmist riet emov '2 x 3.64 rT2
}, -
I b
M l
~
I
~
os %
/'
S m-e 63 INCH DE.V I
8 e-o81 INCH ELEV "i ' i 7
.':%O a-* 93 INCH ELCV
~
a
(!
\\
+---+ 117 INCH ELEV' v
w a
g85 i
I r
r f!
a n
y n;q s$4,M 1
I
=a 0.0 80.0 1.10.o 18'0.0 210.0
%3.0 TIFE (SEC)
Figure 4.2-15: High Power Bundle Clad Temperatures (LBLOCA-ES) a.
+
- q: =.
I e-c63 INCH ELEV M
U
'+'w e-o 81 INCH ELEV m - a99 INCH ELEY 5
J
)
+--+ 117 INCH ELEV r y I
2
'i.. i 8si g3 L
M I
f j.
I I
J I
g.
h.i-bukh I
og.. -
0.o e,0.0 320.o too.o 2U.0 300.0 TIME (SEC1 I
Figure 4.2-16: High Power Bundle Qualities (LBLOCA-EB) I
VCR'ONT TRNK!I H$$$ LICCNSinc N000.
CRSC CB tJWtGC BRERK tip APPDCIX K RCSLA.TS I
QCG REC!*C LOOP DISONGC PIPC BRDF (2 X 3.64 fT2) d C
c.e i I l
E Q-e-e 63 INCH ELEY s$
e-o81 INCH ELEV R
- -* 99 INCH CLEV g
+--+ 117 INCH CLEY s
t 8 $.
k.
i I
t 1
9 I
e n
I k
I I
h j
l s-.
o g
l t
. A
_ = = r e r r ;: r r- -___ JS,
)- !IOhb O
l'
^ - - - -
0.0 EC.0 120.0 180.0 240.0 30C.0 TIME (SEC)
Figure 4.2-17: Long Term Heat Transfer Coefficients (LBLOCA-EB)
I d'
C l
a 8 g
l l g
h, g e--cG3 INCH ELEV l
9o e--o 81 INCH ELEV l
W
- - *G9 INCH ELEV
~
+--+ 117 INCH ELEV l
U l
l'a
~
E d-I l
LI 8
3-I oo sd.0 so.O e6.0 ti0.0 is0.0 TIME (SEC)
Figure 4.2-18: Degraded Heat Transfer Coefficients (LBLOCA-EB) I
VDIMDdT YFMTE NSSS LICDfSING MODEL CMC E8s LARGE BRDIK LDCH FFPDOII K RESULTS DES RECIRC LOOP O!SOWIRGE P!PE BREM (2 X 3.64 FT2)
I N
I b
C 8[
rf e
0
.O h
/
g 81 Inch Elev j
m--cLOW POWER BUNDLE C9 j
b{~
k e-o AVG POWER BUNDLE I
h
/y g
- - 4HI POWER BUNDLE
)
s 8
< b,k.
I E9 T
_ n '*.
hE)
)
F*
g"
't 5
c
[
SAT o
V
. f..
r q_
g k @ g gf Abill
, i ni,1 '
N I
a i
i g
2i0.0 3*.0 E
o.o ch.0 120.0 180.0 TIME (SEC)
Figure 4.2-19:
Maximum Bundle Clad Temperatures (LBLOCA-EB)
I 1
I a
81 Inch Elev o--cLOW POWER BUNDLE
- --o AVG POWER BUNDLE g
- -*HI POWER BUNDLE g
s, I
E y.
.).
E i
I 4
ff f
.N.
I I
I
}
I i
^
d h-I
-^____:: r r r = r r r r-5--
cr 0.0 60.0 120.0 180.0 240.0 300.4 TIME (SEC)
Figure 4.2-20:
Bundle Heat Transfer Coefficients (LBLOCA-EB)
I I
__.__..___.____..._._ ___ _ i
m snatcc Mc t.tcp51'88 M0001-p oner ce, vesc encs LDen etx x WM OcG RcCIRC & DI W 'I" a
N c
L C
\\
-e63 INCH ELEY F
t e-o81 INCH ELEV L
..f'.
/
A d.
7 w
'4
.-a99 INCH ELEV t
E m.
/
j %'*y
---+ 117 INCH ELEY p
y e.
F e
I V I L
J o.
I, if I
k
'i o
i._
4 IL l
c e
o l
N 00
,o*o taa.o 18 0 TIME (SEC1 Figure 4.2-21: VY-HC Avg Rod Clad Temperatures (LBLOCA-EB)
O N
r
(
h 9
b
-e 63 INCH ELEV
.t e--o 81 INCH ELEV
' g f*~
[
j ' ANN N7 NCH V
e 1
/
I 4
'e l L
o ing it
[
T f
SAT o
i t
P V
6he E
o E
N o.o 120.0 l# 0 TIME (SEC1
[
Figure 4.2-22: VY-HC Hot Rod Clad Temperatures (LBLOCA-EB) E
VCRNONT YNerE HC LIEN $ LNG MOOCL casC Csi vesC ORCm ux:n nPPCN0ix K RCSLLTS DCG RECIRC LOOP 015CHRRGC PIPC OREM (2 X 3.64 -T21 0
C l l rL b"
01' N
a-e63 INCH ELEV
~
IL
{o4 o-o 81 'NCH ELEV
- - *99 INCH ELEY i
+--+ 117 INCH ELEY m
E F
i og-IL' o
N r
_b
. l l
{ (
lj i
Ib l
$o
= = = = = = = = +
ny L!
=a "-
=
0.0 80.0 120.0 180.0 240.0 300.0 TIME (SEC1 u
Figure 4.2-23: VY-HC Avg. Rod Heat Transfer Coefs (LBLOCA-EB) b 0
C l
L gm e-e63 INCH ELEV i3o4-e-o81 INCH ELEV
~
S
- - *99 INCH ELEY
+-- + 117 INCH ELEV L
E
(
E s
0-A l
l
\\
d
,'l.
l o
0.0 80.0 120.0 180.0 240.0 300.0 TIME (SEC)
Figure 4.2-24: VY-HC Hot Rod Heat Transfer Coefs (LBLOCA-EB)
-100-jI
I VERf10NT TANKCC $ LICD4 SING MOOCL CRSC 08: t#GC BREAK LOCA frPCEfx K RESULT 5 I
DEG RECIRC LOOP O!SCHMGC PIPC 8RCN (2 x 3.64 T21 2
d i
g l
8 O
{8 I
g' n
I e-e63 INCH ELEV EE!
e-o81 INCH ELEV ena
- - *99 INCH ELEV L
+--* 117 INCH ELEV d
I S. -....
g t
b; J,
US s.
- d ed.0 120.0 ts0.c 0.o sc.0 so.0 TIME (SEC)
Figure 4.2-25: VY-HC Degraded Heat Transfer Coefs (LBLOCA-EB) 0 k
4
f N
k
- g. E..."
- 1.
e-c63 INCH ELEV o-o 81 INCH ELEV I
0{
l
- -a99 INCH ELEV I
f" hee
+. 4117 INCH ELEV u l ri d'
's Q
g'.
t i
I s
"d-e J v
I g
II.
f
. M I
y n
Y 0.0 00.0 120.0 180.0 240.0 300.0 TIME (SEC)
Figure 4.2-26: VY-HC High Power Bundle Qualities (LBLOCA-EB)
-101-I
I I
VCRMONT YANKCC NSSS LICCNSING NODCL CASC CB LARGC BREAK LOCA APPENDIX K RCSULTS I
DCG RCCIRC LOOP O!SCHARGE PIPC BREAK (2 X 3.64 FT2) a i
I
~
C EA EB 7
,A g l
J
//
l j
l s_
/
L B
/
EB p
e.
@i I
I a
l8 y
'(
EA i
I.
(ri SAT l
's J ' M r i
i
' F t1 '
o.o so.o 3a0.0 iso.o ae.o 300.o TIMC (SCC)
Figure 4.2-27: Peak Clad Temperatures for Cases EA & EB I
I I
l I I
-102-I
5.0 SAMPLE PROBLEM 3:
SMALL NECIRCULATION LOOP BREAK This section presents licensing analysis results for a small (approximately three-inch diameter) break in the discharge pipe of one I
recirculation loop. The method used complies with 10CFR50.46 and Appendix K L
thereto. The accident assumptions are summarized in Table 5.0.1.
As in the previous sample problems, the break occurs at 4.0E-6 seconds with a coincident loss of normal auxiliary power. HPCI is postulated to fall in compliance with the single failure criterion. Furthermore, RCIC is also postulated to fall.
Therefore, the ECCSs available to mitigate the accident are:
(a) two LPCS Systems, and (b) two LPCI Systems.
IL Table 5.0.2 summarizes the timing of significant events for this accident case. Figures 5.0-1 through 5.0-14 present the Vermont Yankee NSSS results for the system response. Figures 5.0-15 through 5.0-22 present the Vermont Yankee NSSS results for the reactor core response. Figures 5.0-23 r
l through 5.0-27 present the Vermont Yankee HC (hot channel) results. The Vermont Yankee HC results are presented in order to demonstrate the implementation of the methodology described in Section 3.0.
Figures 5.0-1 and 5.0-2 show the reactor power and net reactivity.
In the first second, the power drops sharply due to about -33 cents reactivity caused by voidance. Then, voidance is temporarily arrested causing the power to stabilize and briefly increase from 0.5 to 3.0 seconds. Finally, RPS 7
underfrequency generates a scram signal at 3.0 seconds. At 3.56 seconds, control rod insertien is initiated. This quickly overpowers the short lived reactivity increase. At 7.23 seconds, the rods are fully inserted, and the s
E chain reaction is ended. Beyond this time, the reactor power follows the Appendix K decay heat values.
F Figure 5.0-3 shows the early feedwater and main steam line flow rates.
l The loss of auxiliary power to the condensate and feedwater pumps causes the feedwater flow to coast down in 5.0 seconds. The steam line flow drops in the e
first three seconds due to the decrease in power and feedwater flow. At 3.56 seconds, the Main Steam Isolation Valves (MSIVs) begin to close and are fully closed 10 seconds later. The main steam line flow ramps to zero accordingly.
r
-103-L
_ _ - ~ _ _ _ _ _ _ _ -. - _ - - _.
The turbine stop and bypass valve signals at 18.9 cnd 19.05 ccernda, 1
respectively, are inconsequential. This is because they occur after the steam flow has been cut off by the MSIVs.
Figure 5.0-4 shows the narrow-range water level measured with respect to the top of the enriched fuel. The level generally declines and reaches the I
low level setpoint at 6.4 seconds.
The low low level signal (15.5 seconds) combines with the high drywell pressure signal (18.4 seconds) to initiate the countdown for the actuation of the Automatic Depressurization System (ADS).
In the meantime, following MSIV closure and prior to PDS actuation, Safety Relief Valve 1 (SRV1) is cycled 6 times to maintain the pressure between its lower (1047 psid) and upper (1080 psid) setpoints. The valve cycling and the corresponding pressure behavior are shown in Figures 5.0-6 and 5.0-5, respectively. Finally, I
following the normal delay (120.4 seconds), the ADS operates as designed.
This causes SRVs 1 through 4 to open completely at 138.4 seconds. Their combined discharge is also shown in Figure 5.0-6.
The attendant sharp depressurization is clearly visible in Figure 5.0-5.
I Figure 5.0-7 shows the break flow. prior to ADS actuation at 138.4 seconds, the flow is usually choked and fairly constant at around 500 lb/sec. During this period, the break is discharging liquid water, as evidenced in the break void fraction plot of Figure 5.0-8.
Following ADS I
actuation up to 230 seconds, the break flow declines sharply due to voidance (Figure 5.0-8).
Between 170 and 230 seconds, the break flow is oscillatory.
Given the void pattern of Figure 5.0-8, the mean values for the break flow between 200 and 230 seconds are felt to be too high. This behavior is currently believed attributable to the explicit implementation of Moody's break flow model. This belief is based on the observation of successive reductions in mean break flow over this period (200 to 230 seconds), as successively smaller time steps were r;ed in the calculations. While the issue deserves attention from a best estimate standpoint, it is not felt to be I
significant from the evaluation model perspective. This is because a break spectrum is required for licensing and because there is evidence to support the conclusion that the higher break flow gives conservative results.
I
-104-I
Figures 5.0-9 end 5.0-10 show tha timing and injection flow rates of the two combined LPCS Systems and both LPCI Systems. The brief shutoff
{
periods have been previously explained in connection with the large break (Section 4.1).
F L
Figure 5.0-11 shows the net flow into the system, which remains negative until a few seconds after the injection of LPCI is initiated.
Figure 5.0-12 presents the NSSS mass inventory. The key points are:
(a) the rapid mass depletion due to the ADS actuation at 138.4 seconds, and (b) the minimum inventory and turnaround due to ECC injection at the time of
[
zero net inficw (Figure 5.0-11), and finally, (c) the recovery of inventory to u
the pre-ADS actuation level at 300 seconds.
r Figures 5.0-13 and 5.0-14 show regional mass histories. The bypass region mass decreases slowly prior to 100 seconds and then begins to deplete lL more rapidly. This region begins to refill simultaneously with the initiation of LPCI injection. The upper plenum mass increases in the first 30 seconds.
F L
This r*Sult8 from tha cara void collapse due to the reactor scram. The hydrostatic head imbalance and inertia drive fluid from the downcomer through I
the core into the upper plenum. The MSIV closure then leads to an inventory L
rise. Eventually, the upper plenum mass reaches a maximum (30 seconds) and p
begins to decrease following the system trend. At 140 seconds, the initiation 1
of the ADS causes a brief surge in upper plenum inventory due to flashing inside the core shroud. The brief surgo is followed by a sharp decline. This l
region's inventory remains very low until it recovers a few seconds after LPCS L
injection begins. The lower plenum and the control rod guide tube region mass histories are shown in Figure 5.0-14.
Both regional inventories are nearly insensitive to the break flow itself. A significant depletion is seen only
[
during ADS actuation. Finally, the lower plenum responds to ECC injection with a small delay and begins to refill at 240 seconds.
F Figures 5.0-15 and 5.0-16 present local inventories within the reactor core itself. In all three regions, viz., central average, low power, and high g
b power, there is an early inventory surge which is due to void collapsing and inflow from the downcomer, as previously described in connection with the upper plenum. In all three regions, the inventory remains high until the ADS
[
-105-
[
is actuated. This causes rapid depletion in all three regions. Eventually, the low power region begins to refill sooner than the central average and high
[-
By 280 seconds, all regions have regained their initial power regions.
invaltories.
Figures 5.0-17 and 5.0-18 present the outer clad surface temperatures
{
and the qualities in the high power assembly at four elevations. In spite of the failure of HPCI, the outer clad remains well cooled prior to ADS actuation. Following this event, the quality in the hot assembly rises sharply, causing all elevations to experience critical Heat Flux (CHF) at 157.0 seconds. At about 170 seconds, the rods are cooled by the fluid surge c
L and lower v.ss.1 voiding caus.d my Aos actuation (Figur. 5.0-1s).
rhis cooldown, however, is not sustained and a new heatup begins. The maximum temperatures occur shortly after the Ecc injection begins. Shortly after that, the temperatures drop sharply to the saturation temperature also shown F~
in Figure 5.0-17.
L The heat transfer coefficients for the four elevations are shown in 7
Figure 5.0-19.
The values are seen to drop sharply at 157 seconds which coincides with the onset of CHF.
The cooldown and final quench are evidenced L
dy two periods of high values. The intervening period is characterized by iow values (approximately 20 Btu /hr ft F) similar to those calculated for the large breaks.
p Figures 5.0-20, 5.0-21, and 5.0-22 compare the outer clad surface temperatures, qualities, and heat transfer coefficients for the three regions, viz., low power peripheral, central average, and high power for the 81-inch g
L elevation. rhe dehavior in the other two regions is consistent witt the previous description for the high power assembly.
Finally, the hot channel results are presented in Figures 5.0-23 to 5.0-27.
These results replicate the corresponding Verinont Yankee NSSS results as discussed in section 3.2.
E r
-106-L
~
TABLE 5.0.1 Summary of the Vermont Yankee Small Break Accident Assumptions 2
1.
Small recirculation discharge break (0.05 ft ) at 4.0E-6 seconds.
2.
Loss of auxiliary power occurs at 4.0E-6 seconds.
3.
Reactor scrams after 0.5-second delay from first RPS signal. Scram curve 67B-EOC is used.
4.
Feedwater coasts down to 0.0 lbm/see at 5.0 seconds.
5.
MSIVs close in 10.0 seconds after isolation signal plus 0.5-second delay.
6.
Recirculation pumps in A and B loops coast down with decreasing power from loss of MG sets.
7.
ADS may actuate if appropriate signals exist. Thereafter, ADS cycles open/close at 12 psid between steam line and drywell any time ADS criteria are currently met.
8.
HPCI steam turbine admission valve fails to open on demand. Thus, HPCI fails to inject.
(This is the single failure.)
9.
No credit for RCIC operation.
10.
Two LPCS Systems inject on demand.
I 11.
LPCI-A injection valve opens upon demand.
12.
LPCI-B injection valve opens upon demand.
13.
Drywell pressure and quality are assumed constant at 16.4 psia and 1.0 F for fluid sink conditions. High drywell pressure is 0
I conservatively estimated to occur at 18.4 seconds for this case by a containment calculation.
14.
Wetwell pressure and temperature are assumed constant at 14.7 psia and I
1650F for fluid source and sink conditions.
15.
EM point reactor kinetics initially at 1,664 MWth.
16.
EM core heat transfer.
17.
Passive heat structures are included.
18.
Moody two-phase critical flow model used at the break location.
19.
1971 ANS Decay Heat Standard plus 20%.
-107-
TABLE 5.0.2 Sequence of Events for Small Break Case EY I
Event Time (sec) 4.0E-6 1.
Break opens.
4.0E-6 2.
Loss of auxiliary power.
I 3.
Control rod insertion initiated 0.5 second beyond estimated 3.56 RPS underfrequency reactor trip signal.
3.56 4.
MSIVs begin to close.
5.
Feedwater flow coasts down to zero.
5.0 13.56 6.
MSIVs completely closed.
7.
Low-low level signal.
15.6 8.
Recirculation pump motors trip on low frequency at their 17.0 MG sets.
18.45 9.
High drywell pressure.
18.9 10.
Turbine stop valve begins to close.
11.
Turbine bypass valve begins to open.
19.05 19.65 12.
Turbine bypass valve completely open.
138.4 13.
ADS valves open.
14 Earliest nodal CHF.
157.0 206.07 15.
LPCS injection begins.
16.
Recirculation loop discharge valves begin to close.
206.07 17.
Itinimum primary system inventory (99,483 lb) occurs.
214.40 18.
Peak clad temperature occurs (inner 742.3 F; 221.90 0
outer 740.50F).
I 230.47 19.
LPCI begins to inject.
20.
Average core and high power regions are well cooled.
251.0 I
-108-l
VERMONT YANKCC MS$$ LICDISING t00CL GASC_C7: SMFtt BRCAK LDCA wrClotX K RC$tA.TS RCCIRC LOOP OISCHPetGC PIPC BRCRK 10.05 FT2)
I YN o,
I G
!od-g 1
L I
Es e
o I
i i
5'. 0 1$.0 15.0 25.0 MA I
0.0 TIME (SEC)
Figure 5.0-1: Reactor Power History (SBLOCA-EY)
I 9o '-
9 U
bon-d' U9 s
E@
I 1,
/
=
g 9'
{
o.
15.0 20.0 as.o s'.0 10.0 0.o TIME (SEC)
Figure 5.0-2:
Net Reactivity (SBLOCA-EY)
I
-109-I
vov o n Y W CC NSSS LIC O SING N000.
CASC CT: $ MALT. ORCAK L(XA VPDOlX K RC$ ULT 5 9
RCCIRC LOOP DISCHRRGE PIPC ORCN (0.05 N 21 e
L 8
U9 bN
[
~l 0
\\ -..'
~9
.),..,~,
I
'N L-
's e-e FEEDWATER
\\
e--enRIN STEfV1LINE M9
's,
- f. k' N,
u l
\\
g
's,',
9 m
2
'\\,
'g ed y_
~
c2 3
42 o r
h a
' O.0
$.0 lb.0 l$.0 20.0 25.0 TIME (SEC)
L Figure 5.0-3:
Feed and Main Steam Flows (SBLOCA-EY) o S
~
I-i 3
8-E' O
g9
/\\
m L
=3 v
d E
89 J R-a N
E e
awi u
w x
g-E e
I a
$.0 10.0 15.0 20.0 25.0 0.0 TINE (SEC) p L
Figure 5.0-4:
Vessel Water Level (SBLOCA-EY)
-110-r L
s
E VCRMONT Tf9FCr NS$3 LICENSING NOCO.
CPSC CY $MLL BREAK LOCA frPDoix 0 RCSLLT1 i
RECiftC LOOP DISCHARGE PIPC BFCAK 10.0$ FT2)
I o
.b_
E.39 bl I
Ii!se LJ g.
\\
I o
,e I
N m
o N
od i
0.o 6b.0 120.0 18'0.0 2e.0 300.0 TIME (SEC)
Figure 5.0-5: Vessel Pressure History (SBLOCA-EY) o I
-go m g--
2-S_
I mW e - cSRV1 FLOW c o e--o flDS FLOW l
b
'b o
5' N' E
- g**,
\\
~
5 m
+
9 1.
2
'- e6.o 120.0 180.c 2e.o 300.0 Oo TIME (SEC)
Figure 5.0-6:
S/RV and ADS Flowrates (SBLOCA-EY)
-111-
~
M RNONT Y44rfC NS$$ LICCNSING N00CL l
CASC EY SFWLL BREN LOCA WPCN0!X K RESILTS RCClRC LOOP otSCHmGC P!PC BREN (0.05 FT21 I
o iI D
l g!
l 1
I ei i
D f,
i I
O W,
b o.C Sb.0 12o.0 ab.0 260 20 TIME (SEC)
Figure 5.0-7:
Break Flowrate (SBLOCA-EY) i I
[
T.4 I
g z
.I s.
w e-
+
i ca 2
g m
l 5
L::::::
3 ::: :
II S'
as.o s00.o o.o so.o tac.o too.o NME (SEC1 Figure 5.0-8:
Break Junctions Void Fraction (SBLOCA-EA)
-112-
R VERMONT TfectCE NSSS LICOeSING MO E L l
CASE E7 SMftL BRE N LOCA fFPC @ lx K RCStLTS l
RcciRc toor 01scHest rirt ORom to.os rT2 l
)I a
1,!
E l
l
/
l 8=
eg 2
f I
--J g.
I 9
I g
tio is.s----
ano son.o
'o o -
- e.o TInc (SEC)
Figure 5.0-9: LPCS Flowrate (SBLOCA-EY)
I 9
l n
I h
I IM.
9 l
!: KIN o
'i 1
g i
I
!l l
8 e--e LPCI A 8
o--c LPCI B l l:
9
(
f
~
l l l
l ni l
o I
l i
ll 8'
9o ::: :
=
- _ _180.0 240.o 3o0.0 o.O 80.0 120.o TINE (SEC)
Figure 5.0-10: LPCI A and B Flowrates (SBLOCA-EY)
-113-l
.I i
E tTRMONT TfWEE NSSS LICCNSING N00CL CAE EY: SFWILL 8RCRK LOCR frPC@tX K RCSLA.?$
RCCIRC LOOP DISCNFGC PIPE BRCN (0.05 F721 15 z$
=-
{
l
- y y w
y I
se I
H 1:
I
/
s, 7
g 0.0 d.0 120.0 teIO.0 2e.0 s00.0 TIME (SEC)
Figure 5.0-11: Net Flowrate into NSSS (SBLOCA-EY)
I "S ;
I*
I O,
=
0 l
0
~
m Ee i,
v I
I 0.0 85.0 120.0 180.0 240.0 s00.0 TIME (SEC)
Figure 5.0-12: NSSS Fluid Mass Inventory (SBLOCA-1Y)
~~~
I
I VERMON? YANK [I NS$$ LICCNSING N000L f
CASC CYr SMRLL BRCAK LOCA ffPCNDIX K RCSULTS RCc!RC LOOP DISC MRGC PIPC BREAK (0.05 m l
- oo I
yi Ii e-eBYPBSS REGION o
y e - oUPPER PLENUM s.
4 i
1 I
.I ja /s'\\
O l
I
\\
9 M
m I
I
^\\
\\
I k so i
I i,
a l
f
\\
p m
(\\
i l
o
\\
LJ
(.g.
I o
0.0 60.0 120.0 180.0 210.0 300.0 TIME (SEC)
Figure 5.0-13: Bypass and Upper Plenum Fluid Mass (SBLOCA-EY)
I e:
I **
4 I W
I o
h i
E I
I 3o d'
\\
s 5
/h p
o I
=
E.%,/^ %**, ^ ^'g- #'/
0-
--- e-eLOWER PLENUM I
e--c CRGT REGION i
od 0.0 8b.0 12.0 teIO.C 240.0 300.0 TIME (SEC1 I
Figure 5.0-14:
CRGT and Lower Plenum Fluid Mass (SBLOCA-EY) a l
-115-1 L
igMONT Tf90GT NS$$ LICCNSING M00CL CRei CYr SMALL BRCRK LDCA fFPCMitz K RCStLT5 ReclRC LOOP Ol!OWWtGC PIPC BRCfW: 10.05 FT2) h9 xR k
9 t
E 9
~
[
s.
E Ya c
Se p,,
l a:
a;-.,>-~,y'..,
B
,*,e
,, n,s-a:,
/
'-k
--W 9.
g a--c LOW POW REGION l
1 e'I e-e CEN fWG REGION p.g /
l W
l
= :
,/
o 0.0 65.0 120.0 15.0 2N.0 300.0 L
TIME (SEC1 Figure 5.0-15: Outer and Central Core Fluid Mass (SBLOCA-EY)
I 9
s
[
29 m
l r
E Y
h N
[
E h
I I
5y s-
__1
[
r i
9 0.0 80.0 120.0 100.0 240.0 300.0 '
TIME tSEC)
Figure 5.0-16: High Power Assembly Fluid Mass (SBLOCA-EY)
-116-
VERMONT Y%ef[r NSSS LICOfSING McKL CASC CT: SMALL BREN LOCR &PCM)lx K RESULTS RCCIRC LOOP OISOtmGC PIPC 8REN to.05 W21 o
k c-L
=8
~
a e-e63 INCH ELEV e - oB1 INCH ELEV
- - *99 INCH ELEV
{
+--+ 117 INCH ELEV w
l12 e
I k 'i l
+
mo 6
., (*"} %. *..f g
I w
f
/
3 Q
SAT u
E o EN I
o.o si.o 120.o too.c ae.o 300.0 TIME fSEC1
{
Figure 5.0-17: High Power Bundle Clad Temperatures (SBLOCA-EY)
L r
i y
u
' f.
E
?
d 0-
- ;1 -gy, 8
e-e63 INCH ELEY l
I I
L e-o 81 INCH ELEV o
m 99 INCH ELEY i
~
4'
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Figure 5.0-18: High Power Bundle Qualities (SBLOCA-EY)
-117-w
VCRMONT TftdKCC NSSS LICDfS1 4 N000.
CASE CYr $NALt. BREAK LDCA APPC@!X K RC$lLT5 RECIRC LOOP DISCHARGE PIPC 8REN 10.05 FT2) o C
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k 9
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o 0.0 05.0 1$0.0 100.0 240.0 300.0 TINC CSEC1 Figure 5.0-19: Long Term Heat Transfer Coefficients (SBLOCA-EY)
F 9
b r
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cm o 8 g.
)
81 Inch Elev
~
i e--eLOW POWER BUNDLE L
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x
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TIME (SEC)
Figure 5.0-20: Maximum Bundle Clad Temperatures (SBLOCA-EY)
-118-
I tTRMONT VfwCC NSSS LICENSl$ 6 l
EASE CY. $NRLL BRCfW LOLS frPEN0lX K CESLLTS REclRC LOOP DIS M GC PIPC m to 05 M 21 I
0 l
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a--cLOW POWER BUNDLE o-oRVG POWER BUNDLE k$4
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Bundle Static Qualities (SBLOCA-EY)
-119-
I vtRMONT YfteIC W t.fCDtSIse PC0il CR$C eye $MLL BREN LDCR fMCNDIX E RC91TI RECTRC LOOP OlSCMRGE l'IPC OREN 10.05 FT2)
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e-o81 INCH ELEV F
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Figure 5.0-24:
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-120-I
E VERMONT Y4ECC HC 1.lCDf5f@ N00CL CRSC CY8 SMLL BRCRK LOCA &PDolX K RCSLA.TS RECIRC WOP OISCHNtGC PIPC BREAK (0.05 FT21 4
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ll h.
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u ba#1' e--e 63 INCH ELEY e-o 81 INCH ELEV s
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Figure 5.0-26: VY HC Hot Rod Heat Transfer Coefs (SBLOCA-EY)
-121-I
ttmon swrCC H: ucCwstNo M000.
CRSC CYe $MR1 BRCfW LOCA WN0!X K RCSLVS RECIRC LO(P DISCWGC PIPC p (0.8 N21 na L
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r e
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-122-E E
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I
6.0 CONCLUSION
S The Vermont Yankee BWR loss-of-coolant accident licensing analysis method has been described. This method is based upon the RELAPSYA and FROSSTEY computer codes.
Each code has been extensively assessed against experimental data, analytical solutions, and sensitivity studies summarized in Tables 1.3.2 and 1.3.3 and Reference A-2.
This assessment has shown that these codes will yield conservative and reliable predictions for both large I
and small break LOCAs when applied in a conservative manner.
The method uses two Vermont Yankee base input models: Vermont Yankee NSSS and Vermont Yankee HC.
Conservative initial fuel rod conditions for these models are derived from FROSSTEY calculations. Each base input model uses licensing assumpt. ions and computational options within the RELAPSYA code that comply with 10CFR50.46 and Appendix K requirements. Additional conservative assumptions, including certain items listed in Tables 4.0.1 and 5.0.1, are incorporated in the input data for the two base models.
I Three LOCA sample problems have been presented to demonstrate the application of this method for the Vermont Yankee Nuclear Power Station.
These include two large break cases and one small break case. For each case, the calculated results show the following:
The calculated peak clad temperature for the two large break and I
a.
the small break cases are below the 2,200 F peak clad temperature limit specified in 10CFR50.46(b)(1).
I b.
The total cladding oxidation at the peak location is below the 17%
limit specified in 10CFR50.46(b)(2).
The hydrogen generated in the core by cladding oxidation during c.
these accidents is less than the 1% limit specified in 10CFR50.46(b)(3).
I I
I
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E d.
The calculations chow thst the ccra estainsd a ecolable gsometry for each case analyzed. The fuel rod models for the three core regions show minor dimensional changes but no clad ruptures are calculated. Thus, the coolable geometry criterion in 10CFR50.46(b)(4) is satisfied for these three cases.
I Each calculation shows that the available ECCS successfully e.
initiated and the core was well cooled in less than 300 seconds.
Therefore, the long-term core cooling criterion in 10CFR50.46(b)(5) is satisfied for these cases.
I Together, the RELAPSYA computer code, FROSSTRY computer code and the two Vermont Yankee base input models will be used to obtain LOCA ECCS results for Vermont Yankee that. comply with 10CER50.46 criteria and Appendix K requirements.
I I
I I
I I
I I
I
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4se -
I
7.0 REFERENCES
I 1-1.
Final Safety Analysis Report: Vermont Yankee Nuclear Power Station.
Vermont Yankee Nuclear Potter Corporation, Brattleboro, Vermont, as amended.
1-2.
Fernandez, R.
T., R. K. Sundaram, J. Ghaus, A. Husain, J. N. Loomis, L. Schor, R. C. Harvey, and R. Habert, RELAP5YA - A Computer Program I
for Light-Water Reactor System Thermal-Hydraulic Analysis. Volume I:
Code Description, Yankee Atomic Electric Company Report YAEC-1300P (October 1982).
(Proprietary) 1-3.
Fernandez, R. T., R. K. Sundaram, J. Chaus, A. Husain, J. N. Loomis.
L. Schor, R. C. Harvey, and R. Habert, RELAPSYA - A Computer Program for Light-Water Reactor System Thermal-Hydraulic Analysis. Volume II:
I Users Manual, Yankee Atomic Electric Company Report YAEC-1300P (October 1982).
(Proprietary) 1-4.
Fernandez, R. T., R. K. Sundaram, J. Ghaus, A. Husain, J. N. Loomis.
L. Schor, R. C. Harvey, and R. Habert, RELAPSYA - A Computer Program for Light-Water Reactor System Thermal-Hydraulic Analysis. Volume III:
Code Assessment, Yankee Atomic Electric Company Report YAEC-1300P I
(October 1982).
(Proprietary) 1-5.
YAEC Letter to USNRC, " Response to the NRC Questions on RELAPSYA."
I FYR 85-22, FVY 85-18 (March 1, 1985).
1-6.
YAEC Letter to USNRC, " Response to the NRC Questions on RELAPSYA,"
FYR 85-48 (April 30, 1985).
1-7.
YAEC Letter to USNRC, " Response to the NRC Questions on RELAPSYA,"
FYR 85-72 (July 1, 1985).
1-8.
YAEC Letter to USNRC, " Response to the NRC Questions on RELAPSYA,"
FYR 85-87 (August 15, 1985).
1-9.
YAEC Letter to USNRC, "P.esponse to NRC Questions on the RELAP5YA Computer Program." FYR 85-121 (November 1, 1985).
1-10 YAEC Letter to USNRC, " Response to Additional NRC Questions on the RELAP5YA Computer Code," FVY 85-122 FYR 85-139 (December 31, 1985).
2-1 VerPlanck, D. M., Methods for the Analysis of Bolling Water Reactors l
Steady-State Core Physics, YAEC-1238, March 1981.
2-2 VerPlanck, D. M., SIMULATE-2:
A Nodal Core Analysis Program for f
Light-Water Reactors, YAEC-1392P, June 1984.
1 2-3 Letter, D. B. Vassallo, NRC, to J. B. Sinclair, " Methods Reports j
I Acceptance," NVY 82-157, September 15, 1982.
,3 3-1 Steves, L. H., et al., HUXY: A Generalized Multi-Rod Heatup Code With g
10CFR50. Appendix K Heatup Option User's Manual, Exxon Nuclear Company, Inc., IN-CC-33(A), devision 1, November 14, 1975.
-125-lI
4-1.
Meyars, L.
L., BWR R* fill-Reflood ProRreta: Final R port, EPRI NP-3093, L
GEAP-30157 NUREG/CR-3223, General Electric Company, San Jose, CA (April 1984).
4-2.
Lee, L.
S., G. L. Sozzi, S. A. Allison, BWR Large Break Simulation Tests: BWR Blowdown / Emergency Core Cooling Program, Volume 2 EPRI NP-1783 GEAP-24962-2, NUREG/CR-2229, General Electric Company, I
San Jose, CA (July 1982).
L 4-3.
Lee, L.
S., G. L. Sozzi, S. A. Allison, BWR Large Break Simulation Tests: BWR Blowdown / Emergency Core Cooling Program, Volume I,
{
EPRI NP-1783, GEAP-24962-1, NUREG/CR-2229. General Electric Company, u
San Jose, CA (April 1982).
I Eu a
m lI l
lI lI I
lI 2.-
I
I APPENDIX A: FUEL ROD INITIAL CONDITIONS FROM FROSSTEY A.1 FROSSTEY DESCRIPTION RELAP5YA requires the user to supply information concerning the initial 1
conditions of the fuel contained within the NSSS model. The information is
)
extracted from the FROSSTEY (Fuel Rod Steady-State Thermal Effects) code (References A-1 and A-2) developed by Yankee Atomic Electric Company. The FEOSSTEY code has undergone extensive benchmarking against both commercial and I
test reactor fuel rods irradiated under a wide variety of conditions.
FROSSTEY has proved capable of predicting the response of these rods in a manner consistent with its intended use (i.e., commercial rods operating at low to high burnups at moderate power levels). THe FROSSTEY code has been reviewed by the NRC's core performance branch and a SER on the code's use in non-LOCA applications at low to moderate exposures has been issued. YAEC is submitting an updated version of the code with the intention of gaining I
approval for use at high fuel burnup and for LOCA applications.
I The FROSSTEY code is specifically designed to provide fuel rod temperature distributions, fuel-to-clad gap conductances, fuel rod dimensional changes, fission gas release and gap inventories, internal pressure, and stored energy predictions as a function of the fuel rod operating history.
The RELAPSYA code requires a number, but not all, of these parameters to be specified as input.
I The general component models which comprise the FROSSTEY code are listed in Table A.1.
The code models the fuel rod as a number of axial I
segments and a plenum region which are in communication with each other. The power history of the fuel rod is represented by a series of constant power steps. A file management structure within the code allows the user to store
(
appropriate variables to allow a problem restart at a particular time step in l
subsequent computer runs.
I l
-127-l l
I A.2 SYSTEM CALCULATION DATA I
The Vermont Yankee NSSS core was divided into three radial zones for analysis purposes in the FROSSTEY fuel performance calculations. These radial zones correspond to the three regions defined in the RELAp5YA NSSS model. The low power peripheral region contains 116 fuel bundles with the highest exposures, but relatively low power. The average power central region contains 248 fuel bundles with lower exposures but with a higher power. The high power central region represents the four highest power bundles within the core.
The response of the regions is defined to be the response of a single fuel rod operating with the region's average parameters. All fuel was modeled as p8X8R bundles containing 62 fuel rods at 2.89 w/o U-235 within a UO 2 matrix. The active fuel length was represented as 25 axial slices each 6 inches in length. The FROSSTEY model employed the recommended set of options.
The power history of the three regions is based on Cycle 10 which is fairly representative of Vermont Yankee's present operation. The power level chosen for the central average region is representative of the power level at I
the exposure intervals (end-of-cycle) for those bundles. The high povar central region power is conservatively set at a peaking factor of 1.50.
The 1.50 peaking factor bounds all assembly radial peaking factors seen to date at Vermont Yankee. The average power central region has a peaking factor of 1.186.
The peripheral region's peaking factor is 0.5851 and is defined as the remainder of the core total power. The exposure of each region is defined by the upper and lower bound of the batch average exposures within the regions.
The fuel rod response, predicted by the FROSSTEY code, was based on a I
long-term exposure with a Haling axial power shape, and is restarted for a short-term exposure step with a highly peaked chopped cosine axial power shape. During the long-term exposure, the reactor was assumed to be operating at 1593 MWth (100% N3R), and during the short-term was assumed to be at 1664 MWth (104.5% NBR). The switch to the higher power is designed to encompass the Appendix K power uncertainty (1.02%) without incurring the permanent effects on certain phyr.ical processes which are history-dependent.
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[
A.3 MOT. CHANNEL CALCULATION DATA s
The hot channel calculations are performed in a manner that provides assurance that the bundle power history bounds all fuel bundles in the core.
The response is predicted in a manner similar to the method used to define the
[
NSSS region's responses. The response is based on a long-term exposure employing bounding power and a Haling axial shape followed by a short-term exposure at a target power level employing a highly peaked shape. The target r
l power level is defined to be high enough to bound all allowable peak location powers.
L.
The switch to the short-term power level is accomplished by raising the r
power to correspond with operation at 1664 MWth and a change to a higher
!w peaked axial shape. The switch occurs at. the time corresponding to the f
exposurs point where the LOCA limit is to be evaluated. The hot rod and the L
average bundle peak power location are both placed on the target power level.
c w
[
[
Eu L
F L
-129-IL
A.4 REFERENCES A-1 Schultz, S. P. and K. E. St. John, Method for the Analysis of Oxide Fuel Rod Steady-State Thermal Effects (FROSSTEY) Code /Model Description Manual, YAEC-1249P, April 1981.
A-2 Schultz, S. P. and K. E. St. John, Method for the Analysis of Oxide Fuel Rod Steady-State Thermal Effects (FROSSTEY) Code Qualification and Application YAEC-1265P, June 1981.
A-3 Letter, D. B. Vassallo (USNRC) to R. W. Capstick (VYNPC), " Approval of Use of Fuel Performance Code FROSSTEY," September 27, 1985.
l i
I I
)
-130-I
TABLE A.1 FROSSTEY Component Models Closed Coolant Channel Heat Balance coolant-to-Cladding Hest Transfer Clad Crudding Clad Corrosion Clad Creepdown Clad Elastic Deflection l
Clad Axial Growth Themal Dimensional Changes Thermal Flux Depression Fuel Densification I
Fuel Relocation / Cracking Fuel Fission Product Swelling (History-Dependent)
Fission Gas Production and Release (History-Dependent)
Indigenous (Sorbed) Gas Release and Reaction Pellet Dish Effects Grain Crowth and Themal Restructuring Fuel Melting Dimensional Changes Fuel Cladding Gap Conductance (Component Analysis)
Rod Internal Pressure Fuel Heat Capacity and Stored Energy i
. I
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