ML20095G822
| ML20095G822 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 06/30/1984 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML112230638 | List: |
| References | |
| NEDO-21082-03, NEDO-21082-03-APP-A, NEDO-21082-3, NEDO-21082-3-APP-A, NUDOCS 8408280209 | |
| Download: ML20095G822 (46) | |
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CLASSI JUNE 1984 APPENDIX A l
LOSS-OF-COOLANT ACCIDENT i
ANALYSIS REPORT FOR l
DUANE ARNOLD ENERGY CENTER-(LEAD PLANT)
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l O! S GENER AL h ELECTRIC L...
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NEDO-21082-03 Class I June 1984 Appendix A LOSS-OF-COOLANT ACCIDENT ANALYSIS REPORT FOR DUANE ARNOLD ENERGY CENTER (LEAD PLANT) o f
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'NEDO-21082-03 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The only undertakings of General Electric Ccmpany respecting information in this doctor:ent are contained in the contract between Iotxt Electric Light and
. Power Company and General Electric Capany and nothing contained in this document shall be construed as changing the contract.
The use of this infor-mation by anyone other than Iowa Electric Light and Pouer Ccmpany or for any purpose other than that for which it is intended, is not authorized; and trith
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respect to any unauthorized use, General Electric Cmpany makes no represen-
- tation or carranty, and aceumee no liability as to the completenees, accuracy,
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or usefulness of the information contained in this document.
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NEDO-21082-03 Page A.1 INTRODUCTION A-1 A.2 LEAD PLANT SELECTION A-2 A.3 ^
INPUT TO ANALYSIS A-4
'A.4
'LOCA' ANALYSIS COMPUTER CODES A-4 A.4.1 Results of the LAMB Analysis A-4 A.4.2 Results of the SCAT Analysis A-4 A.4.3 Results of the SAFE Analysis A-5 A.4.4 Results of REFLOOD Analysis A-5 A.4.5 Result of the CHASTE Analysis A-6 A.4.6 Methods A-7 A.5 BREAK SPECTRUM CALCULATIONS-A-7 A.5.1 Large Break Analysis A-8 A.5.2 Small Break Analysis A-10 A.5.3 Intermediate Break Analysis A - A.6 CONCLu 'ONS A-12 REFERENCES A-13 iii
c NEDO-21082-03 Title Page Table-31 BWR/4 With Loop Selection Logic Important LOCA/ECCS A-14 Parameters 2
Significant Input Parameters to the Loss-of-Coolant
-A-15 Accident Analysis A-16 3-Summary of Break Spectrum Results A-17 4
LOCA Analysis Figure Summary - Lead Plant Sa MAPLHCR Versus Average Planar Exposure (P8DRB301L/
A-18 BP8DRB301L)
Sb MAPLHGR Versus Average Planar Exposure (P8DPB289/
A-18 BP8DRB289)
Sc MAPLHGR Versus Average Planar Exposure (P8DRB284H/
A-19 BP8DRB284H) 5d MAPLHCR Versus Average Planar Exposure (P8DRB299/
A-19 BP8DRB299)
A-20 6
Single Failure Evaluation J
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NEDO-21082-03 LIST OF ILLUSTRATIONS Figure Title Page la Water Level Inside the Shroud and Reactor Vessel Pressur,e Following a Design Basis Accident, LPCI Injection Valve A-21 Failure lb Water Level Inside the Shroud and Reactor Vessel Pressure Following a Break of the Recirculagion Line, LPCI Injection Valve Failure, Break Area = 1.0 ft (LBM)
A-22 1c Water Level Inside the Shroud and Reactor Vessel Pressure Following a Small Break of the Recirculation 2 (ne, LPCI Li Injection Valve Failure, Break Area = 1.0 f t SBM)
A-23 1d Water Level Inside the Shroud and Reactor Vessel Pressure Following a Small Break of the Recirculation Line, LPCI Failure, Break Area = 0.R ft2 (SBM)
A-24
.le Water Level Inside the Shroud and Reactor Vessel Pressure Following a Small Break of the Recirculation Line, HPCI Failure, Break Area = 0.07 f t2 (SBM)
A-25 2a Peak Cladding Temperature Following a Design Basis Accident, A-26 LPCI Injection Valve Failure 2b Peak Cladding Temperature Following a 1 f t Break, LPCI A-27 Injection Valve Failure (LBM) 2c Peak Cladding Temperature Following a Small Break of the Recirculation Line, LPCI Injection Valve Failure, Break Area = 1.0 f t2 (SBM)
A-28 2d Peak Cladding Temperature Following a Small Break of the Recirculation Line, LPCI Injection Valve Failure, Break Area = 0.8 ft2 (SBM)
A-29 2e Peak Cladding Temperature Following a Small Break of the 2
. Recirculation Line, HPCI Failure, Break Area = 0.07 ft A-30 (SBM) 3a Fuel Rod Convective Heat Transfer Coef ficient During Blowdown A-31 at the High Power Axial Node for DBA Fuel Rod Convective Heat Transfer Coef ficjent During Blowdown 3b at the High Power Axial Node for a 1.0 ft Break (LBM)
A-32 4a Normalized Core Average Inlet Flow Following a Design Basis A-33 Accident 4b Normalized Core Average Inlet Flow Following a 1.0 f t A-34 Break (LBM) 5a Minimum Critical Power Ratio Following a DBA A-35 Sb Minimum Critical Power Ratio Following a 1 f t Break (LBM)
A-36 6
Normalized Power Versus Time A-37 v
NEDO-21082-03 LIST OF ILLUSTRATIONS (Continued)
Figure Title Pg 7
Peak Cladding Temperature Versus Break Area A-38 Variation With Break Area of Time for Which Hot Node
.8 Remains Uncovered A-39 t
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NED0-21082-03 A.1 INTRODUCTION The purpose of this document is to provide the results of the loss-of-coolant accident (LOCA) analysis for the Duane Arnold Energy Center (DAEC). The analysis was performed using approved General Electric (GE) calculational models.
This analysis of the plant LOCA is provided in accordance with the NRC requirement (Reference 1) and to demonstrate conformance with the ECCS accep-tance criteria of 10CFR50.46. The objective of the LOCA analysis contained herein is to provide assurance that the most limiting break size, break location, and single failure combination has been considered for the plant.
The required documentation for demonstrating that these objectives have been satisfied are given in Reference 2.
The documentation contained in this re-port is intended to satisfy these requirements.
The general description of the LOCA evaluation models is contained in Reference 3, with a detailed description provided in Reference 4.
Plants are separated into groups for the purpose of LOCA analysis (Reference 5).
Within each plant group there will be a single lead plant analysis which pro-vides the basis for the selection of the most limiting break size yielding the highest peak cladding temperature (PCT). Also, the lead plant analysis provides an expanded documentation base to provide added insight into evalua-tion of the details of particular phenomena. The remainder of the plants in that group will have non-lead plant analyses referenced to the lead plant analysis. This document contains the lead plant analysis for DAEC which is a BWR 4 with loop selection logic (plants that have not incorporated the low pressure coolant injection (LPC1) system modification) group plant and is con-sistent with the requirements outlined in Reference 2.
The same models and computer codes are used to evaluate all plants. Changes to these models will cause changes in phenomenological responses thac are sim-ilar within any given plant group. Dif ferences in input parameters are not expected to result in significantly dif ferent results for the plants within a given group. Emergency core cooling system (ECCS) and geometric differences A-1
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between plant groups may result in different responses for different groups but within.any group the responses will be similar. Thus, the lead plant concept
.is still valid for.this evaluation.
A.2 LEAD PLANT SELECTION Lead plants are selected and analyzed in detail to permit a comprehensive re-view and eliminate unnecessary calculations. This constitutes a generic analy-sis for each plant of that type which can be referenced in subsequent plant sub-mittals.
Based on the criteria given in Reference 5, all operating General Electric Boiling Water Reactors (BWR) have been divided into groups. A lead plant was selected for each group whose LOCA response would be representative of the entire group. The three main criteria for selecting the lead plants are:
.(1) typical blowdown and reflood characteristics; (2) 1 typical reactor thermal power; and (3) number of plants.
The first of these is important because it establishes that the shape of the break spectrum will be. typical of plants in the class. The second and third are important because they establish the degree to which the lead plant analysis can
.be' considered " generic".
The duration of nucleate boiling is the best measure of blowdown heat transfer.
It is determined.largely by_ the ratio of downcomer volume to break area.
Shorter periods of nucleate boiling during blowdown result in higher PCT's since less i
stored energy is removed.
Reflood time is the best measure of overall emergency core cooling effective-It is determined primarily'by the complement of ECCS equipment available, ness.
.and the number of bundles with alternate flow path holes.
Shorter reflood times result in lower PCT's since the fuel has less time to heat up after nucleate boiling is lost.
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Based on these consideraticus, the first criteria is the primary reason for
' differentiating between the various groups of plants. The balance between heat transfer during blowdown and reflood time is of primary importance in deter-mining how plants respond to the LOCA.
The reactor thermal power is used to determine the lead plant from a large number of plants in a group to permit a more comfortable extrapolation of the generic results to other reactor sizes where no plant is clearly identified by the first criterion.
The number of similar plants is used as a criterion for selecting lead plants since it maximizes the number of plants that the lead plant analysis is directly applicable to.
As a result of the above criteria, all currently operating jet-pump GE-BWR's have been separated. into three groups for the IDCA analysis.
The three groups 4
are identified as BWR/3, BWR/4 with loop selection logic, and INR/4 with LPCI modification. The basis for selecting DAEC as the lead plant for the BWR/4 with loop selection logic group is discussed in the following paragraph.
I Since the last lead plant analysis some of the plants which were previously in this group have installed the LPCI modification. Only two domestic operat-ing BWR/4 plants have not installed the LPCI modification. The previous lead plant for this group was FitaPatrick which has since installed the LPCI mod-ification. Of the two remaining plants in this group, there is no technical reason to select one above the other. Table 1 shows a comparison of the in-portant LOCA/ECCS parameters for both plants remaining in this group. Based I
on Criterion 1 there is no significant difference between these two plants.
Since there are only two plants remaining in this group Criteria 2 and 3 are not applicable. Thus, DAEC has been selected as lead plant for this group since
-it was scheduled for completion of the ECCS analysis first.
A third plant, Pilgrim, has be.9n added to this group. Although Pilgrim is classified as a BWR/3, its design incorporates finger springs on the fuel and alternate flow path holes in the fuel bundle lower tieplates, and it incor-porates the same ECCS design as the BWR/4 plants with loop selection logic.
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NED0-21082-03 This difference between Pilgrim and the other BWR/3's results in Pilgrim hav-ing a significantly reduced reflood time. Thus, Pilgrim is closer in ECCS 1
performance to a BWR/4 than a BWR/3 and has been included with the BWR/4's.
A.3 INPUT TO ANALYSIS A list of the significant plant input parameters to the LOCA analysis is pre-sented in Table 2.
A.4 LOCA ANALYSIS COMPUTER CODES ~
A.4.1 Results of the LANB Analysis This code is used to analyze the short-term blowdown phenomena for large pos-tulated pipe breaks (breaks in which nucleate boiling is lost before the water
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level drops and uncovers the active fuel). in jet pump reactors.
The LAMB out-(core flow as a function of time) is input to the SCAT code calculation of put blowdown heat trar.sfer.
4 The LAMB results presented are:
Core Average Inlet Flow Rate (normalized to unity at the begin-l e
ning of the accident) following a Large Break.
A.4.2 Results of the SCAT Analysis l
This code completes the transient short-term thermal-hydraulic calculation l
for large breaks in jet pump reactors. The GEXL correlation is used to track the boiling transition in time and location. The post-critical heat flux heat transfer correlations are built into SCAT which calculates heat transfer coefficients for input to the core heatup code, CHASTE.
1 The SCAT results presented are:
e Minimum Critical Power Ratio following a Large Break.
Convective Heat Transfer Coefficient following a Large Break.
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e NEDO-21082-03 A.4.3 Results of the SAFE Analysis This code is used primarily to track the vessel inventory and to model
-ECCS performance during the LOCA.. The application of SAFE is identical for all break sizes.
The code is used during the entire course of the postulated accident, but af ter ECCS initiation, SAFE is used only to calculate reactor system pressure and ECCS flows, which are pressure dependent.
The SAFE results presented'are:
Water Level inside the Shroud (up-to the time REFLOOD initiates) e and Reactor Vessel Pressure A.4.4 Results of REFLOOD Analysis This code is used across-the break spectrum to calculate the system inventories
.after ECCS actuation. The models used for the design basis accident (DBA)
. application ("DBA-REFLOOD") was described in a supplement to the SAFE code description transmitted to the USNRC December 20, 1974. The "non-DBA REFLOOD" analysis is nearly identical to the DBA version and employs the same major assumptions. The only differences stem from the fact that the core may be partially. covered with coolant at the time of ECCS initiation and coolant
-levels change slowly for smaller breaks by comparison with the DBA. More precise modeling of coolant level behavior is thus required principally to determine the' contribution of vaporization in the fuel assemblies to the counter current flow limiting (CCFL) phenomenon at the upper tieplate. The differences from the DBA-REFLOOD analysis are:
(1) The non-DBA version calculates core water level more precisely I
than the DBA version in which great precision is not necessary.
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(2) The non-DBA version includes a heatup model similar to but less detailed than that in CHASTE, designed to calculate cladding tem-i
.perature during the small break. This heatup model is used in calculating vaporization for the CC"L correlation, in calculating swollen level in the core, and in calculating the peak cladding temperature.
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The REFLOOD results presented are:
- Water Level inside the Shroud Peak Cladding Temperature and Heat Transfer Coefficient for e
breaks calculated with small break models A.4.5 Result of the CHASTE Analysis This code is used, with suitable inputs from the other codes, to calculate the fuel cladding heatup rate, peak cladding temperature, peak local cladding oxidation, and core-wide metal-water reaction for large breaks. The detailed fuel model in CHASTE considers transient gap conductance, clad swelling and rupture, and metal-water reaction.
The empirical core spray heat transfer and channel wetting correlations are built into CHASTE, which solves the transient heat transfer equations for the entire LOCA transient at a single axial plane in a single fuel assembly.
Iterative applications of CHASTE determine the maximum permissible planar power where required to satisfy the
- requirements of 10CFR50.46 acceptance criteria.
The CHASTE results presented are:
Peak Cladding Temperature versus time e
Peak Cladding Temperature versus Break Area o
Peak Cladding Temperature and Peak Local Oxidation versus Planar e
i Average Exposure for the most limiting break size Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) versus Planar Average Exposure for the most limiting break size A summary of the analytical results is given in Table 3.
Table 4 lists the figures provided for this analysis. The MAPLHGR values for each fuel type presently in the DAEC core are presented in Tables Sa through 5d.
A-6
NED0-21082-03 A.4.6 Methods In the following sections, it will be useful to refer to the methods used to c.nalyze DBA, large breaks, and small breaks. For jet-pump reactors, these are defined as follows.
- . DBA Methods. LAMB / SCAT / SAFE /DBA-REFLOOD/ CHASTE.
Break size: DBA.
b.
Large Break Methods (LBM). LAMB / SCAT / SAFE / non-DBA REFLOOD/ CHASTE, Break sizes:
1.0 ft.
< A < DBA.
c.
Small Break Methods (SBM). SAFE /non-DBA REFLOOD, Heat transfer coefficients: nucleate boiling prior to 2
core uncovery, 25 Btu /hr-ft _oF after recovery, core spray when appropriate. Peak cladding temperature and peak local oxidation are calculated in non-DBA-REFLOOD.
2 Break sizes: A < 1.0 ft,
A.5 BREAK SPECTRUM CALCULATIONS For convenience in describing the LOCA phenomena, the break spectrum has been separated into three regions:
small breaks, intermediate breaks and large breaks. The selection of the break sizes to be included in each region is dependent on the most limiting single failure and the ECCS evaluation method used. The potentially limiting single active failures considered in establishing the various break regions are given in Table 6.
The small break region is defined as that portion of the break spectrum where the high pressure coolant injection (HPCI) is the most limiting single failure. In this region, the "small break methods" are used.
A-7
NED0-21082-03 The intermediate break region is defined as that portion of the break spectrum up to the transition break where the LPCI injection valve is the most limiting single failure. The transition break is defined as the 2
1.0 ft break size.
This break size has been chosen in order to be consistent with previous analyses thus allowing for a better comparison between the old and new analyses.
The calculational techniques employed in the SBM are intended to conservatively model small breaks only. As the break size increases (I 1.0 f t ) the SBM becomes overly conservative and does not appropriately de-scribe some of the phenomena (e.g., radiation heat transfer, blowdown heat trans-fer).
The transition break has been analyzed with both the large and small break methods with the same single failure to allow a comparison between the methods.
The analysis of the transition break is shown in Figures 1b through Sb for the large break method and Figures 1c and 2c for the small break methods.
In the intermediate break region, small break methods are used.
The large break region is defined as that portion of the break spectrum between the transition break and the DBA. The DBA is defined as the complete severance of the largest pipe in that portion of the system which yields the highest peak cladding temperature when the most limiting single failure is assumed. The most limiting single failure in this region is the failure of the LPCI injection valve.
In the large break region, large break methods are used. For the DBA, the DBA methods are used.
A.S.1 Large Break Analysis In this region, the vessel depressurizes rapidly and the HPCI has an insignificant effect on the event.
Consequently, failure of the core spray or LPCI is more severe. Analyses have demonstrated that failure of the LPCI is the most severe failure among the low pressure ECCS because unlike the core spray which must pass through the CCFL regions at the top of the core, LPCI is injected into the lower plenum through the jet pumps. Thus, the LPCI injection valve is the worst single failure in the large break region.
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NED0-21082-03 The characteristics that determine which is the most limiting break are:
(a) the calculated hot node reflooding time, (b) the calculated hot node uncovery time, and (c) the time of calculated boiling transition.
The time of calculated boiling transition increases with decreasing break size, since jet pump suction uncovery (which leads to boiling transition) is determined primarily by the break size for a particular plant. The calculated hot node uncovery time also generally increases with decreasing break size, as it is primarily determined by the inventory loss during the blowdown. The hot node reflooding time is determined by a number of inter-acting phenomena such as depressurization rate, counter current flow limiting and a combination of available ECCS.
The period between hot node uncovery and reflooding is the period when the hot node has the. lowest heat transfer. Hence, the break that results in the longest period during which the hot node remains uncovered results in the highest calculated PCT. If two breaks have similar times during which the hot node remains uncovered, then the larger of the two breaks will be -
limiting as it would have an earlier boiling transition time (i.e., the larger break would have a more severe LAMB / SCAT blowdown heat transfer analysis).
Figure 8 shows the variation with break' size of the calculated time the hot node remains uncovered for DAEC. Based on these calculations, the DBA was determined to be the break that results in the highest calculated PCT in 2
the 1.0 ft to DBA region. Confirmation that this is the most limiting break over the entire break spectrum is-shown in Figure 7.
The DBA results are presented in Figures la through Sa, 6, and 8.
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A.S.2 Small Break Analysis 1
In this region. the vessel depressurizes ' relatively slowly (or not at all, j
depending on.the break size) because the. break is small. HPCI is the most severe. equipment failure in this region because its loss results in a loss of ECCS delivery' capability at high pressure. With HPCI available, the core remains covered for longer periods of time.than for cases with a single ADS valve failed or for cases with low-pressure core spray or LPCI failures.
For the BWR/4 plants with. loop selection logic, the remaining ECCS assuming.
- an HPCI failure are 2 low pressure core spray and 4 LPCI pumps. With all of the LPCI flow directly available to the lower plenum, the reflooding 4
5 time is' rather insensitive to the small changes in CCFL (which affect the delivery of core spray.to the lower plenum) that result from the model and input changes. Therefore, the uncovered time for the fuel in this region of the break spectrum is only slightly altered when new inputs and models are applied. The change in PCT from the previous ECCS analysis (Reference 6) is due, primarily, to a.different heatup rate during the period of fuel uncovery. Any change in MAPLHGR calculated from the limiting break is fed back into the small break heatup calculation by a change in the power of the f~
hot rod. The change in PCT due to this power change is approximately pro-
-portional to the change in MAPLHGR.
2 For DAEC, the limiting break size (0.07 ft ) has been identified as a result of the several sizes considered as shown on Figure 7.
The results 2
of the 0.07 ft analysis are shown on Figures le and 2e.
For all BWR/4's with loop selection logic, the small break temperatures are on the order of 1500*F (specific results for DAEC are shown on Figure 7) and since MAPLHGR will seldom increase by more than about 15%, the maximum expected increase in small-break temperature is about 15% of (1500 - 500'F),
or about 150'F.
l-This. trend will be experienced by all plants in this group because:
4 (1) For small breaks with LPCI flow, the core spray flow into the lower plenum through the alternate flow path holes contributes-A-10 4
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NED0-21082-03 a very small amount to the total reflooding flow. Thus, the effect of the alternate flow path is overshadowed by the LPCI flow.
(2) The small break has characteristically a linear heatup. Thus, the effect of a change in reflooding time can be accurately predicted from a previous analysis.
(3) The change in the' slope of the heatup curve is directly pro-portional to the change in power (i.e., MAPLHGR).
A.5.3 Intermediate Break Analysis In this region, the vessel depressurizes rather rapidly through the break and the high-pressure delivery capability of HPCI is less significant than it is for smaller breaks. Consequently, failure of low pressure core spray or LPCI is more severe. Analyses have demonstrated that the LPCI is the
.most severe failure among the low pressure ECCS because, unlike core spray which must pass through the CCFL regions at the top of the core, LPCI is injected into the lower plenum through the jet pumps. Thus, the LPCI injec-tion valve failure, which results in no LPCI being available is the worst single failure in the intermediate break region.
For DAEC, the limiting break size (0.8 f t ) in this region has been identified as a result of the several sizes considered as shown on Figure 7.
The results of the 0.8 ft analysis are shown on Figures ld cod 2d.
Throughout the entire intermediate break spectrum (%0.3 ft to 1.0 ft )
there is a similar reduction in the PCT as evidenced by the 1.0 ft analysis (Figure 7).
This same trend will be experienced by all plants covered by this lead plant analysis.
A-ll
-'A.6 CONCLUSIONS The results of the analysis' demonstrate that the ECCS will perform its func-
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tion in an acceptable manner and that.the ECCS acceptance criteria of 10CFR50146 are met.
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NED0-21082-03 I
REFERENCES i
1.
Letter, George Lear (NRC) to Duane Arnold (IEL&P), "Re:
Duane Arnold Energy Center," dated March 11, 1977.
2.
Letter, Darrell G. Eisenhut (NRC) to E.D. Fuller (GE),." Documentation of the Reanalysis Results for the Loss-of-Coolant Accident (LOCA) of Lead and Non-Lead Plants," June 30, 1977.
3.
General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K, NEDO-20566 (Draft), submitted August 1974, and General Electric Refill Reflood Calculation (Supplement to SAFE Code Description) transmitted to the USAEC by letter G.L. Gyorey (GE) to Victor Stello, Jr. (NRC), dated Decmber 20, 1974.
4.
General Electric Standard Application for Reactor Fuel (U.S. Supplement),
NEDE-240ll-P-A-6-US, April 1983.
5.
Letter, G.L. Gyorey (GE) to V. Stello, Jr., dated May 12, 1975, "Com-pliance with Acceptance Criteria of 10CFR50.46."
6.
Duane Arnold Energy Center Loss-of-Coolant Accident Analyses Conform-ance with 10CFR50 Appendix K (Jet Pump Plant), June 1975.
A-13
NED0-21082-03 Table 1 BWR/4 WITH LOOP SELECTION LOGIC IMPORTANT LOCA/ECCS PARAMETERS
' Parameter' DAEC PilgrLa Browns Ferry 3 4
Power,-HWt 1691 1998 3293 Vessel Id, in.
183'- 224 251 Recirc Line ID, in. 22 I' 28 28 Number of Fuel Bundles 368 580 764 Number of Drilled 368 428 764 Fuel Bundles Fuel Design 8x8 8x8 8x8 (Operating Reactors) ECCS Available, DBA ,2CS 2CS 2CS ECCS Available, 2CS 2CS 2CS Small Break 4LPCI 4LPCI 4LPCI ADS ADS ADS e 4 A-14
NED0-21082-03 Table 2 SIGNIFICANT INPUT PARANETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS PLANT PARAMETERS: Core Thermal Power 1691 MWt, which corresponds to 102% of rated steam flow 6 Vessel Stesa Output 7.344x10 lbm/h, which corresponds to 102% of rated steam flow Vessel Steam Dome Pressure 1055 psia Recirculation Line Break Area (DBA) 2 2 for Large Breaks 2.51 ft, 1.0 ft 2 2 2 Recirculation Line Break Area 1.0 ft, 0.8 ft, 0.07 ft for Small Breaks Number of Drilled Bundles 368 FUEL PARAMETERS: Peak Technical Initial Specification Design Minimum Linear Heat Axial Critical Fuel Bundle' Generation Rate Peaking Power Fuel Type Geometry (kW/ft) Factor Ratio A. P8DPB289/ 8x8 13.4 1.4 1.2 BP8DRB289 B. P8DRB299/ 8x8 13.4 1.4 1.2 BP8DRB299 C. P8DRB284H/ 8x8 13.4 1.4 1.2 BP8DRB284H D. P8DRB301L/ 8x8 13.4 1.4 1.2 BP8DRB301L A-15
NEDD-21082-03 Table 3
SUMMARY
OF BREAK SPECTRUM RESULTS Core-Wide o Break Size Peak Local Metal-Water o Location o Single Failure PCT (*F) Oxidation % Reaction % o 2.51 ft2 (DBA) 1959 1.1 0.08 o Recirc suction LPCI Injection Valve o o 1.0 ft2 (LBM) o Recire Suction 1765 Note 1 Note 2 LPCI Injection Valve o o 1.0 ft2 (SBM) o Recirc suction 1358 Note 1 Note 2 LPCI Injection Valve o 2 o 0.8 ft o Recire Suction 1142 Note 1 Note 2 LPCI Injection Valve o 2 o 0.07 ft Recire Suction 1566 Note 1 Note 2 o o HPCI NOTES: 1. Less than DBA (1,1%) 2. Less than DBA (0.08%) A-16
J Table 4 LOCA ANALYSIS FIGURE
SUMMARY
- LEAD PLANT Small Break Method Large Break Method Limiting Break
- DBA - Transition Break Transition Break Limiting Break (LPCI Inj. Valve (LPCI Inj. Valve (LPCI Inj. Valve (LPCI Inj. Valve Failure Failure) Failure) Failure) (HPCI Failure) 2 (2.51 ft ) (1.0 ft2) (1.0 ft2) (0.8 ft2) (o,07 fc ) Water Level Inside la lb ic Id le I Shroud and Reactor Vessel Pressure 2a 2b 2c 2d 2e Peak Cladding Temperature Heat Transfer 3a 3b 2c 2d 2e [ 4 Coefficient og u Core Average 4a 4b Inlet Flow w Minimum Critical Sa Sb Power Ratio Peak Cladding Tem-2a perature of the Highest Powered Plane Experiencing Boiling Transition Normalized Power 6 Peak Cladding Tem-7 perature vs. Break Area Hot Node Uncovered 8 Time vs. Break Area
NED0-21082-03 Table 5a MAPLHGR'VERSUS AVERAGE PLANAR EXPOSURE PLANT: DAEC BUNDLE TYPE: P8DRB301L/BP8DRB301L Average Planar Exposure MAPLHCR PCT 0xidation (mwd /t) (kW/ft) (*F) Fraction 200 11.5 1947 0.011 1000 11.5 1936 0.011 5000 11.9 1927 0.011 10000 12.3 1935 0.011 15000 12.4 1959 0.011 20000 12.2 1941 0.011 25000 11.3 1854 0.008 35000 9.9 1679 0.004 45000 8.7 1559 0.002 Table 5b MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT: DAEC BUNDLE TYPE: P8DPB289/BP8DRB289 Average Planar Exposure MAPLHCR PCT 0xidation (mwd /t) (kW/ft) (*F) Fraction 200 11.2 1915 0.010 1000 11.2 1906 0.010 5000 11.8 1923 0.010 10000 12.0 1914 0.009 15000 12.1 1926 0.010 20000 11.9 1921 0.010 25000 11.4 1863 0.008 30000 10.8 1786 0.006 35000 10.3 1712 0.005 40000 9.6 1646 0.004 45000 8.9 1580 0.003 A-18
NEBO-21082-03 Table Sc MAPLHGR VERSUS AVERACE PLANAR EXPOSURE PLANT: DAEC BUNDLE TYPE: P8DRB284H/BP8DRB284H Average Planar Exposure MAPLHGR PCT 0xidation (tfWd/ t) (kW/ft) (*F) Fraction 200 11.2 1912 0.010 1000 11.2 1900 0.010 5000 11.7 1909 0.010 10000 12.0 1921 0.010 15000 12.0 1926 0.010 20000 11.8 1918 0.010 25000 11.1 1837 0.007 30000 10.4 1744 0.005 35000 9.8 1674 0.004 40000 9.1 1608 0.003 45000 8.5 1541 0.002 Tabic 5d MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT: DAEC BUNDLE TYPE: P8DRB299/BP8DRB299 Average Planar Exposure MAPLHGR PCT 0xidation (mwd /t) (kW/ft) (*F) Fraction 200 10.9 1874 0.009 1000 11.0 1869 0.009 5000 11.5 1874 0.009 10000 12.2 1929 0.010 15000 12.3 1950 0.011 20000 12.1 1948 0.011 25000 11.5 1890 0.009 30000 11.0 1807 0.007 35000 10.3 1725 0.005 40000 9.7 1661 0.004 45000 9.0 1601 0.003 A-19
NEDD-21082-03 Table 6 SINGLE-FAILURE EVALUATION The following table shows the single, active failures considered in the ECCS performance evaluation. Suction Break Asstaned Failure Systems Remaining LPCI Injection Valve ADS, 2 CS, HPCI Diesel Generator (D/G) ADS, 1 CS, HPCI, 2 LPCI HPCI ADS, 2 CS, 4 LPCI One ADS Valve ADS minus one, 2 CS, HPCI, 4 LPCI Other postulated failures are not specially considered because they all result in at least as much ECCS capacity as one of the above assumed failures. A-20
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i VESSEL PRESSUFE (PSIR) , DAF - 16.74 FEET , TAF - 29.25 FEET WATER LEVEL, FEET = 1.2 60 =10' C ~m 0.8 40 O_ i 4 Q m ,5n /, / ,' W LLI u-s D i J S E O l" y ~ i w E L2J ui h J T j "~ \\V, / U l/ d v t2 3 W O LJJ i "1#- 0.'I' o 0. 100. 200. 300. 400. TIME [ SECONDS) Figure Id. Water Level Inside the Shroud and Reactor Vessel Pressure Following Small Break of the Recirculation Line, LPCI Injection Valve Failure, Break Area = 0.8 ft2 (SBM)
1 1 VESSEL PRESSURE (PSIA) 2 BAF -16.74 FEET 3 TAF -29.25 FEET 4 WATER LEVEL, FEET 60 =10' i E 40 b 0.8 i Q_ z . a....e -
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_a _ = ^ t=s = m u u LL3 g -- b [ b ~ [ s s s i s s _s s d s 3 v ( y (D oo 7 wu W = 8 W ew [ A j 20 tc n_ 0.4 N il = _] LLJ (D CD Lt.] i k i i i ~ l'
=='o* 0 0. 0. 200. 400. 600. 800. TIME (SECONDS) Figure 1e. Water Level Inside the Shroud and Reactor Vessel Pressure Following a Small Break of the Recirculation Line, HPCI Failure, Break Area = 0.07 ft2 (SBM)
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I 1 PEAK CLAD TEMPERATURE 2 HTC BTU /HR-FT2-DEG F 3 PCT LIMIT -- 2200.0 3. =10' s i s s s s s 3 { I tb 2. til z O o ~ Peak Cladding Temperature h O_ 2; t I u 00 y ll l 1 i i-- O i C 1' \\ u _J C 2_op) 2 Below: Heat Transfer Coefficient (Btu /hr-fL r 25 ~ Core spray Heat Transfe 2 2 2 2 2 N
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i 1 PEAK CLAD TEMPERATURE 2 HTC BTU /HR-FT2-DEG F 3 PCT LIMIT-2200.0 3. =10' { 5 5 s s 3 3 3 3 I tb 2. LLI O ~ m Mea k Cladding Tempera ture go Q-h E W o a g o b O C 1. j d x y C ' C L1J 2 1 10,000 Core Spray Below: Heat Tra 1sfer Coefficient I Btu /hr-ft OF) 25 Heat Jransfer 0.l' obi b l 2 nr - o. 200. .m. 800. ooo. TIME [ SECONDS) Figure 2e. Peak Cladding Temperature Following a Small Break of the Recirculation Line, HPCI Failure. Break Area = 0.07 ft2 (SBM)
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