ML20073H264
| ML20073H264 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 03/31/1991 |
| From: | OMAHA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML19298E441 | List: |
| References | |
| OPPD-NA-8303-NP-R03, OPPD-NA-8303-NP-R3, NUDOCS 9105070038 | |
| Download: ML20073H264 (138) | |
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Omaha Public Power District !
, Nuclear Analysis Reload Core Analysis Methodology l
. I i Transient and Accident Methods and Verification !
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l OPPD-NA-8303-NP :(
Rev. 03 March 1991 t
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!9105070038'910419 '
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t Table 2 i TRANSIENT AND ACCIDENT METHODS AND VERlFICATION J
OPPD-NA-8303-NP Rev.03 Eagg Section Change All All Added reference to CENTS. Changed *The D! strict" or j
' District" to *OPPD* l lil, IV All Renumbered pages to indicato corroct page numbers of <
sections.
Y All Renumbered pages to agree with tables in document.
Changed title of Table 5.1-1. ,
vi, vil All Renumbered figures to be consistant with other OPPD methodology topicals. Renumbered pagos to reflect actual figure page number, vill Added Revision 03 to list.
1 1.0 Added information on CENTS as computer code.
2 2.2.1 Revised to indicate present rather than future tense.
5 Changed LOCA analyst from CE to W.
6 3.0 Defined TM/LP as Thermal margin / low pressure trip 7 3.0 Deleted Excess Load event from TM/LP transient term.
Added Excess Load to DNB LCO events.
7 4.0 Added reference to CENTS and Indicated use of CENTS as plant simutation code.
8 -10 4.1 Additional CENTS Description 15 5.1.3E Added "of* between " coefficient " and " reactivity" 17 Table 5.1-1 _ Changed " Initial Conditions" to ' Key Parameters" 20 Reclassiflod boron dilution event.
21 5.2.5 Added multiplier 'm" to account for density changes.
Changed "Teo" to 'Oliution timo constant
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L Table 2 (Conti_nued)-
TRANSIENT AND ACCIDENT METHODS AND VERIFICATION OPPD-NA-8303-NP Rev.03 32 5.6.1 Added information on Excess Load as DNB margin event, Transferred margin requirement from LSSS to LCO.-
33,34 5.6.1 Added information on AT power calculator.
34,35 5.6.3 - Changed objective from bias term to ROPM value. Added -
objective on validating VHPT.
5.6.4 Added information on ROPM and reference to Reference 5-2. Added and changed key parameters to reflect -
calculation of ROPM. !
'35 5.6.5 ' Changed method to calculate ROPM term rather than bias .
term.'
36' Table 5.6.4-1 Changed /added key parameters.
37- 5.6.6 Changed 10 CFR 50.59 criterla to reflect ROPM results from =
bias term.
'43- 5.8.4 L - Changed "197,000" to "196,000" for LCO flow rate to be used.'
50 t ' 5.1'0.2/3 Revised analysis criteria for W methods, 51 . 0.11 Revised LOCA to indicate W methods.
59-72,77-78 L - 6.0 - - Added information on CENTS verification and use. Revised -
figure numbers to be consistent with other OPPD methodologies.'
'70 -- Ref. 2-1 Added reference for W LOCA methods.
Ref.s 4-11,12 and 13 : Added references for CENTS.. '
' 80 : Ref 5-5 Changed reference from CE to W for CEA Ejection methodologies.--
Ref. 5-7 Added reference for W LOCA methods.
Ref. 5-8 Corrected typographical error, t
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Table 2 (Continued)
TRANSIENT AND ACCIDENT METHODS AND VERIFICATION OPPD-NA-8303-NP Rev. 03 80 Refs, 6' 3,4 Added references to CENTS.
81 Figure 5-1 Added figure as referenced in Section 5.6.5 82-125 Figures Added CENTS Information and renumbered figures.
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ABSTRACT This document is a Topical Report describing Omaha Public Power District's reload core transient and accident rnithods for application to Fort Calhoun Station Unit No.1. The report addresses the District's transient and acc! dent analysis methodology and its application to the analysis of reload cores. In addition, comparisons of results using the NSSS timulation code to results from experimental measuromonts and independent calculations are provided.
OPPD-NA-8303-NP, Rev. 03 i
p Proprietary Data Clause This document is the property of Omaha Public Power District (OPPD) and contains proprietary information, Indicated by brackets, developed by Combustion Engineering (CE). The CE Information was purchased by OPPO under a proprietary information agreement.
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OPPD-NA-8303-NP, Rev 03 Il
TAB'.E OF CONTENTS
1.0 INTRODUCTION
AND
SUMMARY
1 2.0 CHAPTER 14 EVENTS CONSIDERED IN THE RELOAD CORE ANALYSES 1 2.1 Criteria 1 2.2 USAR, Chapter ;4, Safety Analysis Evonts Not Considered in Reload Core Analyses 2 2.2.1 MalposlVoning of Group N CEAs (formerly Part-Length CEAs) 2 2.2.2 Idle-Loop Sth1up Event 3 2.2.3 Turbino Generatcc Overspeed Event 3 2.2.4 Loss of Load Event 3 2.2.5 Malfunctions of the Feectwater System 4 2.2.6 Steam Generator Tube Rupturo Accident 5 2.2.7 Loss of Coolant Accident 5 2.2.8 Containment Pressure Analysis 5 2.2.9 Generation of Hydrogen in Containment 5 2.2.10 Fuel Handling Accident 5 2.2.11 Gas Decay Tank Rupture 6 2.2.12 Waste Liquid Event 6 2.3 USAR, Section 14, Events Considered In a RoloadCore Analysis 6 3.0 TRANSIENT AND ACCIDENT ANALYSIS AND TECHNICAL SPECIFICATIONS 6 4.0 TRANSIENT AND ACCIDENT ANALYSIS MODELS 7 4,1 Plant Simulation Model 7 4.2 DNBR Analysis Models 10 4.3 Application of Uncertainties 10 5.0 TRANSIENT AND ACCIDENT METHODS 12 51 CEA Withdrawal Event 13 5.2 Boron Dilution Event 19 5.3 Control Element Assembly Drop Event 22 5.4 Four-Pump Loss of Flow Event 26 5.5 Asymmetric Steam Generator Event 29 5.6 Excess Load Event 32 5.7 RCS Depressurization Event 38 5.8 Main Steam Line Break Accident 40 OPPD-NA-8303-NP, Rev. 03 111
s TABLE OF CONTENTS (Continued) 5.9 Sol 200 Rotor Accident 46 5.10 CEA Ejection Accident 49 5.11 Loss of Coolant Accident 51 5.12 Loss of Load to Both Steam Generators Event 51 5.13 Loss of Feedwater Flow Event 55 6.0 TRANSIENT ANALYSIS CODE VERIFICATION 59 6.1 Introduction 59 6.2 Comparison to Plant Data co 6.2.1 Turbine-Reactor Trip 60 6.2.2 Four-Pump Loss of Coolant Flow 62 6.3 Comparison Between OPPD Analyses and independent Analyses Previously Performed by the Fuel Vendors 64 6.3.1 Dropped CEA 65 0.3.2 Hot Zero Power Main Steamline Break 65 6.3.3 Hot Full Power Main Steamline Break 67 6.3.4 RCS Depressurization 68 6.4 Comparison Between OPPD Analyses and independent Analyses Previously Performed with CESEC-ill for Verification of CENTS 69 6.4.1 Dropped CEA 70 6.4.2 RCS Depressurization 70 6.5 Summary 72
7.0 REFERENCES
79 OPPD-NA-8303-NR Rev. 03 IV
l LIST OF TABLES TABLE . TITLE PAGE 5.0-1 Reactor Protectivo System and Safety injec' ion 12 l
5.1-1 Key Parameters Assumed in CEAW Event Analysis 17 l
5.3.4-1 Key Parameters Assumed in the Full Length CEA Drop Analysis 25 l
l 5.4.4-1 Key Parameters Assumed in the Loss of Coolant Flow Analysis 28 g I
5.5.4 -1 Key Parameters Assumed in the LL/ ISO Event 31 I I
5.6.4 -1 Key Parameters Assumed in the Excess Load Event Analysis 36 I :
5,7,4- 1 Key Parameters Assumed in the RCS Depressurization Event Analysis 39 g 5.9.4-1 Key Parameters Assumed in the Solzed Rotor Analysis 48 5.12.4-1 Key Parameters Assumed in the Loss of Load to Both Steam 53 Generator Analysis 5.13.4-1 Key Parameters Assumed in the Loss of Feedwater Flow Analysis 58 6.3-1 Comparison of Parameters including Uncertaintles Used in 73 the CEA Drop Analysis for Cycles 6 and 8 6.3-2 Comparison of Parameters locluding Uncertainties Used in 74 the HZP Main Steamline Break Analysis for Cycles 6 and 8 6.3-3 Comparison of Parameters locluding Uncertaintics Used in 75 the HFP Main Steamline Break Analysis for Cycles 6 and 8 6.3-4 Comparison of Parameters including Uncertainties Used in 76 the RCS Depressurization Analysis for Cycles 6 and 8 6.4-1 Comparison of Paremeters including Uncertainties Used in 77 I
the CEA Drop Analysis for Cycle 12 6.4-2 Comparison of Parameters including Uncertainties Used in 77 the RCS Depressurization Analysis for Cycle 12 OPPD-NA-8303-NP, Rev. 03 v
LIST OF FIGURES FIGURE TITLE PAGE 5-1 Excess Load Event ROPM Methods 81 1 1
6-1 Full Power Turbine Trip Core Power vs Time 82 6-2 Full Power Turbinu Trip Pressurizer Pressure vs Time 83 6-3 Full Power Turbine Trip Pressurizer Level vs Time 84 6-4 Full Power Turbine Trip RCS Hot Leg Temperatures 85 6-5 Full Power Turbino Trip Steam Generator f1 Pressure 86 6-6 Full Power Turbine Trip Steam Generator #2 Pressure 87 6-7 Full Power Turbine Trip Feedwater Flow vs Time 88 6-8 Full Power Turbine Trip Steam Generator #1 Steam Flow 89 6-9 Full Power Turbine Trip Steam Generator #2 Steam Flow 90 6-10 Four Pump Loss of Flow Total RCS Flow Fraction vs Time 91 6-11 Four Pump Loss of Flow Pressurizer Pressure vs Time 92 6-12 Four Pump Loss of Flow Pressurizer Level vs Time 93 6-13 Four Pump Loss of Flow Core Power vs Time 94 6-14 Four Pump Loss of Flow Steam Generator Pressure vs Time 95 6-15 Four Pump Loss of Flow RCS Temperatures vs Time 96 6-16 Four Pump Loss of Flow Main Feedwater Flow vs Time 97 6-17 Four Pump Loss of Flow Steam Flow vs Time 98 6-18 CEA Drop incident Core Power vs Time 99 6-19 CEA Drop Incident Core Average Heat Flux vs Time 100 6-20 CEA Drop incident Coolant Temperature vs Time 101 6-21 CEA Drop Incident Pressurizer Pressure vs Time 102 6-22 Zero Power MSLB Accident Core Power vs Time 103 6-23 Zero Power MSLB Accident Core Average Heat Flux vs Time 104 6-24 Zero Power MSLB Accident Total Reactivity vs Time 105 OPPD-NA-8303-NP, Rev. 03 vi
LIST OF FIGURES (Continued)
FIGURE TITLE PAGE 6-25 Zero Powei MSLB Accident Coolant System Pressure vs Time 106 6426 Zero Power MSLB Accident Steam Generator Pressure vs Time 107 6-27 Full Powar MSLB Accident Core Power vs Time 108 6-28 Full Power MSLB Accident Core Average Heat Flux vs Time 109 6-29 Full Power MSLB Accident Total Reactivity vs Tirr.e 110 6-30 Full Power MSLB Accident Coolant System Pressure vs Time 111 6-31 Full Power MSLB Accident Reactor Coolant Tempnratures 112 vs Time: Cycle 6 6-32 Full Power MSLB Accident Reactor Coolant Temperatures 113 vs Time: Cycle 8 6-33 Full Power MSLB Accident Steam Generator Pressure vs. Time 114 6-34 RCS Depressurization incident RCS Pressure vs Time 115 6-35 RCS Depressur!zation incident Core Power vs Time 116 6-36 RCS Depressurization incident Core Average Heat Flux vs Time 117 6-37 RCS Depressurization incident RCS Pressure vs Time 118 6-38 Dropped CEA Incident Core Power vs Time 119 6-39 Dropped CEA Incident Core Average Heat Flux vs Time 120 6-40 Dropped CEA incident RCS Temperatures vs Time 121 6-41 Dropped CEA Incident Pressurizer Pressure vs Time 122 6-42 RCS Depressurization Incident Core Power vs Time 123 6-43 RCS Depressurization incident RCS Pressure vs Time 124 6-44 RCS Depressurization incident Preseurizer Pressure vs Time 125 OPPD-NA-8303-NP, Rev. 03 vil
OMAHA PUBLIC POWER DISTRICT RELOAD CORE ANALYSIS METHODOLOGY TRANSIENT AND ACCIDENT METHODS AND VERIFICATION REVISION DATE 00 September 1983 01 November 1986 02 April 1988 03 March 1991 g OPPD-NA-8303-NP, Rev. 03 vill
Omaha Pubile Power District Reload Core Analysis Methodology Transient and Accident Methods and Verification
1.0 INTRODUCTION
AND
SUMMARY
This report discusses the methodology the Omaha Public Power District utilizes to analyze transients and accidents for reload cores. In addition, the report discusses OPPD's verification of the Combustion Engineering System Excursion Code (CESEC) and the Combustion Engineering Nuclear Transient Simulation code (CENTS) for Fort Calhoun Station transients.
The purpose of this verification is to demonstrate OPPD's ability to properly utilize the CESEC and CENTS codes.
OPPD's transient and accident analysis methodology for reload cores is based upon the -
reanalysis of those Updated Safety Analysis Report (USAR), Chapter 14 events whose consequences may be adversely affected by changes in parameters associated with any reload core. The USAR Chapter 14 events which must be considered during a feload core analysis are discussed in Section 2.0. Section 3.0 discusses the transient analyses which determine certain parameters spe0ified in the Technical Specifications. OPPD's transient analysis models are discussed in Section 4.0. OPPD's application of these transient analysis models to the various Chapter 14 events is discussed in Section 5.0. The verification of the NSSS simulator model used by OPPD is discussed In Section 6.0. References are provided in g Section 7.0.
2.0 - CHAPTER 14 EVENTS CONSIDERED IN THE RELOAD CORE ANALYSES This section discusses the criteria utilized to detarmine if a Chapter 14 event need be considered in reload core analyses. Each event which h not formally considered in a reload core analysis is discussed and the reasons given for not normally including the event in the reload core analyses. The methodology applied to taese events will not be discussed in this
- report.
2.1 Criterla
- The criterion used to determine the events considered in reload core analyses is that changes in various neutronics parameters adversely affect the safety analyses of these events. The core parameters considered are the pin peaking factors, Fr and Fxy, the
- Moderator Temperature Coefficient (MTC), the Fuel Temperature Coefficient (FTC) or Doppler Coefficient, the boron concentration, the inverse boron worth, the neutron kinetics parameters, the CEA reactivity worth and the cooldown reactivity associated with a steam line break. If these parameters change such that the previously reported results for a Chapter 14 event are no longer conservative, then this event must be OPPD-NA-8303-NP, Rev. 03 1 of 125
2.0 CHAPTER 14 EVENTS CONSIDERED IN THE RELOAD CORE ANALYSES (Continued) 2.1 Criteria (Continued) reanalyzed if these parameters are conservative with respect to the values assumed in the referenced safety analyses, the criteria of 10 CFR 50.59 are met and this event is not reanalyzed. If a change in some of the parameters may cause the results of a safety analysea to be nonconservative, the event is reanalyzed, if the criteria for the event are still met, then the requirements of 10 CFR 50.59 are satisfied. The event is reported as being reanalyzed and that It has been determined that no unreviewed safety question exists for the event. In some cases it may be possible that an event is reanalyzed and it is determined that an unreviewed safety question exists in these cases the analyses for these events are submitted, in addition, any safety analyses which are performed as a result of a change in the Technical Specifications are reported as part of the supporting documentation for a Facility License Change.
Criteria not directly associated with the reload core but which may be considered in a reload analysis are changes to plant systems which would take place during a refueling and would first be utilized during the operation of the subsequent core, in cases where either physical modifications or modifications in operating procedures are made that do impact the safety analyses, the results of the revised safety analyses are reported in a reload core analysis. This methodology report does not consider tne methodology that is required to analyze all events which could be affected by this criteria, rather, if submittals are made which require analyses of events other than those discussed in this report, revisions to this methodology report will be made to incorporate the methodology used for those events.
2.2 USAR. Chaoter 14 Safety Analysis Events Not Considered in Reload Core Analvses This section discusses the USAR, Section 14, safety analyses which are not normally considered !n a reload core analysis. The USAR section is discussed and the reasons for not including it in the scope of these analyses is discussed. Typically, the reasons for not analyzing these events are that the operating modes considered in the events are no longer allowable at Fort Calhoun Station, the event is not associated with ar;y core parameters of the event is analyzed by a fuel vendor for OPPD. B 2.2.1 Maloositionir$ of Grouc N CEAs (formerly Part-Length CEAs)
This event is not analyzed in the reload core analysis because the Group N part-length CEAs were replaced with full length CEAs in Cycle 11, The use of Group N during power operations is prohibited by the Technical Specifications. The drop of a Group N CEA is considered in the fulllength CEA analysis.
OPPD-NA-8303-NP, Rev,03 2 of 125
2.0 CHAPTER 14 EVENTS CONSIDERED IN THE RELOAD CORE ANALYSES (Continued)
- 2.2 USAR. Chaoter 14 Safety Analysis Events Not Considered in Reload Core Analvsg][(Continued) 2.2.2 Idle-Loco Startuo Event This event is not analyzed because part-loop operation is not permitted by the Fort Calhoun Technical Specifications. i l
2.2.3 Turbine Generator Oversceed Event Tl is event is an analysis of the consequences of a turbine wheel failure and is unrelated to any reload core changes.
I 2.2.4 Loss Of Load Event !
A. The loss of load to both generators is assessed to determine if: i The pressurizer safety valves limit the reactor coolant system pressure to a value below 110% of design pressure (2750 psla) In accordance with Section lit of the ASME Boller and Pressure Vessel Code, and sufficient thermal margin is maintained in the hot '
fuel assembly to assure that Departure from Nucleate Bolling
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(DNB) does not occur throughout the transient. This event is not analyzed with respect to the first criterla since the relief capacity of the pressudzer safety valves does not change and the initial energy contained in the reactor coolant system will not change unless power level is raised above 1500 MW or the reactor coolant' system inlet temperature is significantly increased. Section 14.9 of the USAR reports that the DNBR for the loss of load transient never decreases below the initial value considered in the analysis, e Therefore, it is concluded that any change in a parameter which -
could effect the DNBR foi this event would much more significantly
- effect other events and that_lt is not necessary to analyze this
- event with respect to DNBR criteria.
- Steam generator tube plugging performed during a refueling outage has the potential for altering the heat transfer characteristics assumed in Section 14.9.1 of the USAR. Section
. 5.12 of this document addresses the methodology to be employed should the steam generator tube plugging exceed or be expected to exceed the current USAR analysis assumptions.
OPPD-NA-8303-NP, Rev. 03 3 of 125
2.0 CHAPTER 14 EVENTS CONSIDERED IN THE RELOAD CORE ANALYSES (Continued) 2.2 USAR. Chnoter 14. Safety Analysis Events Not Considered in Reload Core AnnIvses (Continued) 2.2.4 Loss Of Load Event (Continued)
B. The loss of load to one steam generator is discussed in this methodology report as one of the asymmetric steam generator transients.
2.2.5 Malfunctions of the Feedwater System The analyses which are reported in USAR, Section 14.10 Malfunctions of the Feedwater System, are the total loss of feedwater flow and the loss of feedwater heating The results of the total loss of feedwater flow show that the minimum DNBR does not decrease below its initial steady stato value and that no safety limits are approached durin0 the event. Therefore, this event is not reanalyzed in a reload core analysis.
The loss of feedwater heating is the most adverse feedwater malfunction in terms of cooling on the RCS This event, like the excess load event, is more limiting at EOC. This event has the same effect on the primary system as a small increase in turbine demand which is not matched by an increase in core power. As a result, the DNBR degradation associated with this event is less severe than that for the excess load where a large effective increase in turbine demand is analyzed. The excess load event analysis is reported in Section 5.6 in this riocument.
Steam generator tube plugging performed during a refueling outage has the potential for degrading the heat transfer characteristics assumed in Section 14.10.1 of the USAR for the Loss of Feedwater Flow Event.
Section 5.13 addresses the methodology to be employed should steam generator tube plugging exceed or be expected to exceed the assumptions of the current Loss of Feedwater Flow Event. Reduced heat transfer for the Loss of Feedwater Heating Event does not require reanalysis, since it is an overcooling event and the increase in plugged tubes reduces the consequences of the event.
OPPD-NA-8303-NP, Rev. 03 4 of 125
2.0l CHAPTER 14 EVENTS CONSIDERED IN THE RELOAD CORE ANALYSES (Continued) 2.2 ljSAR. Chanter 14. Safety Ar'alysis Events Not Considered in Reload Core Analyses (Continued) i 2.2.6 Steam Generator Tube Ruoture Accident The steam generator tube rupture accident is analyzed to determine if the offsite dose acceptance criteria of 10 CFR Part 10018 met. The analysis is a radioactive material release analysis based upon 1% failed fuel within )
the core. It is not dependent upon any reload core analysis related parameters, therefore, it la not analyzed in the reload core analysis. In the future, the steam generator tube rupture accident analysis may be verifled for high burnup fuel and/or a change in heat transfer characteristics for an increase in the number of plugged tubes in the generators.
2.2.7 Loss of Coolant Accident I The loss of coolant accident as reported in USAR, Section 14.15, is l 1
analyzed for OPPD by W, The large and small break u.~.alyses were performed by W using NRC approved Methods. A summary of the methods used by W for the large and small break LOCA analyses is provided in Reference 2-1. OPPD confirms the assumptions used in these analyses are valid for each reload core, if reanalysis is required, the reanalysis is done by a nuclear fuel vendor. OPPD does not perform any g loss of coolant accident analyses.
2.2.8 Containment Pressure Analvsis Containment pressure analysis is dependent upon the initial liquid mass 1 and energy contained in the primary or secondary system. Since these parameters do nct change when the core is refueled, the containment
^
. pressure analysis is not done in a reload core analysis.
2.2.9 Generation of Hvdrocen in Containment The generation of hydrogen in containment analysis is independent of any reload core parameters, therefore, the analysis is not performed during the course of a reload core analysis.-
2.2.10 Fuel Handlino Accident The fuel handling accident is a function of the isotopic inventory contained
. In the fuel pins. This is not normally considered in a reload core analysis, however, it may be necessary to reconsider this analyses for high bumup fuel.
OPPD-NA-8303-NP, Rev. 03 5 of 125
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l 2.0 CHAPTER 14 EVENTS CONSIDERED IN THE RELOAD CORE ANALYSES (Continued) 2.2 USAR. Chacter 14. Saletv Analysis Events Not Considered in Reload C.gre Analvses (Continued) 2.2.11 Gas Decav Tank Ruotum The gas decay tank rupture is independent of any parameter associated with refueling the core. Therefore, the analysis is not performed during a normal reload core analysis, 2.2.12 Wasto Liquid Event The waste liquid event analysis is not affected by refueling the core.
Therefore, the waste liquid event analysis is not performed in the course of a normal reload core analysis.
2.3 USArt Section 14. Events Considered in a Reload Core Analysis The reload core analysis consists of analyzing several events which are considered in the USAR and two events which previously were not analyzed in the USAR. These events are analyzed in accordance with the criterla discussed in this report and to determine if an unreviewed safety question would exist for a reload core. The USAR Chapter 14 events considered in a reload core analysis are the Control Element Assembly Withdrawal (CEAW) event, the boron dilution event, the Control Element Assembly (CEA) drop event, the loss of coolant flow event, the excess load event, the steam line break accident, the CEA ejection accident, RCS depressurization event and the seized rotor accident, in addition, an analysis is performed for incidents resulting from the malfunction of one steam generator. The analysis for each of these events will be discussed in detail in Section S.0 of this report.
3.0 TRANSIENT AND ACCIDENT ANALYSIS AND TECHNICAL SPECIFICATIONS Results of transient and accident analyses are used in the Technical Specifications in two ways. The first way is that values from the Technical Specifications are included in the initial conditions of the transient analyses. These Technical Specifications guarantee that the various transient and accident analysis acceptance criteria will not be exceeded if the reactor is operated within the bounds of these Technical Specifications. Technical Specifications of this type include the limits on F,, Fry, the PDIL and the Moderator Temperature Coefficient. The second type of values factored into the Technical Specifications are those that are determined by transient analysis. These parameters consist of the transient response term applied to the Thermal Margin / Low Pressure (TM/LP) trip equation, the minimum required shutdown margin, g the linear heat rate LCO and the DNBR LCO. The transient response term applied to the TM/LP OPPD-NA-8303-NP, Rev. 03 6 of 125
3.0 TRANSIENT AND ACCIDENT ANALYSIS AND TECHNICAL SPECIFICATIONS (Continued) equation in the Technical Specifications is a result of the analysis of the RCS depressurization g event. The minimum required shutdown margin at hot shutdown conditions is determined by the steam line break accident. This value is also confirmed for the boron dilution event. The minimum required shutdown margin for cold shutdown and refueling shutdown conditions is determined by the boron dilution event of the five percent subcriticallty requirement for refueling. The values used in the linear heat rate LCO are typically determined by the loss of coolant accident. These values are also confirmed for the dropped CEA event. The LCO on DNBR margin is calculated based on results from the dropped CEA analysis, the loss of four g pump flow analysis, the excess load analysis or the CEA withdrawal analysis, p 4.0 TRANSIENT AND ACCIDENT ANALYSIS MODELS OPPD utilizes the latest version of the CESEC code (CESEC-Ill and hereafter referred to as CESEC) and the CENTS code in the simulation of plant response to non-LOCA initiating eventa. OPPD utilizes the CETOP and TORC computer codes for calculation of DNBR during these events.
4.1 Plant Simulation Model OPPD utilizes the CESEC and CENTS digital computer codes, References 4-1, 4-2, and 4-11, to provide the simulation of the Fort Calhoun Station nuclear steam supply system. Both codes calculate the plant response to non-LOCA initiating events for a wide range of operating conditions. Additionalinformation on the CESEC modelis provided in Reference 4-3, The CESEC program, which numerically integrates one dimensional mass and energy conservation equations, assumes a node / flow-path network to model the NSSS. The primary system components considered in the code include the reactor vessel, the reactor core, the primary coolant loops, the pressurizer, the steam generators and the reactor coolant pumps. The secondary system components include the secondary side of the steam generators, the main steam system, the feedwater systern and the various steam control valves, in addition, tne program models some of the control and plant protection systems.
CESEC self initializes for any given, but constant, set of reactor power level, reactor coolant flow rate and steam generator power charing. During the transient calculations, the time rate of change In the system pressure and enthalpy are obtained from solution of the conservation equations. These derfvatives are then numerically Integrated in time under the assumption of thermal equilibrium to give the system pressure and nodal enthalples. The fluid states recognized by the code are subcooled and saturated; superheating is allowed in the pressurizer. Fluid in the reactor coolant system is assumed to be homogenous. Reference 4-1 provides a description of the i
OPPD-NA-8303-NP, Rev. 03 7 of 125
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4.0 TRANSIENT AND ACCIDENT ANALYSIS MODELS (Continued) 4.1 Plant Simulation Model (Continued)
CESEC code, including the major models, and the input, output and plot packages.
The pressurizer model is described in Reference 4-1 and further discussed in Reference 4-2. OPPD utilizes the wall heat transfer model to permit simulation of g voiding in any node in which steam formation occurs. Volding may occur in events such as a steam line break or steam generator tube rupture. Nodalization of the closure head, described in Reference 4-1 and further discussed in Reference 4-2, allows for the formation of a vold in the upper head region when the pressurizer empties. Flow to the closure head is terminated in simulations of those events in which natural circulation occurs and in those events such as the steam line break where this action delays safety injection.
The capabilities and limitations of the CESEC code are discussed in References 4-1 and 4-2. OPPD's CESEC model of Fort Calhoun Station is valid as indicated in g Reference 4-3 for the transients discussed in Section 5 of this report, with the exception of the CEA Elect lon Analysis and LOCA Analysis. The CESEC modelis also valid for analysis of the loss of load, malfunctions of the feedwater system and the steam generator tube rupture incidents.
The CESEC code is maintained by CE on the CE computer system in Windsor, Connecticut. OPPD accesses the code through a time sharing system. CE maintains g all documentation and quality assurance programs related to this code.
The CENTS primary system model is based on the design version of the CEFLASH-4AS code (Ref. 4-12), lhe thermal hydraulla response is modeled by a node and flowpath networ1<. Nodes enclose control volumes which represent fluid mass and energy. Flowpaths connecting nodes represent fluid momentum and have no volume. Nodes are provided to model primary system components such as the inner vessel, upper head, hot and cold legs, pressurizer, steam generator (separate nodes for hot and cold sides of tubes), reactor coolant pump suction legs, reactor vessel downcomer, and the control element assembly guide tubes. The secondary side is represented by three nodes for eacn steam generator (downcomer, evaporator, and steam dome) and one node for the main steam line header. The secondary system model also models the secondary safety valves, atmospheric and condenser dump valves, main steam isolation valves, turbine bypass and admission valves, and main and auxillary feedwater.
The RCS thermal hydraulle model is formulated with five one dimensional conservation l
OPPD-NA-8303-NP, Rev. 03 8 of 125
4.0 TRANSIENT AND ACCIDENT ANALYSIS MODELS (Continued) 4.1 Plant Simulation Model (Continued) equations. Conservation of mixture (tiquid and steam) mass, liquid mass, mixture energy, steam energy, and mixture momentum are all considered. Mass and energy for ilquid and steam are calculated for each node, while mass flowrate is calculated for each flowpath. Transient thermal hydraulic response la calculated by integrating the five conservation equations. Pressure in each nodo is calculated after solution of the conservation equations, CENTS also models phase separation within a node into a separate steam region and a liquid or two-phase region consisting of a continuous liquid phase with dispersed bubbles. Nodes with phase separation provide a discrete two phase mixture level in the node. And where appropriate, the fluid level imp 6 cts on heat transfer rates and on the quality of fluid mixture exiting through flowpaths connected to the node.
CENTS provides a full range of thermodynamic fluid states for all primary nodes.
Nodes with homogeneous or fully mixed fiuld are at equilibrium. The possible states for non-homogeneous or phase-separated nodes with separato two phase mixture and steam regions are (a) saturated liquid with saturated steam (equilibrium), (b) subcooled liquid with saturated steam, (c) saturated liquid with superheated steam, and (d) subcooleo dquid with superheated steam.
Core power is calculated by CENTS using either a point kinetics model or a three dimensional core neutronics model. In the point kinetics model, the calculated power is distributed axially according to a usst input axlal power shape. The tfvee dimensional model calculates a detailed power distribution with local power levels and fuel operating conditions for each fuel assembly.
CENTS features a flexible, modular method to handle control systems for the core, primary system, secondary system, and the reactor protective system. This method has been used to model the reactor protective system and other control systems for Fort Calhoun Station.
The capabilities and limitations of the CENTS code are discussed in References 4-11 and 4-13. OPPD's CENTS model of Fort Calhoun Station is valid for the transients discussed in Section 5 of this report, with the exception of the CEA Ejection Analysis and LOCA Analysis. The CENTS modelis also valid for analysis of the loss of load, malfunctions of the feedwater system and the steam generator tube rupture incidents.
OPPO-NA-8303-NP, Rev. 03 9 of 125
l 4.0 TRANSIENT AND ACCIDENT ANALYSIS MCDELS (Continued) 4.1 E13nt Simulation Model (Continued)
CENTS is maintained by OPPD on an OPPD workstation computer. CE maintains all g documentation cnd quality assurance programs related to this code. E 4.2 DNBR Analysis Models The DNBR analysis is currently performed using either the TORC code, Reference 4-4, or both the TORC and CETOP codes, Reference 4-5. The TORC code is used as a benchmark for the CETOP code model. TORC solves the conservation equations, es applied to a three-dimensional representation of the open lattice core, to determine the local coolant conditions at all points in the core. Lateral transfer of mass and energy between neighboring flow channels (open core effects) are accounted for in the calculation of local coolant conditions. These coolant conditions are then used with a Critical Heat Flux (CHF) correlation supplied as a code subroutine to determine the minimum value of DNBR for the reactor core. The CE-1 CHF correlation (References 4-6 and 4-7) is used for the Fort Calhoun reactor as approved in Reference 4-8. The Detailed TORC code Ic used directly in the seized rotor analysis.
The CETOP code has been developed to reduce the computer time needed for thermal hydraulic analyses while retalning all of the capabilities of the TORC desl0n model.
The CETOP model provides an addition: tmplification i to the conservation equations due to the specific geometry of the model. A complete description of the CETOP code is contained in Reference 4-5 and a description of OPPD's application of the CETOP g code is contained in Reference 4-9.
The fraction of inlet flow to the hot assembly in the CETOP model is adjusted such that the model yields appropriate MDNBR results when compared to the results of the TORC analysis for a specified range of operating conditions.
The CETOP code is used to calculate DNBR for all transient analyses discussed in Section 5 with the exception of the seized rotor analysis.
4.3 Acolication of Uncertainties Uncertaintles are taken into account either by deterministic or statistical methods. The deterministic method applies all uncertaintles adversely and simultaneously when calculating the approach to a limit.
Uncertainties in DNBR calculations are taken into account by statistical methods. The statistical method takes into account the likelihood that the uncertainties will all be adverse. The statistical method is discussed in Reference 4-10. In this method the 1
OPPD-N A-8303-NP, Rev. 03 10 of 125
l 4.0 TRANSIENT AND tiCCIDENT ANALYSIS MODELS (Continued) 4.3 Acolication of Uncertainties (Continusd) impact of component uncertainties on DNBR is assessed and the DNBR SAFDL is increased to include tne effects of the uncertainties. Since the uncertainties are accommodated by the increased DNBR SAFDL in the statistical method, engineering factors are not applied to the DNBR analysis model. The statistical method of applying uncertainties is applied in the CEA withdrawal, CEA drop, loss of RCS flow, excess load, seized rotor and asymmetric steam generator event DNBR calculations.
5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS This section addresses the evaluation of the various transients and accidents that are performed during a reload core analysis. Specific methods are desciibed for each transient and accident. For each accident or transient the following material is described:
A. Definition of the Event - A brief description of the causes, consequences, and RPS tilps involved in the incident.
B. Analysis Criteria - A brief description of the classification of the event and the Specified Acceptable Fuel Design Limit (SAFDL) or the offsite dose critoria which must be met, C. Objectives of the Analysis - A brief description of the methods that are used to assure that the critoria of the analysis are met.
D. Key Parameters and Analysis Assumptions - A description of the key parameters and assumptions used in the analysis.
E. Analysis Method - A description of the methodology employed by OPPD to analyze g the event.
F. Analysis Results and 10 CFR 50.59 Criteria - The expected results of the analysis and a discussion of the methods used to determine if the event meets the criteria of 10 CFR 50.59.
G. Conservatism of Results - A description of the conservatism of the analyslG.
The values of the trip setpoints and trip delay times used in these analyses are shown in Table 5.0- 1.
OPPD-NA-8303-NP, Rev 03 '
11 of 125
Table 5.0-1 REACTOR PROTECTIVE SYSTEM TRIPS AND SAFETY INJECTION Used in Analysis Setooint Uncertainty Delav Time (sec) Setooirt Irig 2.6 dec/ min 10.5 dec/ min 0.4 2.1 dec/ min High Rate-of-Change of Power 107 % 5.0% 0.4 112 %
High Power Level 9.1% above set 0.9 % 0.4 10% above Variable High Power Level initial power power level to a low of 19.1% fevel 0.65 93 %
Low Retetor Coolant Flow 95 % 12%
2400 psia 122 psi 0.9 2422 psia High Pressurizer Pressure 1750 psia 122 psi 0.9 1728 psia Thermal Margin / Low Pressure (1) 500 psia 122 psi 0.9 478 psia Low Steam Generator Pressure 31.2% of narrow 110 in. (5.7% of 0.9 25.5% of Low Steam Generator Water Level span range span narrow range span) l Steam Generator Differential 0.9 175 psid Pressure 135 psid 140 psi 5 psig 10.4 psi 0.1 5.4 psig Containment Pressure High 1600 psia 122 psi 12A 1578 psia High Pressure Safety injection (1) Values represent the low limit of the thermal margin / low pressure trip. The setpoint of this trip is discussed in Reference 5-3.
A Pump start -loop valve opening time.
OPPD-NA-8303-NP, Rev. 03 12 of 125
(
, 5.0 : TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 0.1 - CEA Withdrawal Evgg 5.1.1 Definition of the Event A sequential CEA Group W!thdrawal Event is assumed to occur as a result of a failure of the control element assembly drive mechanism control system or by operator error. The CEA Block System eliminates tne possibility of an out of j sequence bank withdrawal or single CEA withdrawal due to a single failure.
Any controlled or unplanned withdrawals of the CEAs results in a positive reactivity addHlon which causes the core power, core average heat flux and reactor coolant system temperature and pressure to rise and in turn decrease the DNB and Linear Heat Rate (LHR) margins. The pressure increase, if large enough, activates the pressurizer sprays which mitigate the pressure rise, in j the presence of a positive Moderator Temperature Coefficient (MTC) of l reactivity, the temperature increase results in an additional positive reactivity -
addition further decreasing the margin to 'he DNB and LHR limits. '
Withdrawal of the CEAs causes the axlai power distribution to shift to the top of the core, The associated increase in the axial peak is partially compensated by the corresponding decrease in the integrated radial peaking factor. The magnitude of the 3-D peak change depends primarily on the initial CEA configuration and axlal power distribution.
The withdrawal of the CEAs cauces the neutron flux as measured by the excore
- detectors to be decalibrated due to CEA motion, i.e., rod shadowing effects. ,
- This decalibration of -
r s
/
L
- i OPPD-NA-8303-NP, Rev. 03 13 of 125
I 5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5,1 :
CEA Withdrawal Event (Continued) 5.1.1 Definition of the Event (Continued)-
excore detectorsihowever, is partially compensated by neutron attenuation 1 rising from moderator density changes (i.e., temperature shadowing effects).- -l As the core power and heat flux increase, a reactor trip on high power, variable
, high power, or Thermal Margin / Low Pressure may occur to terminate the event depending on the initial operating conditions and rate of reactivity addition.
Other potential trips include the axlal power distribution and high pressurizer pressure trips, If a trip occurs, the CEAs drop into the core and insert negative reactivity which quickly terminates further margin degradatiort if no trip occurs and corrective action is not taken by the operators, the CEAs fully withdraw and the NSSS achieves a new steaoy state equilibrium with higher power, temperature, peak linear heat rate and lower hot channel DNBR value, 5,1,2 - Analysis Criteria The CEA Withdrawal (CEAW) event is classified as an Anticipated Operational
. Occurrence (AOO) for which the following criteria must be met:
A. The transient minimum DNBR is greater than the 95/95 confidence irterval limit for the CE-1 correlation, and B,
' The Peak Unear Heat Generation Rate (PLHGR) does not exceed 22 kw/ft (Reference 5-1);
5.1.3 Oblec6ves of the Analvsis -
The objeciives of the analysis performed for the *at power" CEAW event is to calculate the Required Overpower Margin (ROPM) which must be factored into -
the setpoint o'.ialysis.
The objective of the analyals for the hot zero power CEAW event is to demonstrate that the Variable High Power Trip (VHPT) is initiated in t!me to Insure that the analysis criterla are met.
OPPD-N A-8303-NP, Rev. 03 14 of 125 '
j 5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5,1 CEA Withdrawal Event (Continued) 5.1 A Kev Parameters and Analvsis Assumotiong l
The initial condit'ons assumed in the CEAW analysis are shown in Table 5.1-1, l
The reactor state parameters of primary importance in calculating the margin I degradation are:
A. CEA withdrawal rate * (i.e., reactivity insertion rate),
B. Gap thermal conductivity (Hgap),
C, initial power level, D. ' Flux power level determined from the excore detector response l l
during the transient, E. The moderator temperature coefficient of reactivity, and 3 F. Changes in the axla! power distrnution and planar and Integrated radial peaking factor during the transient.
- NOTE: ~ The term CEA withdrawal rate and CEA reactivity insertion rate are used interchangeably in this -
report.
The excore responses for each initial power level analyzed are based on the CEA insertions allowed by the Power Dependent insertion Limit (PDIL) at the selected power level, the changes in CEA position prior to trip, and the co7esponding rod shadowing and temperature attenuation (shadowing) factors, For the CEAW cases where combinations of parameters result in a reactor trip,'
the scram reactivity versus insertion chaiacteristics are assumed to be those
- associated with the core average axlat power distribution peaked at the bottom of the core. The scram reactivity versus incertion characteristics associated a 'with this bottom peak shape minlmize the amount of negative reactivity inserted during initial portions of the' scram following a reactor trip.
OPPD-NA-8300-NP, Rev. 03 15 of 125 -
t
- 5.0
TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) .
- 5.1 CEA Withdrawal Event (Continued) s 5.1,4- Kev Parameters and Analysis Assumotions (Continued) ;
All control systems except the pressurizer pressure control system and the pressurizer level control system are assumed to be in a manual mode. These are the most adverse operating modes for this event. The pressurizer pressure '
- control system and pressurizer level control system are assumed to be in the
}
I automatic mode since the actuation of these systems minimizes a rise la the coolant system pressure. The net effect, is to delay a reactor trip until a high power trip is initiated. This allows the transient increases in power, heat flux
. and coolant temperature to proceed for a longer period of time, in addition,-
minimizing the pressure increase is conservative in the margin degradation r
--calculations since increases in pressure would offset some of the DN8 margin
- degradation caused by increases 10 the core heat flux and coolant temperatures. ,
5.1,5 Analvsls Methodoloov -
- The methodology used for analysis of the CEAW event is described in CEN-121(B)-P, Reference 5-2 OPPD does not perform all parametric- a analyses discussed in Reference 5-2 for Fort Calhoun Station. Rather, OPPD '
r ' utilizes the analyses performed in Reference 5-2 to limit the number of
- analyses necessary for Fort Calhoun Station. Specifically, OPPD utilizes the g i result that [
V ..
+
) In addition, the result from Reference 5-2 that_ [ )
j when combined with [
-) can be used to perform sensitivity analyses on the CEA withdrawal rate Y to achieve [ .
j is utilized.
-i i '
n OPPD-NA-8303-NP, Rev. 03 r
16 of 125
, ,, ,m .-_- , - -..v- --, --- ,. +~v y -- w -- . m . t --
Y
. 5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS'(Continued)
Table 5.1-1 Key Parameters Assumed in CEAW Event Analysis l Parameter 1,ln[tg yglug Initial Core Power MWt 1 (HZP) /1530 (HFP)*
Inltlal Core Inlet Coolant 'F 532 (HZP)*
Temperature Maximum allowed by Tech. Specs.
Moderator Temperature Coefficient x10 dAp/'F Tech. Spec. Range Initial RCS Pressure psia Minimum allowed byTech. Specs.*
Fuel Temperature Coefficient x10 dap/*F Least Negative Predicted During a Cycle ,
Initial Core Mass Velocity- x108 lbm/hr Minimum allowed byTech. Specs.*
Fuel Temp. Coeff. Uncertainty .% -15.0
. Oap Thermal Conductivity BTU /hr-ft2 [ ]
CEA Differential Worth . x10 4 / Inch' [ ]
CEA Withdrawal Speed IrVmin - 46.0 Radial Peaks Maximum Allowed byTech Spec, for a Given Initial Power Level
- Scram Reactivity '%. Minimum Predicted During a Cycle
- High Power Trip Analysis Setpoint % of 1500 MWt .112.0
' Variable High Power Trip Analysis % Abovel Initial 10.0 Setpoint - Power Level Temperature: Shadowing Factor: % Power /'F .[ ]
- For DNBR calculations, effects of uncertaintles are combined statistically.
I 1
i i
=
OPPD-NA-8303-NP, Rev. 03 17 of j 125 a- . .
4 L 5.0 < TRANSIENT AND ACCIDENT' ANALYSIS METHODS (Continued) 5.11 CEA Withdrawal Event (Continued) 5.1.5 Analvsla Methodoloov (Continued)
- The rod shadowing factors for the Fort Calhoun Station full power case with Bank 4 inserted are the inverse of the. rod shadowing factors used in Reference l 5-2 (The rod shadowing factors for Fort Calhoun Station are such that the l
excore deiectors see more flux when the rods are withdrawn than when they are Inserted. Therefore, the [
] during a full power CEA withdrawal event). Because of this effect, it may be necessary to assume a [ ] in order to achieve [L
) The analysis at 'l Intermediate power levels is the same as documented in Reference 5-2. !
l
. The hot zero power CEAW event is analyzed assuming the variable high power trip is initiated at 29.1 % (19.1% plus 10% uncertainty) of rated thermal power. '
in addition, the analysis assumes that the maximum CEA withdrawal rate is ;
combined with the maximum differential rod worth. This case is anelyzed _
- using CESEC and the minimum DNBR la calculated using CETOP using the assumptions discussed in Reference 5-2.
The CEAW event analyzed to determirm .he closest approach to the fuel
- centerline melt SAFDL assumes those values of the CEAW rate and Hgap discussed in Reference 5-2. This combination of CEAW rate and Hgap was used to determine the PLHGR at all power levels.
]
5.1.6 : TVolcal Analvsls Results and 10 CFR 50 59 Criteria i 1
The results of the analyses of the CEAW event for Fort Calhoun Station at' full power and at Intermediate power levels are expected to be similar to those :
presented in Reference 5-2. The results of the hot zero power CEA withdrawal
- analysis are expected to be similar to those discussed in the Cycle 8 'r eload - !
submittal and the 1983 update of the USAR. The 10 CFR 50.59 criteria are met L if the analysis for the full power and intermediate power level CEAW events L shows that the required overpower margin for these events is less than the i available overpower margin required by tne current Technical Specification .
DNB and PLHGR LCOs. The 10 CFR 50.59 criteria is satisfied for the hot zero-power CEAW event if the minimum DNBR is greater than that reported in the latest submitted analysis.
3 OPPD-NA-8303-NP, Rev. 03 -
18 of' 125
l l
5.0 TRANSIENT AND ACCIDENT ANAL.YSIS METHODS (Continued) 5.1.7 C.ommausm of Rcatts Conservatism of the results of the CEAW Incident analysos is discussed in Reference 5 2 for the full power, intermodlato power level und hot zero poWor Cases.
5.2 Doron Dilution Event 5.2.1 D.chnition of Event Boron dilution is a manual oporation, cunducted under strict procedurcl controls which specify pom'issible limits on the rato and magnitudo of any required change in boron concentration Boron concontration in the reactor coolant system can be decreabod by olther controlled addition of unborated makoup water with a corresponding removal of reactor coolant or by using the deberating lon exchangers. To offect boron dilution the makoup controller modo selector of the chemical and volume control system (CVCC) must bo set to "dlluto" and then the dominorall zed water batch quantity selector tot for the desired quantity. When the opccific amount has boon injected, the dominoralized water control velvo is shut automatically. An Inativertent boron tillution can occur only if there la a combination of operator error and a CVCS malfunction occurring at tho same time. No RPS trips are assumed to terminate this incident.
5.2.2 Annivsis Criterin The boron dilution event is classified as an AOO for which the following critoria cannot be exc00ded:
A. DNBR greater than the 95/95 confidence interval limit using the CE-1 correlation, and B. The PLHGR less than 22 kw/ft.
5.2.3 @jectives of the Analysis The DNBR and PLHGR critoria are met by showing that sufficient time exists for the operator to take correctivo action to terminate the event prior to exceeding the SAFDLs. This is accomplished by calculating tho timo interval in which the minimum Technical Specification shutdovo margin is lost. The acceptable time Interval for the operator to take corrective actions beforo shutdown margin is lost aro 15 minutes for Modes 2,3 and 4 and 30 minutes in Mode 5.
OPPD-NA-8303-NP, Rev 03 19 of 125
1 i
5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.2 Boron O!!ution Eycat (Continued) 5.2.4 Key Parameters and Analy11!LA11umotions The boron dilution ovent at po+or (Modo 1) is bounded by the fastor reactivity insertion rate of the CEA withdrawal event and it !scks the local power peaking associated with the withdrawn CEA. For the boron dilution event in Modos 2 through 5, it is assumed that all three charging pumps are operating at their maximum capacity for a total charging rate of 120 gpm. For the d;lution at hot standby (Modo 2) the event is assumed to bo initiated at the Technical Specification shutdown margin requiremont at an RCS temperature betwoon 515 'F and 545 'F. The minimum volume of the reactor coolant system is conservatively assumed to be 5,506 cubic foot.
The boron dilution event betwoon hot and cold shutdown (Modos 3 and 4) is assumod to be initiated from the Technical Spoc'fication shutdown margin requiremont at an RCS temporaturo betwoon 210 'F and 515 'F. The boron dilution event at cold shutdown (Mode 4) is initiated from the Technical Specification minimum shutdown margin requirement at an RCS temperature between 68 'F and 210 *F. The analysis is conducted for two RCS volumos, one of 5,506 cubic foot and the other of 2,036 cubic f 001, which is the volume consoivatively assumed to be the volume for a refuo'ing operation conomon.
The analysis for the lower volumo cold shutdown condition assumos that shutdown groups A and B are withdrawn from the crto and all regulating groups are inserted in the core with the exception o'114 most reactive rod which is assumed to be stuck in its fully withdrawn position. Tnose assumptions are consistoct with the Tochnical Specifications for cold shutdown conditlor.e The boron dilution ovent during refueling is analyzed assuming that reactor refueling has just boon completed and the head is h place but the coolant volumo is sufflclont to only fill the reactor vessel to the bottom of the piping nozzles (2,036 cubic feet) and the minimum permissible boron concontration allowed by Technical Specification for refueling oxists. All CEAs are withdrawn from the core.
Those assumptions represent shutdown conditions for the various modos wheroin the core reactivity is greatest, the water volume and total boron content is at a ininimum, and the rate of dilution is as large as possible.
Hence, these conditions represent the minimum time to achlove inadvertent criticality in the event of an uncontrolled boron dilution.
OPPD-N A-8303-NP, Rev. 03 20 of 125
5.0 TRANSIENT AND ACCIDENT ANALYSid METHODS (Continued) 5.2 Bcron D.llullonEycal(Continued) 5 2.5 AnalystsRettleds The method used to calculate the dilution time to criticality from Modos 2 ttvough 5 is ttvough the use of the following equation:
to,n = m *tno*In(( CBC + SDM
- IDW) / CBC ) E Whore too = boron dilution time constant, which is a function of RCS volume and temporature (soc) l CBC = critical boron concentration (ppm) '
SDM = shutdown marg!n (% Ap)
IBW = Inverso boron worth (ppm /ap) m = multiplier to conservatively account for density effects due to the !
temperature difference betwoon the makoup water and the reactor coolant water. The multipiler is the ratio of the specific volumo of the makoup water to the specific volume of the reactor coolant water at the highest temperature in the range for the modo of oporation of Interest. Makeup water is assumod to be at room temperaturo (68 'F).
As can be seen from this equation, the dilution time to criticality is minimlzod with a greator critical boron concentration, a smaller inverse boron worth, or a smaller dilution timo constant. E 5.2.6 Analy11SEESulls and 10 CFR 50.69 Cr;tetig The analysis results are similar to those reported in the Cycle 11 safety analysis report and in the 1987 updato of the USAR. The critoria of 10 CFR 50.59 ate satisfied if the Technical Specification requirements on shutdown margin and the refueling boron concontration are unchanged as a result of this analysis.
5.2.7 Conservatism oLBeaulta Because of the proceduros involved in the boron dilution and the numerous alarm indications available to the operator, the probability of a sustalnod or erronoous boron dilution is very low. There is usually a large interval betwoon the calculated time and the timo limit for the boron dilution at hot standby and hot shutdown modes. Thoroforo, the results show considerable margin to tho l limit. The calculated time to critical for l
OPPD-NA-8303-NP, Rev. 03 21 of 125
v s
5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.2 Boron Dllution Eygal (ContinaJed) l 1
5.2.7 conservatiam of nonutts (Continued) the boron dilution at cold shutdown with the minimum RCS volume is -;
reasonably close to the acceptance criterla; however, the event is analyzed with only shutdown groups A and B being fully withdrawn from the core. Cold ,
shutdown is normally achieved with the shutdown groups A and B fully inserted '
in the core and, therefore, the core has a much lower k,n than assumed in the analysis. The boron dilution at refueling 18 conservative since 11 is improbable *
~ that more than a few CEAs will be removed at any one time during a refueling and the approach to critical following refueling is done under strict '
administrative control with only one bank of CEAs removed at a time. The '
analysis assumes that all CEAs are witharawn from the core, 5.3 Control Element Annembly Droo Evtigt -
5.3.1 Definttloa gttygal $
n The control element assembly (CEA) drop event is defined as the inadvertent j release of a CEA causing it to drop into the reactor core. The CEA drive is of the rack and pinion type with the drivo chaft running parallel to and driving the _
rack tryough a pinion gear and a set of bevel gears. The drive mechanism is equipped with a mechanical brake which maintains the position of the CEA.
The CEA drop may occur due to an inadvertent interruption of power to the l
CEA drive magnetic clutch or an electrical or mechanical failure of the
- mechanical brake in the CEA drive mechanism when the CEA is being moved.
'The full ;ength CEA drop event is classified as an AOO which does not require an RPS trip to provide protection against exceeding the SAFDLs.1he CEA drop results in a redistribution of the core radial power distribution and an increase in the radial peaks which are not directly monitored by the RPS and which are not among those analyzed in determining the DNB and LHR LCOs and LSSSs. As such, Inltlal steady state margin must be bullt into the : 1
- Technical Specification LCOs to allow the reactor to "rlde out" the event
.without exceeding the DNBR and LHR SAFDLs.
5 i
f OPPD-NA-8303 NP, Rev. 03 22 of 125
, , , -..,..-n,,- .---,...~,_....,.--n.,,. -
. %. ..,-.._,,.,m_...w . -,,r-m- -.*me.-,..-m--,.-._m._--.,e , , ,_ _ . ....m , , . . ..,_..,.,,--w----,-.-
i I
i l
5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.3 Control Element Assembiv Dron Event (Continued) 5.3.2 - Analys!n Criteria The full-length CEA drop event is classified as an Anticipated Operational Occurrence for which the following criteria must be met:
i A. The transient minimum DNBR must be greater than or equal to the I 95/95 confidence Interval limit, using the CE-1 correlation, and B. The Peak Linear Heat Rate (PLHR) must be less than or equal to 22 !
kw/ft.
5.3.3 Oblectives of the Analvsla The objective of the analysis is to determine the Required Overpower Margin (ROPM) which must be built into the LCOs to assure the DNBR and LHR ,
SAFDLs are not exceeded for the CEA drop which produces the highest I
distortion in the hot channel power distribution. Since the ROPM is dependent upon initial power level, rod configuration and axlal shape index, an analysis parametric in these variables is performed.
5.3.4 Kev Parameters and Analvals AnatunoljQDa l
Table 5.3.4-1 contains a list of N key parameters assumed in the full-length j CEA drop analysis. Assumptions used in the analysis include:
A. The charging pumps and proportional heater systems are assumed to be innperable during the translent. This maximizes the pressure drop during the event.
B.: The roc 8 block system la assumed to prevent any other rod motion during the translent.
C. The turbine admission vaVes are maintained at a constant position during the transient < This is because the turbine admission valve posit lon is set manually at Fort Calhoun Station and, therefore, the turbine
- admission valves wl!! not automatically open in response to a redtced electrical generation output. -
OPPD-NA-8303-NP, Rev. 03 23 of 125
. . - _ , _ - - . . ~ , _ _ . _.___..___u_. _ . - ~ . _ . . . _ _ _ . _ . _ _ _ . . . _ _ . . . - - . - - _ . . . _ - _
t i
l 5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) !
- 5.3 Control Element Assembly.Qton Event (Continued) l 5.3.5 Ana!vsla Method I 1
The analysis methods utillzed by OPPD to analyze the CEA drop event are g discussed in Section 8 of Reference 5-3.
5.3.0 Analvsin Resulta and 10 CFR 50 50 Crlierla Typical snalysis results are contained in Section 8 of Reference 5-3 and in the l
1987 update of the Fort Calhoun Station Unit No.1'USAR. The criteria of 10 !
CFR 50.59 are rnet if the required overpower margin calculated for this incident is less than the overpower margin being maintained by the current Technical Specifications.
1 5.3.7 QQQ19fvallam of Renutta l l
The following areas of conservatism are included in the analysis:
A. The most negative moderator ( ) coefficients of reactivity are utilized because these coefficients produce tne mlnlmum RCS coolant temperature decrease.
- B. The[ . ) distortion factor at any time during core life is combined with the [ ] CEA worth at any time during core life.
C. The moderator temperature coefficient assumed in the analysis is the most negatNe value allowed by the Technical Specifications; The' actual end of life value, including measurement uncertainty, is less
. negative.
D. . The manual mode of the pressurizer pressure and level control systems are assumed in the analysis, if the AUTO mode of operation is assumed, the RCS pressure would be maintained at a higher va!ve, i thereby lowering the DNBR margin required for this event,
.I OPPD-NA-8303-NP, Rev 03 24 of 125
k l i
4 l
) 5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) -
Table 5.3.4-1
)
KEY PARAMETERS ASSUMED IN THE FULL LENGTH CEA DROP ANALYSIS i 1
ParametCI Ulita Value initial Core Power MWt 1500*
Inlllal Core inlet 'F Maximum allowed *
- Temperature by Tech. Specs. i Initial RCS Pressure pela Minimum allowed
- by Tech. Specs, initial Core Mass Flow Rate x10*lbm/hr Minimum allowed
- by Tech. Specs.
1
-+ Moderator Temperature x10 'ap/'F Most negative Coefficient allowed by Tech.
i' Specs.
- CEA insertion % insertion Maximum allowed by Tech. Specs.
Radial Peaking Distortion _-
Maximum value predicted
.i Factor. during core life Dropped CEA Worth %Ap Minimum value predicted during core life for the CEA producing the maximum distonion factor Core Average Hgap - BTU /hr-ft2 *F Maximum value predicted -
during core life.
- Fuel Temperature x10'ap/*F Most negative value pre- -
Coefficient dicted during core life
'
- For DNBR calculations, the effects of uncertainties on these parameters are combined statistically.
OPPD-NA-8303-NP, Rev. 03 25 of 125
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5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.4 four-Pumn La$3 of Flow Event 5 4.1 Defin! tion of the Eypal "he four-pump loss of Coolant flow event is initiated by the $lmultaneous loss of electrical power to all four reactor coolant pumps. The loss of AC power to reactor cool 6nt pumps may result from eitner the completo loss of AC power to the plant, or the failure of the f ast transfer breakers to close after a loss of offsite power.
Reactor trip for the loss of coolant flow is initiated by a low coolant flow rato as dolormined by a reduction in the sum of the steam genorhtor hot to cold log pressure drop. This cignal is compared to a sotpoint wtilch is a function of the number of reactor coolant pumps in operation (which curiont Technical Specifications requito to be four). A reactor trip would bo initiated when the flow rato drops to 93% of full flow (95% minus 2% uncertainty),
5.4.2 Analygirdtlinda The four-pump loss of flow event is classified as an AOO for which the transient minimum DNBR must be greator than the 95/95 porcent confidence Interval limit using the CE-1 conolation.
5.4.3 OMCC11yfts of the Analy31g The objective of the analysis is to dolormine the required ovorpower margin that must be built into the DNB LCOs such that in conjunction with the low flow trip the DNBR SAFDL is not exceeded. Since the required overpower margin is dopondent upon both axial shape index and the CEA rod configuration, an analysis paramotric in those paramotors is performed.
5.4.4 Kev Parameters and Analysis Assumotions The closost approach to the DNBR SAFDL occurs for a loss of flow event initiated from the full power conditions. Tablo 5.4.4-1 gives the key paramotors used in this analysis. The flow coast down la calculated in the CESEC code.
OPPD-NA-8303-NP, Rev. 03 20 of 125
i:
5.0 TRANSIENT AND ACCIDENT ANALYSIS ME'iHODS (Continued) 5.4 Four-Pumo Logg of Flow Event fContinued) 5.4.5 Analvala Method 1 The analysis method used by OPPD to analyze the four-pump loss of coolant ,
flow is discussed in Section 7 of Reference 6-3. OPPD utilizes the CESEC-TORC method to analyze axial power distributions characterized by both negative and positive shape indices. The STRIKiN-TORC method is not
' utilized by OPPD because of the high rotational energy of the pumps (N =
1185 rpm, I = 71,000 lb-ft2/ pump). OPPD also utilltes the [
}
6.4.6 Analysis Beaulit.SQd 10 CFR 60.59 Critorig f
Expected analysis results are presented in Section 7.1 of Reference 6 3. The main difference between these results and the results for Fort Calhoun Station is that the ROPM will be significantly reduced for Fort Calhoun Station. This is because of the higher rotational energy of the Fort Calhoun reactor coolant pumps. The criteria of 10 CFR 50.59 are met if the required overpower margin calculated for the four-pump loss of coolant flow ev3nt is less than the !
, overpower margin being maintained by the current Technical Spocifications.
5.4.7 Conservattam of Renutta A. - Field measurements of the CEA magnetic clutch decay is more rapid than assumed in the safety analysis.
B. The available scram worth is higher than assumed in the safety analysis.
C. The MTC at full power is more negative than the value assumed in the safety analysis.
D. The actuct CEA drop time to 90% inserted is faster than that assumed in the safety analysis, E. The conservatism of the CETOP calculations is discussed in Section 7 of Reference 5-3.
OPPD-NA-8303-NP, Rev. 03 27 of 125 -
, . - .- . . . - - - . - . . . - - - . _ . ~ . . . - . . - . - - . - . . . _ - . . .- . . _ . _ .
5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued)
Table 5.4.4-1 KEY PARAMETERS ASSUMED IN THE LOSS OF COOLANT FLOW ANALYSIS Paramelet unlis __Natue initial Coro Power MWt 1500*
Initial Coro inlet Maximum allowed
- Temperaturo *F By ' loch. Specs, inillal RCS Prossure psia Minimum allowed
- by Toch. Spoca, Initial Coro Mass x10a Ibm /hr Minimum allowed
- Flow Rato by Toct. i. Spocs.
Modorator Temporaturo x10 ' ap/*F Maximum ellowed Coefficient by Toch. Spoca.
Fuel Temperaturo x10 'A p/*F Least nogative Coefficient prodicted during core life.
Low Flow Trip Delay Time soc Maximum l CEA Drop Time soc Maximum allowed i by Toch. Specs.
l Scram ReactMty Worth %op Minimurn predicted during coro lifetimo Scram ReactMty Consistent with axial i
Curve shapo of interest Core Average Hgar. BTU /hr-ft2 *F Minimum predicted during coro lifetimo
- For DNBR calculations, offects of uncertainties on thoso paramotors woro combined statistically, L
l l
OPPD-NA-8303-NP, Rev. 03 28 of 125
(
l 5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.5 Asymmetric Steam Generator Event 5.5.1 Definition of the Event The asymmetric transients arising from a secondary system malfunction in one steam generator result in changes in core power distribution which are not inherently covered by the TM/LP or APD LSSS. Consequently, these events must be analyzed to determine the initial steady state thermal margin which is l built into and maintained by the Technical Specification LCO such that l
assurance la provided that the DNBR and peak linear heat rate SAFDLs are not !
exceeded for these transients. The four events which effect the steam generator are:
= -)
A. Loss of load to one steam generator. -1 I
B. Loss of feedwater to one steam generator, j C. Excess feedwater to one steam generator.
D. Excess load to one steam generator.
The possible RPS trips whlch can occur to mitigate the consequences of these -
ever,ts include the low steam generator level, TM/LP, low steam generator pressure, and the asymmetric steam generator transient protection trip function (ASGTPTF). The particular trip which Intervenes is dependent upon the event .
initiator and the initial operating conditions, The ASGTPTF trip was instand in the Fort Calhoun Station RPS prior to operation of Cycle 9 to reduce the margin requirements associated with these ;
asymmetric events and to insure that these events do not become a limiting :
AOO for establishing initial margin which must be maintained by the LCO. A system description of the ASGTPTF is presented in Appendix B of Reference 5-3.
5.5.2 Analvsin criteria
- The asymmetric steam generator events are classified as AOOs for which the
. following crlierta must be met:
A, The translent minimum DNBR must be greater than or equal to the 95/95 confidence Interval limit using the CE-1 correlation, and B. The peak linear heat must be less than or equal to 22 kw/ft.
OPPD-NA-8303-NP, Rev. 03 29 of ' 125
- z. a- . . - . . . . -
____-_-____i
'i
. 5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 'l
- 5.5 Anvmmetric Steam Generator Event (Continued) 1 1
5.5.3 Oblectives of the Analvsla The objectives of the analysis are to determine the required overpower margin that must be built into the LCOs such that in conjunction with the ASGTPTF the 4
DNBR and PLHGR SAFDLs is not exceeded.
5.5.4 Kev Parameters and Analvtic Annumot}QDG Section 7 of Reference 5-3 demonstrates that the loss of load to one steam generator (LUISG) is the limiting asymmetric steam generator transient for establishing inillal steady state thermal margin which must be maintained by the Technical Specification LCO. Therefore, information is only provided for this asymmetric steam generator event. The key parameters used in the analysis of the LUISG event are given in Table 5.5.4-1. The charging pumps j and proportional heater systems are assumed to be inoperable during the trans!ent. This maximites the pressure drop during the event. The turbine i admission valves are assumed to maintain a constant position throughout the event since the turbine control system at Fort Calhoun utilizes manual setting i of the turbine admission valves.
l 5.5.5 Analysis tdttbQd l
The method utilized by OPPD to analyze the LUtSG 18 discussed in Section 7. J of Reference 5-3.
5.5.6 L Annivsis Reg (Ita t and 10 CFR 60 59_Crlierla f
lhe results of the analysis for the LUISG event are discussed in Section 7 of Reference 5-3. -
.The results for Fort Calhoun Station are expected to be similar. The criteria of 10 CFR 50.59 are satisfied if the required overpower margin calculated for the-LUISG event is less than the overpower margin being maintained by the -
current Technical Specifications.
1 OPPD-NA-8303-NP, Rev. 03
'30 of 125
. _ _ _ _ _. . _ . - _ . - _ . _ _ _ _ . .J . _., _ _ _ _._-. _ _. _ .. _. _ -,- _ _... .-_ _
l l
l 5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) i Table 5.5.4-1 l KEY PARAMETERS ASSUMED IN THE LUISG EVENT i
EnfMitlCI UnitG VoluD Initial Coro Power "". 1530*
Initlai Coro inlet *F Maximum allowed
- j Temperaturo by Toch. Spocs.
Initial Roactor Coolant psia Minimum allowed
- i System Pressure by Toch. Spocs. l Modorator Tomporaturo x10 'Ap/*F Most negativo allowed l
Coefficient by Toch. Specs.
Fuel Temperaturo x10'op/*F Mobt negativo predicted Coefficient during core life, i
Coro Avorago H g , BTU /hr-ftt *F Maximum valuo prodicted during coro lito, inillal Coro Mass x1061bm/hr Best estimato flow
- Flow Rato Scram Reactivity Worth %op Minimum prodleted during core lifo.
- For DNBR calculations, offects of uncertaintios on those parametors were I combined statistically.
OPPD-NA-8303-NP, Rov. 03 31 of 125
3 I
l 5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) j 5.6 ' Excess, Load Event i
5.0.1 Definition of Evbnt 1 l
An excess load tran6iont is cettned as any rapid increase in the steam generator steam flow other than a steam line break. Such a rapid increase in steam flow results in a power mismatch between the reactor core and the steam generator load demand. In addition, there is a decrease in the reactor )
coolant temperature and pressure. Under these conditions the negative I moderator temperature coefficient reactivity cat 1ses an increase in core power and heat flux This results in a decrease in DNB margin and an increase in '
LHR. In Cycle 14 the excess load event was reclassified from a (
) event to a ROPM event, it should be noted that margin requirements l based on explicit AOO transient analyses are absolute quantitles. The use of
]
one method versus the other method does not result in a net gain in mar >)ln, !! '
only transfers the margin requirement from the LSS9 to the LCO or vice gersa. -
These is no gain in both LSSS and LCO space.
' The rapid opening of the turbine admission valves or the steam dump bypass i to the condenser causes an excess load event. TLarbine valves are not sized to accommodate steam flow for powers much in excess of 1500 MWI. The steam !
dump valves and steam bypass valves to the condenser are sized to accommodate 33% and 5%, respectively, of the steam flow at 1500 MW.
Therefore, the following load increase events are examined:
A. Rapid opening of t'he turbine control valves at powen The maximum increase in the steam flow due to the turbine control valves opening is - ,
- limited by the turbine load limit control. The load limit control function la used to maintain load, so unless valve failure occurs, the control -
valves will remain where positioned.
B. Opening of all dump and bypass valves at power due to' steam dump -
- control Interlock failure: The circuit between the steam dump controller and the dump valves is open when the turbine generator is on linec -
Accidental closing of tr o steam dump control Interlock undor full load conditions, according t'o the temperature program of the controller, causes full opening of the dump and bypass valves. Since the reactor ,
coolant temperature decreases during the event, these valves will be closed again after the average reactor coolant temperature decreases -
to 535'F..
OPPD-N A-8303-NP, Rev. 03 32 of 125 s
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5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.0 EECL.S.LOM.hent (Continuod) 5.0.1 Ochaltlen.0theat (Continvod)
C. Opening of the dump and bypass valvos at hot standby conditions due in low reference temporature setting in the steam dump controllor:
When the plant is in hot standby conditions the dump valvo controller is operative but does not act because the hot standby temperaturo is lower than the lowest value required to opon the valves. At hot standby the roactor coolant temperaturo is $32*F, which 10 8'F below the minimum temperaturo toquirod to opon the dump and bypass valvos (540'F). The maxirnum error that can bo introduced in the referenced temporature sotting 18 limited to 17'F sinco a narrow rango instrument is used for this purpose. Reducing the dump valvo controllor reference sotting from 532' te 516' would result in a partial opening of tho va!vos but as soon an the reactor coolant temperaturo dropped to $10'F the valves would a00ln bo completely closed.
D. Oponing the dump and bypass valvos at hot standby due to steam dump controllor malfunction: The most sovoro ovont at hot standby would occur in the event the steam dump valvo controllor yields an incorrect signal and causes the steam dump and bypasa valvos to opon complotoly. This caso is concidorod to bo inuch less probable than caso C abovo but toprosents the most limiting event unoor hot standby conditions.
The possibio RPS trips that might be encountored during this event are:
- 1. Variablo high power trip (VHPT).
- 2. TM/LP tdp.
- 3. Low steam generator water levol trip.
- 4. Low stnam gonorator pressure trip.
The RPG trip initiated to mitigato the consequencos of the event will depend upon th'; inillal conditions and the rate of reactivity insortion duo to modorator feedback offects. The TM/LP and VHP trips uso the auctionoorod (higher) valuo of measured power from olther tho ex coro noutron flux power detectors or the AT power calculator. The ex-core power detectors may becomo decalibrated due to the tempvaturo shadowing. The AT power input responds slowly due to the relatively long timo constants assumod for the Resistance OPPD-NA-8303-NP, Rov. 03 33 of 125
5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 50 Extf;L:LLQa1Eyc01(Continued) 0.0.1 DRItalliQ[LQLEvent (Continued)
Tempotaturo Detectors (HTDs) in the hot and cold logs of the reactor coolant g systom.5.0 2 3 AnalyLLLCultua The excess load event is classified as a AOO for which the following critoria must be mot:
A. The transient minimum DNDR must be greater than or equal to the 95/95 confidence intomal limit using the CE-1 cortolation.
D. The peak linear heat rc'o (PLHR) must be less than or equal to 22 kw/ft.
50.3 Otdecitycs_DLthdanlyils An objectivo of the analysis is to calculato a ROPM value whictils factored into the sotpoint analysis to ensure that the DNOR and LHR SAFDLs are not exc00ded for excess load events for which the TM/LP does not provido protection.
The other objective of the analysis foi the limiting oxcess bad event 18 to demonstrate that tho Variablo High Power Trip (VHPT) is initiated in timo to insure the analysis critoria are mot.
5 0.4 Koy Parnalgters and Analyn[dgsumotions notorenco 5-2 discussos a similar sensitivity study performod by CE consistant with Refcronco 5 3 to demonstrato that the maximum calculated ROPM for the excess load event occurs for the (
) at hot full power conditions. OPPD sensitivity studios show similar results. Thorofore, only the hot full power caso is analyzed. The key paramotors used in the analysis of the excess load event are glvon In Tablo 5.0.4-1. The romaining assumptions are the same as thoso discussed in Reforence 5-3. The roactor stato paramotors of primary importance in calculating the margin degradation are:
A. Axial Power Distributions, B. initial Coro inlet Temperature, OPPD-NA-8303-NP, Rev. 03 34 of 125
5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.6 Excets Loa 1Evrat (Continued) 5.6.4 Key Parameters and AnalyLMSktDR11Qas (Continued)
C. Initial power level.
D. Flux power level dotormined from the excoro detector response during the transient, E. The modorator temperature coefficient of reactivity, and F. Changes in the axial power distribution and planar and integrated radial peaking factor during the transient.
5.6.5 Annivsis MelbQd The stops used for determining the [ ] value and calculating the largest ROPM for all excess load events which rely on the TM/LP trip for DNBR g protection are glvon in Section 5 of Reference 5-3.
The method of providing DNB protection is to build DNB margln into the DNB LCOs. The excess load event is protected by the RPS and sufficient initial margin which is maintained by the LCOs. The DNB ROPM is calculated and compared to the DNB ROPMs of the other AOOs to determine the limiting ROPM which should be incorporated into the setpoint LCO calculations.
This method ensures that sufficient margin is maintalnod by the LCOs in conjunction with the VHPT, Hence, the DNB and LHR ROPMs will be calculated based relative to the VHPT sotpoint. To dotormine a conservatively large margin degradation the initial conditions are set such that a VHPT setpoint is reached when both the AT power calculator and the power from the ex-core detectors are equal. This is accomplished by determining the [
] value as described in Section 5 of Reference 5-3.
To futhor ensure that the most limiting case has boon analyzed, the following key assumptions are modo Since tho ex-core detectors may becomo decalibrated due to the [ ), the [
] factor is used to attenuate the ex-core detector response during the cooldown event. This resutts in a [ ] trip and [ ] ROPMs.
Figure 5-1 providos a block diagram of the methods used to calculate the DNB and LHR ROPMs for the excess load went.
l OPPD-NA-8303-NP, Rev. 03 35 of 125
5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued)
Tablo 5.6.4-1 KEY PARAMETERS ASSUMED IN THE EXCESS LOAD EVENT ANALYSIS Parameter Unfts _Valug initial Core Power MWt 1530*
Initial Coro inlet *F Maximum allowed
- Temperature by Tech. Spoca, initial Reactor Coolant psla Minimum allowed
- System Pressure by Tech. Specs.
Initial Coro Maso x10*lbm/hr Minimum allowed
- Flow Rate by Tech. Spocs.
Axial Shapo index aslu Most Nogative allowed by DNB LCO Tont RTD Dolay Timo see Minimum Hot Log Maximum Cold Log Moderator Temperaturo x104p/'F Negativo values up Coefficient to the most negativo valuo allowed by Tech. Specs.
Radial Peaks Maximum Allowed byToch Spoc. for a Given initial Power Levol Scram ReactMty % Minimum Predicted During a Cyclo High Power Trip Analysis Setpoint % nf 1500 MWt 112.0 Variablo High Power Trip Sotpoint % Above initial 10.0 Power Level Temperaturo Shadowing Factor % Powor/'F ( )
- For DNBR calculations, effects of uncertainties on theso paramotors woro combined statistically.
OPPD-NA-8303-NP, Rov. 03 36 of 125
I 5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.6 Eres1 Loa 1Eycat (Continued) i 50.5 An@ilLidett1Rd (Continued)
]
l The PLHR is calculated by obtaining the core ever so tinoar heat rate at temo l of peak coro power and muttiplying it by the appropriato peaking f actors and associated uncertainties.
50.6 An$sitf1cti,11ts and 10 CFR3039_Cutclla The results of the excess load analysis are similar to thoso presented in Reforonco 5-2. The critoria of 10 CFR 5159 are mot if the ROPM calculated for tfils event is loss than or equal to the overpower margin being maintained by the cunent Technical Specifications.
5.6.7 Contctyattstagliltuults The following points demonstrate the conservatism of the overall results for the excess load event:
A. Tho VHPT is not assumod until the right hand AT Power calculator and the left hand AT Power calculator values are both above the setpoint.
B. The actual scram worths are highor than those in the analysis.
C. Wooro the most negativo MTC is used, the value is more negativo than that monsured during plant operation.
D. The actual Doppler reactMty is more negative than assumod in the analysis.
E. The scram reactivity curve associated with the strong positivo axial power distribution is conservative with respect to the actual power distributions observod in the reactor.
F. The pressurizer levol control is assumed to bo in the manual modo (pressurizer heators off). This maximizos the RCS pressure docrease which maximizos the ROPM for the event.
OPPD-NA-8303-NP, Rev. 03 37 of 125
m _ _ _. . _ _ - _ _ _ _ _ _ _ _ _ _ . _ _ _ .__ _
l I
5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.7 ACS Depreanutgat!Qafvtut 5.7.1 Definition of Event The RCS depressurization evont is characterized by a rapid decrease in the prirnary system pressure caused by either the inadvertent opening of both ,
power operated relief valves (PORVs) or tne inadvertent opening of a single primary safety valve operating at rated thormal power. Following the initiation of the oyent, steam is discharged from the pressurizor steam space to the quench tank whoro it is condensed and storod. To compensate for the decreasing pressure the water in the pressurizor flashes to steam and the proportional heators increase the heat added to the water in the pressurizer in an attempt to maintain pressure. During this time the pressurizer level also
- bog!ns to decrease causing the letdown control valves to closo and additional charging pump 9 to start so as to maintain lovel. As pressuro continuos to drop, the backup heators energize to further assist in maintaining primary pressure. A reactor trip is inttlated by the TM/LP trip to provent exceeding the DNBR SAFDL.
5.7.2 Analysis Cutada The RCS depressurization event is classified as an AOO for which the transient minimum DNBR must be grostor than or equal to the 95/95 percent confldonce interval limit using the CE-1 coriolation.
5.7.3 Oblecityes of tho,fQglygig This event is classified as an ADO for which thoto must be suffielent margin built into the TM/LP trip such that the DNBR SAFDL is not excooded. The objective of this analysis is to calculate a conservative [ ] for incorporation into the TM/LP equation.
5.7,4 Kev Parameters and Analvsle Assumotions The key parameters for the RCS depressurization ovent analysis are glvon in Table 5.7,4-1. Additional assumptions are discussed in Section 5 of Reforonce 5-3, OPPD-NA-8303-NP, Rev. 03 38 of 125
l l
5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued)
Table 5.7.4-1 KEY PARAMETERS ASSUMED IN THE RCS DEPRESSURIZATION EVENT ANALYSIS l
Parameter Unlig _Ydyg l inillal Coro Power MWt 1530*
Initial Coro inlet *F Maximum allowed by*
Tomporaturo Toch. Spocs.
Infllal Reactor Coolant psia Uppor limit of normal l Gystem Pressure oporating rango l Moderator Temperaturo x10 'ap/'F Most negativo allowed Coefficient by Toch. Specs.
Fuel Temperature x10 'ap/'F Most nogative prodicted Coofficient during coro lifo.
Core Averago Hgap DTU/hr-ftt 'F Minimum prodlcted during coro lifo.
t Total Trip Dolay sec 1.4 Timo
- For DNDR ca cciations, offects of uncertaintios on those paramotors were combined statistically.
OPPD-NA-8303-NP, Rov. 03 39 of 125
5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.7 RCSacatetturtration Event (Continued) 5.7.5 Anaty?.dn. Method The methods used by OPPD to analyze the RCS depressurization event are g
contained in Section 5 of Referonce 5-3.
5.7,0 Anslytis Results and 10 CFR 50,M Results of the RCS depressurization transient are discussed in Reference 5-3 and in the 1984 update of the Fort Calhoun Station Unit No.1 USAR. The critoria of 10 CFR 50.59 are satisfied if the [ ] is less than or 1 equal to the value used in the current TM/LP trip equation. !
5.7.7 C2Qantyatitm of Rgaults I The conservatism of the calculated pressure bias term is obtained by using the l combination of the following conservative key parameters: 1 A. Conservativo scram reactivity characteristics are used in the analysis.
B. Conservatively slow RPS response times are uced.
C. Conservatively high primary relief or safety valve areas are used.
D. The RCS pressure 18 initia!!y assumod to be in its upper limit as opposed to the normal operating pressure.
5.8 Main Steam Une Break Aceg[ cat 5.8.1 Definition,st the Event A large break of a pipe in the main stoam system causes a rapid depletion of steam generator inventory and an increased rato of heat extraction from the primary system.
The resultant cooldown of the reactor coolant, in the presence of a negative moderator temperature coefficient of reactivity, will cause an increase in nuclear power and trip the reactor. A severo decreaso in main steam pressure will also initiate reactor trip and cause the main sisam isolation valves to close, if the steam line rupture occurs betwoon the isolation valvo and the steam generator outlet nozzle, blowdown of tne affected steam generator will continue. (However, closure of the check valve in the ruptured steam line, as well as closure of the isolation valves in both steam lines, will terminate blowdown from the intact steam generator).
OPPD-NA-8303-NP, Rev. 03 40 of 125
i 5.0 TRANS!ENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.8 Maltt$leam Line Break Accident (Continued) 5.8.1 Qttiattlon.of the Event (Cor'tinued)
The fastost blowdown, and thorofore, the most rapid reactMty addition, occurs when the brook is at a stoam generator nozzlo. This break location is assumed for the casos analyrod. '
i Both full power and no-load (hot standby) Inillal condition casos woro considered for tvvo-loop operation (i.e., four reactor coolant pumps),
l l
Since the stoam gonorators are designed to withstand toactor coolant system '
oporWing pressure on the tubo sido with atmosphoric pressure on tho shell side, the continued intogrity of the reactor coolant 9ystem barrior is assured.
The most probable trip signals resulting from an MSLB includo low steam generator proscuro, high power, low steam generator water lovel, TM/LP, and high rate-of-change of power (for the no-load case),
5.8.2 Analysis criterin The steam line break accidont event is classiflod as a postulated accident for which the sito boundary dosos must be within the 10 CFR 100 critoria, Acceptablo site boundary dosos are demonstrated by showing that tne critical heat flux la not excoodod. 1 583 Ot1]Qctives of the Analygla The obloctivos of the analycle are to demonstrato that the margins to DNB for the reload coro no-load two-loop and full-load two-loop main steam line break casos are greator than that for the Cycle 1 casos glvon in the original FSAR. This is accompilshed by demonstrating that the return to power during the event for the reload core is loss than the retum to power calculated for Cycle 1, 5.8.4 Egy Parameter and Analysis Assumotions The MSLB accident is assumod to start from steady stato conditions with the inillal power boing 1530 MWt (102%) for the full power caso and 1 MWt for the no-load caso, The reactor coolant system cooldown causes the greatest positivo reactMty insortion into the coro when the Moderator Temperaturo Coefficient (MTC) is the most negativo. For this reason the Technical Specification negative MTC llmit conosponding to the end-of-cycle OPPD-N A-8303-NP, Rov. 03 41 of 125
S.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.8 ldala.Eleam Une B tgak Accident (Continued) 58.4 hev Parameter and Analytidtsumotions (Continued) is assumod in the analysis. Sinco the reactMty change associated with moderator feedback varios significantly over the temperature range covered in the analysis, a curve of reactMty insortion versus temperature rather than a single value of MTC is assumed. This curve la ttorived on the basis that upon reactor trip the most reactive CEA is stuck in the fully withdrawn position thus yloiding the most acNorso combination of scram wonh and reactMty insortion.
Although no single value of MTC is assumed in the analysis, the moderator cooldown reactMty function Is calculated assuming an Inillal MTC equal to the most negative Technical Specification limit.
ReactMty feedback effects from the variation of fuel temperature (i.e., Dopplor) are included in the analysis. The most negative Doppler defect function, v. tion used in conjunction with the decreasing fuel temperature causes the greatest positive reactMty insortion during the MSLB ovent. In addition to assuming the most negative Doppler defect function, an additional 15% uncertainty is assumed, i.e., a 1,15 multipilor. This multiplier conservatively increases the subcritical multiplication and results in a larger retum-to-power.
The delayed neutron procursor fraction, b, assumed is the maximum absolute value including uncertainties for end of cycle conditions. This is conservativo since it also maximizes subcritical multiplication and thus, enhances the potential for a return-to-power.
The steam gonorator low pressure trip, which occurs at 478 psia (including a 22 psla uncertainty below the nominal trip sotting of 500 psla), is the trip assumed in the analysis. No credit is taken for the high power trip which occurs at approximately the same timo for the full power caso, For the casos analyzed, it is assumed that the most reactive CEA is stuck in the fully withdrawn position.11 all CEAs insert (no stuck CEAs), there is no retum-to-criticat and no power transient following trip.
The cold edge temperatures are used to calculato modorator reactMty insortion during the cooldown, thus maximizing the retum-to-critical and retum-to-power potentials.
The Emergency Operating Procedures incorporate the Trip 2/ Leave 2 RCP operating strategy as indicated in Reforonce 5-8. For a steam line broal:,
OPPD-N A-8303-NP, Rev. 03 42 of 125
5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) l 5.8 - Main staarnitne Break. Accident (Continued) 5.8.4 My Parameter and Analvain Ammimotions (Continued) t the Trip 2/ Leave 2 strategy will result in tripping two RCPs (at 1350 pslo), if tho i I
event was misdiagnosed as a LOCA, all four RCPs would be tripped. Ac discussed below, for a main steam line break the consequences of Trip 2/ Leave 2 is bounded by the loss of offsite power and the loss of offsite power case is bounded by tripping no RCPs. Consequently, the limiting main steam lins break accident occurs with all RCPs operating. l The MSLB case with the RCPs tripped 18 similar to the MSLB case with a loss of offsite power since the RCPs coastdown in both events. As discussed in Reference 6-4, the loss of offsite power delays safety injection due to the time delay for the emergency diesel generators to restoro power to the safety ,
injection pumps and causes a coastdown of the RCPs.
The coastdown affects the degree of overcooling and increases the time for safety injection borated water to reach the core midplane, Becauso manual -
tilpping of the RCPs resuits in a later coastdown of the RCPs and because ,
cafety injection is not dolayed since offsite power is available (i.e., the diesol generator startup and pump loading delays are not present), the injected boron
!- will arrive at the core midplane sooner for a MSLB with the RCPs tripped than for a MSLB with a loss of offsite power. Therefore, the reactMty effects of a MSLB with the RCPs tripped are less severe than for the MSLB with a loss of offsite power.
Reference 5-4 states that the MSLB case with a loss of offsite power results in the injected boron being dominant over the RCS cooldown and concludes that the reactMty effects of a MSLB accident would be reduced in soverity with a concurrent loss of offsite power when compared to the same event with offsite power available and the RCPs operating. Because the reactMty effects of a MSLB with the RCPs tripped after SiAS are less severe than a MSLB with a concurrent loss of offsite power, it is concluded that the teactMty effects for tho -l MSLB case with the RCPs tripped utilizing Trip 2/ Leave 2 at 1350 psla are less severe than for a MSLB with offstte power available and RCPs operating; The reactor coolant volumetric flow rato is assumed to be constant during the
' incident.- The LCO flow rate (196,000 gpm) was used in order to obtain the i g
most advorse results.
l OPPD-N A-8303-NP, Rev. 03 43 of 125
. , _ . . --,..._ _ _.-- _ __.__ , _- ~ -- - -, _ _ _ .__ _ . _ . -
5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.8 Mn!n StennLine Break Acclgqu!(Continued) 5.8.4 KeylatAInttcLand. Analysis Astumollons (Continued)
A lower flow rato increases tho initial fuel and avnrago primary coolant temperatures and consequently results in a higher steam gonorator pressuro and a greater steam gonorator mass inventory.1 hose effects causo a longer blowdown, a greater blowdown rato and a greator docreaso in overago primary coolant temperature After MSfV closure the lower flow rato docreases the rate of roverse heat transfer from the intact steam generator, thoroby increasing tho heat extracted from the primary steam by the ruptured steam generator. The overall offect is that the potential for a rotum-to-power is inaximized.
Maximum values for the hoat transfor coefficient across the steam gonorator are used for the no-load inillat condition caso, whilo hominal values are used for the fultload inillal condition. Those heat transfor coefficients result in tho most sovero conditions during the incident because of tho shapo of the reactivity versus modorator temporature function and the difference in avorogo modorator temperature for the maximum and minimum values of the stoam generator host transfor coefficients.
The fast cooldown f.ollowing a MSLD results in a rapid shrinking of the reactor coolant After the pressurizer is emptiod, the reactor coolant pressure is assumod to bo equal to the saturation pressuro corresponding to the highest
- temperaturo in the system, Safoty injection actuation occurs at 1578 psia (i.e.,1000 pt la minus the 22 psla uncertainty) after the pressurizer emptios. Additional time is tequired for pump accolorat!an, valve opening, and flushing of the unborated part of the cafoty injection piping along with the requiremont that the ACS pressure decreaso below the shutoff head of the safety injection pumps (1376 psia for high pressure safoty injection (HPSI) pumps and 201 pela for low pressuro safety injection pumps (LPSI) pumps). The analysis takes credit for one HPSI pump, ono LPSI pump, and the r>afoty injection tanks.
The boric acid is assumed to m!x homogeneously with the reactor coolant at the points of injection into the cold logs. Slug flow is assumed for movement of the mixture throu0h the piping, plena, and core. Attor the boron reachos the coro midplano, the concentration within the core is assumed to increase as a stop function after each loop transit interval.
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i 1
5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.8 Main stemn Une Break Accident (Continued) 1 5.8.4 Kev Parameter and Analysis Assumotions (Continued) l l
The boron concentration of the safety injection water is assumed to be at the i Technical Specification minimum limit. The values of the inverse boron worth I are conservatively chosen to be large to minimize the negative reactivity !
Insertion from safety injection.
Since the rate of temperature reduction in the reactor coolant system increases with rupture size and with steam pressure at the point of rupture, it is assumed that a circumferential rupture of a 26-inch (Inside diameter) steam line occurs at the steam generator main steam line nozzle, with unrestricted blowdown.
Critical flow is assumed al the point of rupture, and all of the mass leaving the i break is assumed to be in the steam phase. This assumption results la the maximum heat removal from the reactor coolant per pound of secondary water, since the latent heat of vaporization is included in the net heat removal. A single failure of the reverse flow check valve in the ruptured steam generator is assumed; so that the intact steam generator will have steam flow through the unaffected steam line and back through and out the ruptured line. Based on sensitivity analyses performed by OPPD, this is the most severe single failure g for the steam line break event. The analysis credits a choke which is Installed in each steam line immediately above the steam generator and assumes the steam flow from the intact steam generator is through a 50% area reduction >
choke installed in a 24 inch steam line. This flow will be terminated upon MSIV closure.
The feedwater flow at the start of the MSLB corresponds to the initial steady state operation. For the full load inillal condition, it is automatically reduced in accordance with the program used in the valve controller. For the no load .
Initial condition, feedwater flow is assumed to match energy input by the reactor coolant pumps and the 1 MWt core power, Feedwater isolation upon the receipt of a low steam generator pressure (at 478 pala) is credited for both the full load and no load cases. A valve closure time of 30 seconds was used.
5.8.5 - Anatvals Method The analysis of the main steam line break accident is performed using CESEC which models neutron kinetics with fuel and moderator temperature feedback, the reactor protactive system, the reactor coolant system, the steam generators and the main steam and feedwater systems.
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5.0 R TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.8 Malajteam Une Break Accident (Continued) 5.8.6 Analvsls Results and 10 CFR 50 50 Critoria l The results of the analysis for the Fort Calhoun steam line break event are discussed in Section 14.12 of the 1983 update of tne Fort Calhoun Station Unit j
No.1 USAR. The critoria of 10 CFR 50.59 are met if the calculated {
retum-to-power is less than the retum-to-power reported for the Cycle 1 analysis, using the current Technical Specification limit on shutdown margin and moderator temperature coefficient.
5.8.7 Conservatism of Rggults Conservatism is added to the analysis by inclusion of uncertainties in moderator and fuel temperature coefficients of reactMty, by saking no credit for vold reactivity feedback, by taking credit for only 1 HPSI pump, by assuming all RCPs operate instead of manually tripping two pumps and by taking no credit for the stuck CEA worth.
5.9 Selzgd Rotor Accident 5.0,1 Definition of Event The seized rotor accident is assumed to be caused by a mechanical failure of a single reactor coolant pump. It is assumed that the rotor shears instantaneously, leaving a low inertia impeller attached to a bent shaft. This latter combination comes to a halt immediately causing a sharp drop in the
~ flow rate. The rapid reduction in core flow will initiate a reactor trip on low flow within the first few seconds of the transient.
5.9.2 - Analvsin Criteria A s' ingle reactor coolant pump shaft seizure is classified as a postulated accident for which the dose rates must be within 10 CFR 100 guidelines.
5.9.3 Oblective of the Analvsla The objective of the analysis is to demonstrate that the radiological releates are within a small fraction of 10 CFR 100 guidelines. This objective is met 11 It can be shown that less than 1% of the pins fall durin0 the event.
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e ro 5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued)
. 5.9 Selzed Rotor Accident (Continued) 5.9.4 Kev Parameters and Analvsls Assumotlorg The key parameterc used in the analysis of the seized rotor event are given in Table 5.9.4-1. The seized rotor is conservatively assumed to result in a 0.1 second rampdown of the core flow from its initial value to the 3 pump value.
For CETOP calculations, [
1 5.9.5 Analvsis Method Two methoas of analyzing the seized rotor event are discussed in this section.
Section 5.9.5.1 discusses a method which does not require transient analysis input. Section 5.9.5.2 discusses a method which utilizes transient analysis input.
l 5.9.5.1 Analvsis Method Without Transient Analvsis Resoonse Inout This method calculates the number of pin failures assuming that the core flow instantaneously decreases to the 3-pump flow rate.
This method utilizes the TORC analysis with a 3-pump inlet flow distribution. The initial RCS pressure and core inlet temperature !
are used as input to TORC and the core average heat flux is conservatively assumed to remain at its initial value. The mcximum value of frTis combined with a conservatively flat power distribution. The TORC calculation [
], the number of pins that have failed is calculated.
5.9.5.2 Analvsis Methods Usina Transient Analvsis This method utuzes the CESEC code to calculate the transient I response for the seized rotor event. The CETOP code is then used to determine the time of minimum DNBR. The TORC code utilizes the 3-pump inlet flow distribution,3-pump core flow rate, and the RCS pressure, core inlet temperature and core heat flux calculated at the time of OPPD-NA-8303-NP, Rev. 03 l l
47 of 125
i 5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued)
Table 5.9,4-1 KEY PARAMETERS ASSUMED IN THE SElZED ROTOR ANALYSIS Parnmeter 110113 Valug initial Core Power MWt $530*
initial Coro inlet *F Maximum allowed
- Temperature by Toch. Specs.
Initial Reactor Coolant psia Minimum allowed
- System Pressurb by Tech. Specs.
Moderator Temperature x10a p/'F Most negative allowed Coefficient by Tech. Specs.
Fuel Temperature x10 'ap/'F Most negative p.edicted Coefficient during core life.
Core Aserage gH , BTU /hr-ft2.'F Maximum value predicted during core life, initlai Core Mass x10albm/hr Best estimate flow
- Flow Rate CEA Drop Time sec Maximum value allowed by Tech. Specs..
Scram Reactivity Worth %ap Minimum predicted during core life.
- For DNBR calculations, offects of uncertalntles on theco parameters were combined statistically.
l OPPD-N A-8303-NP, Rev. 03 48 of 125
l 1
5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.9 Setzed Roto (Accident (Continued 5.9.5.2 Analysis Methods Usino Transient Analysis (Continued) minimum DNBR by CESEC. The steps to determ!ne the number of pin failures is then performed ( ) as discussed in Section 5.9.5.1, 5.9.6. Analvsis Results and 10 CFR 50 59 Criteria The results of the seized rotor analysis are contained in Section 14.6.2 of the Fort Calhoun Station Unit No.1 US AR. The criteria of 10 CFR 50.59 are met, if the number of pin failures is less than one percent.
5.9.7 Conservatism of Results Conservatism in the calculated number of fuel pins predicted to experience DNBR is added through the use of the following assumptions:
A. The most positive MTC is assumed in the analysis. The actual MTC is more negative and would limit core power and heat flux rise.
B. A relatively flat pin census is assumed in the analysis. A more peaked pn census distribution would lower the number of pins pred;cted to experience DNB.
C. For the case without transient analysis, no credit is taken for the pressure increase during the transient and calculating the minimum transient DNBR.
5.10 CEA Election Accident 5.10.1 Definition of Event A CEA election accident is defined as a mechanical failure of a control rod mechanical pressure housing such that the coolant system pressure would eject the CEA and the drive shaft to a fully withdrawn position. The consequences of this mechanical failure is a rapid reactivity insertion which when combined with an adverse core power distribution potentially leads to localized fuel damage. The CEA ejection accident is the most rapid reactivity insertion that can be reasonably postulated. The resultant core and thar.c!
power excursion is limited primarily by the Doppler reactivity effect of the increased fuel tempcratures and is terminated by reactor trip of the remaining CEAs activated by the high power trip or variable high power trip.
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5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5,10 CEA Election Accident (Continued) 5.10.2 Analysis Criteria The CEA ejection event is classified as a postulated accident. The design and limitiri criteria are:
A. Fuel cladding and enthalpy trLUsT)lds (Reference 5-5) are:
r Clad Damage Threshold
- Average Pellet Enthalpy (at hot spot) A 200 cal / gram g Contorline Melting Threshold
' Total Centerline Enthalpy 4 250 cailgram Fully Molten Centerline Threshold Total Centerline Enthalpy A 310 cal / gram 8.- The peak reactor prCsure during a ponlon of the transient will be less than the value that will cause stress to exceed the emergency conditions stress limits as defined in Section 3 of the ASME Boller and Pressure Vessel Code. This objection is achieved if the peak RCS piassure does not exceed 7750 psla.
C. Fuel molting will be limited to keep the offsite dose consequences well within the guidelines of 10 CFR 100..
5.10.3 Qb!ectives of the Analyalg P
- The objective of the analyshis to demonstrate that fuel failures are less vian ttuse reported in Section 14.3.41 of the Fort OalMun Station Unit No.1 USAR of that site boundary doses are within the 10 CFR 10C limhs.
5.10.4 Analvsls Method OPPD utilizes the CEA Election Accident Analysis of our current fuel vendor,_ .
Westinghouse. This analysis methodology is documented in Reference 5 and is perfo~ned by Westinghouse. This methodology utilizes physics parameters, calculated by OPPD in accordance with the methods outlined in Reference 5-6.
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1 S.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.10- CEA Election Accident (Continued) 5.10.5 Analysis Results and 10 CFR 50.59 CI[tedL The results of the CEA Ejection Analysis are reported in Section 14.13 of the Fort Calnoun Station Unit No.1 USAR Criteria of 10 CFR 50.59 are satisfied if fuel failures are less than those assumed for input to the Radiological Consequences portion of the analysis.
5.10.6 Conservatism of Results The major area of conservatism is the calculation method used to obtain the ejected CEA worth and the ejected radial peak. The ejected worth and the ejected radial peak are calculated without any credit for Doppler of Xenon feedback, in addition, the hot full power ejected worth and elected peak are calculated assuming the no-load temperature of 532*F. The lower temperature is more adverse since this causes a power roll to the core periphery which also happens to be the location of the ejected CEA. Also, the ejected worth is calculated assuming the CEAs are fully inserted for hot full power case regardless of PDIL Thus, the ejected worth is conservative.
5.11 Loss of Coolant Accident OPPD does not perform the Loss of Coolant Accident Analysis. The large and small break loss of coolant analyses were perfoimed by Westinghouse. The large break and small break topicals are mentioned in Reference 5-7. OPPD verifles that the physics input assumptions and the maximum Tod bumup are within the bounds assumed in the W large break analysts, g 5.12 Loss of Load to Both Steam Generators Event 5.12.1 Definition of the Event A tMal loss of load to both steam generators usually results from a turbine trip due to a loss of extemal electrical load or to abnormat variations in electrical network frequencies. Other possible causes include the simultaneous closure of all turbine stop valves or main steam isolation valves. All initiating mechanisms result in a corresponding reduction in heat removal from the reactor coolant system due to the loss of secondary steam flow. Although a Reactor Protective System trip signal would normally result from a turbine trip, no credit is taken in the analysis of this event for the turbine trip signal.
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5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.12 Lags of Load to Both Steam Generatorgfygll(Continued) 5.12.2 Analvsts Criteria The loss of load to both steam generators event is classified as an Anticipated Operational Occurrence (AOO) for which the following critoria must be met:
A. The peak RCS pressure does not exceed 2750 psia (110% of design pressure).
B. The transient minimum DNBR is greater than the 95/95 confidence Interval limit for the CE-1 correction limit.
C. The Peak Linear Heat Generation Rate (PLHGR) does not exceed 22 kw/ft.
Criteria B. and C. are not of major concem since DNBR increases during the event and the PLHGR margin required is much less limiting than other AOOs.
Therefore, criterion A. is the main concem in analyzing this event. The loss of load to both steam generators event is the limiting AOO event with respect to peak RCS pressure.
5.12.3 Objectives of the Analysis The objective of the analysis is to demonstrate, for modifications to the plant which potentially degrade RCS heat removal capability (including steam generator plugging) that the peak RCS pressure stays within 110% of the design pressure in accordance with Section 111 of the ASME Pressure Vessel Code. This objective is achieved if the peak RCS pressure does not exceed 2750 psia.
5.12.4 Kev Paramqtgrs and Analysis Assumotions The key parameters used in the loss of load (to both steam generators) event are given in Table 5.12.4-1. Assumptions used in the analysis to maximize heat up of the RCS and consequently peak RCS pressure include:
A. The event is initiated by a sudden closure of the turbine stop valves without a simultaneous reactor trip.
B. No credit is taken for operation of the PORVs, pressurizer sprays, and the turbine steam dump and bypass system, i.e., the pressurizer pressure control system is assumed to be in MANUAL.
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5.0 _ TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued)
Table 5.12.4-1 ,
KEY PARAMETERS ASSUMED IN THE LOSS OF LOAD TO BOTH STEAM GENERATORS ANALYSIS Parameter 1)ntts Valua initial Core Power MWt 1530 101t1a1 Core inlet - *F Maximum allowed Temperature by Tech. Specs, initial RCS Pressure psia Minimum allowed by Tech, Gpecs initial Steam Generator psia Minimum value corresponding to core l Pressure . Inlet temperature operating range, Initlai Core Mass x10s Ibm /hr Minimum allowed Flow Rate . by Tech. Specs.
Moderator Temperature x10 'Ap/*F Most positive Coefficient allowed by Tech. Specs, Fuel Temperature x10 'Ap/*F Least negative !
Coefficient predicted during core life, !
Fuel Temperature Coefficient Multiplier 0.85 CEA Drop Time sec Maximum allowed by Tech Specs.
-Scram Reactivity Worth - %ap Minimum predicted during core lifetime
- Scram Reactivity - Consistent with most Curve : positive axial shape (bottom peaked) {
allowed by Tech, Specs, Core Average Hgap ' BTU /hr-ft2 *F _ Maximum predicted during core lifetime.
Kinetics Parameters EOC parameters (minimum absolute p).
RPS Response Time sec 1,4 OPPD-NA-8303-NP, Rev. 03 53 of 125
1 l
5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.12 Loss of Load to Both Steam GeogIntors Event (Continued) 5.12.4 Kev Parameters and Analysis Assumotions (Continued)
C. The rod block system is assumed to prevent too motion (other than scram) during the transient.
D. Maximum charging flow and zero letdown flow are assumed.
E. Termination of the event occurs as a result of a high pressurizer pressure trip.
5.12.5 Analysis Method The analysis methods utilized by OPPD to analyze the loss of load to both g steam generators event consists of simulating the event using the CESEC computer code, utilizing the analysis assumptions listed in Section 5.12.4 (above) as input, and extract!ng the peak RCS pressure for comparison with the 2750 psla upper limit.
5.12.6 Analvsis Results and 10 CFR 50.59 Criteria The results of the loss of load to both steam generators are contained in the Fort Calhoun Station Unit No,1 USAR, The critoria of 10 CFR 50.59 are met, if the peak RCS pressure is less than 110% of design pressure in accordance with Section ill of the ASME Pressure Vessel Code. This objective is achieved if the RCS pressure does not exceed 2750 psia.
5.12.7 Conservatism of Results The following areas of conservatism are included in the analysis to obtain a conservatively high peak RCS pressure:
A. Field measurements demonctrate that the C.EA magnetic clutch decay time is less than that assumed in the analysis.
B. The actual scram worths are greater than those assumed in the analysis.
C. The actual MTC is more negative during power operation than assumed in the analysis.
D. The steam dump and bypass system and the pressurizer pressure control system (PORVs and sprays) are operated in the AUTO mode rather than MANUAL as assumed in the analysis.
l l
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.5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.12 Loss of Load to Both Steam GQQQIntors Event (Continued) 5.12,7 Conservatism of Resg[t3 (Continued)
E. Actual secondary pressure is highor which results in oather secondary safety valve opening and oarilot alleviation of the primary system temperature and pressure rises.
F. The maximum pressurizer safety valvo capacitlos are assumed to bo 90% of the ASME rated values.
1 G. A one percent pressure uncertainty is appiled to the primary and secondary safety valvo sotpoints,1.0., a 1.01 multiplier.
5.13 Loss of Feedwater Flow Event
- 5. t 3.1 Definition of Event A total loss of main foodwater flow event is defined as a loss of foodwater flow when operating at power without a correspor. ding reduction in steam flow from the steam 00norators. The most likely causes for this event are the loss of all foodwater or condonsato pumps or the inadvertent closure of olther the main feodwater regulating valves or the foodwater isolation valvos due to a foodwater controller malfunction or manual positioning by the oporator. The recult of this mismatch in which turbino domand remains at 100%, is a reduction of the steam gonorator liquid inventories and a degrading RCS heat removal capability. As the heat removal capability is lost, throu0h decreasing steam generator inventories (i.e., levels) the RCS temperatures and pressuro increase. Normally the event would be terminated by a reactor trip on low steam generator level. Sinco no credit 18 taken in the analysis for the steam generator low level trip, a high pressurizer trip eventually results.
Automatic actuation of the auxillary foodwater (AFW) system will also eventually occur (after reactor trip) if either main foodwater is not rostored or manual actuation of the AFW system is not performed by the operator. The
' AFW system actuation ensures the maintenance of a secondary heat sink.
5.13.2 Analvsis Criteria The loss of foodwater flow event is classified as an Anticipated Operational Occurrence (AOO) for which the following criteria must be mot:
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- 5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.13 Loss of Feedwater Flow Event-(Continued) ,
- 5.13.2 Analysis Criteria (Continued)
- A. The peak RCS pressure does not exceed 2750 psia (110% of design pressure).
B. The transient minimum DNBR is greater than the 95/95 confidence interval limit for the CE-1 correlation limit <
C. - The Peak Linear Heat Generation Rate (PLHGR) does not exceed 22 ,
kw/ft.
Criterla B and C. are not of major concem because DNBR does not decrease.
below the initial steady state value and the PLHGR margin required is much -
less limiting than other AOOc. Therefore, only criterion A. requires reevaluation should plant modiflCallons (such 89 steam generator tube plugging) be made which result in degraded secondary heat transfer capability beyond that of this .i event. For Fort Calhoun Station, this event is bounded by the loss of load incident.
5.13.3 ' Objectives of the Analysis
- The objective of this analysis is to demonstrate, for plant modifications which potentially degrade RCS heat removal capability (including steam generator tube plugging), that the peak _RCS pressure stays withl0110% of the design pressure in accordance with Section lit of the ASME Pressure Vessel Code This objective _is achieved if the peak RCS pressure does not exceed 2750
. psla.
5.13,4 Kev Parameters and Analvsis Assumotlons The key parameters used in the loss of feedwater flow event are given in Table :
5.13.4-1. Assumptions in the analysis to maximize heat up of the RCS and consequently the peak RCS pressure include:
t.')j 1 A. . The event is initiated by an instantaneous loss of main feedwater No credit is taken for the low steam generator level trip.
B. The steam dump and bypass system is assumed to be in MANUAL i (l.e., inoperative).
C. The pressurizer pressure control system is in MANUAL (i.e., PORVs and sprays are Inoperable).
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5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.13 Loss of Feedw.ater Flow Everd (Continued) 5.13.4 Key Parameters and Analysis Assumotions (Continued)
D. The pressurizor level control system is in MANUAL with maximum charging end zero letdown flows.
E. The rod block system is assumed to prevent rod motion (other than scram) during the transient.
5.13.5 Analysis Method l
The analysis methods used by OPPD to analyze a loss of main foodwater flow g event consists of using the CESEC computer code to simulato the event, utilizing the analysis assumption listed in Section 5.13.4 (above) as input, and extracting the peak RCS pressure for comparison with the 2750 psia upper limit.
5.13.6 /snalvsis Results and 10 CFR 50.59 Criterla The results of the loss of feedwater flow are contained in the Fort Calhoun Station Unit No.1 USAR. The criteria of 10 CFR 50.59 are met, if the peak RCS pressure is less than 110% of the design pressure in accordanco with Section lli of the ASME Pressure Vessel Code. This objective is achieved if the peak RCS pressure does not exceed 2750 psia.
5.13.7 Conservatisms of Results A. Field measurements demonstrate that the CEA magnetic clutch decay timo is loss than that ascurned in the analysis.
B. The actual scram worths are greator than those assumed in the analysis.
C. Tho actual MTC is more negative during power oporation than assumed in the artalysis.
D. The steam dump and bypass system and the pressurizer pressuro control system (PORVs and sprays) are operated in the AUTO mode rather than the MANUAL mode as assumed in the analysis.
E. Actual e9condary pressure is higher which results in eartior secondary safety valve opening and earlier alleviation of the primary system temperature and pressure rises.
F. No credit is taken for a steam generator low level trip.
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5.0 = TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued)
Table 5.13.4-1 KEY PARAMETERS ASSUMED IN THE LOSS OF FEEDWATER FLOW ANALYSIS Parametcr 1,lnits Valug inillal Core Power MWt 1530 !
Initial Coro Inlet *F Maximum allowod Toreporature by Toch Spocs.
Irh lal RCS Pressure . psia Minimum allowed by Toch. Spocs initi.nl Steam Gonorator ' psla Minimum value 5'icssure corresponding to core inlet temperature operating range.
initial Cero Mass x108 lbm/hr Minimum allowed Flow Rate .by Toch. Specs, Moderator Temperaturo x10*' A p/'F Most positivo Coefficient - allowed by Toch. Specs.
Fuel Temperature- x10 *ap/*F Least negativo Coefficlont predicted during core life.
Fuel Temperaturo -
Coefficient Multiplier 0.85 CEA Drop Time sec Maximum allowed by Tech. Specs.
t Scram Reactivity Worth %op- Minimum predicted during core lifetime Scram Reactivity Consistent with most Curve. positivo axial shapo (bottom peaked) ,
allowod by Toch. Spocs.
Coro Average Hgap -
- BTU /hr-ft2. F Maximum predicted during coro lifetime
. Kinetics Parameters EOC paramotors (minimum absoluto p).
= RPS Responso Timo . sec 1.4 2
OPPD-NA-8303-NP. Rov. 03 58 of 125
6.0 TRANSIENT ANALYSIS CODE VERIFICATION 0.1 Introduction OPPO currently uses the CESEC-Ill computer code to calculate the transient responso of the NSSS during events discussed in thls document. OPPD has also benchmarked and justified the use of the attemate transient thermal hydraulic code CENTS. The CENTS code has advantages over CESEC-lit in both modalling and flexibility of use.
CENTS is a state-of-the-art code that can be used for best estimato, as well as, safety analyses. This allows CENTS to be used for evaluating safety and operability issues, as well as for reload safety analyses. Another incontive for replacing CESEC-Ill is that CENTS can model the as-built facility at Fort Calhoun morb accurately than CESEC-lit.
An example of this increased ability is shown in the main steam safety valve models. l CESEC-lit requires a constant cross sectional flow area for the valve discharge, which I means that the power operated safety valves can not be modelled as they have a l
smaller cross sectional area than the other main steam safeties CENTS is more flexible and all five of the main steam safety valves can be accurately modelled.
Combustion Engineering (CE) also plans to replace CESEC-lit with CENTS for performing reload transient analyses. Once CENTS has been licensod, CE will provido only limited support for CESEC-Ill.
Tne verification of both CESEC-Ill and CENTS are documented in this chapter sinco CESEC-Ill is currently licensed as a safety analysis codo, and will continue to be used f
for licensing purposes until CENTS is licensed for safety related applications.
Combustion Engineering has provided overall verification of the CESEC-Ill codo in Referencoe 6-1 and 6-2, and CENTS in References 6-3 and 6-4. The purpose of the work documented in this chapter is to demonstrato OPPD's ability to correctly utilizo both the CESEC-Ill and CENTS codes, in order to demonstrate Omaha Public Power District's ability to correctly use both computer codes, verification work has been performed by benchmarking against both actual plant transient data and independent safety analyses previously accepted by the NRC. The plant transients which were benchmarked were the Turbine-Roactor trip and Four-Pump Loss of Coolant Flow events. The independent safety analyses which were benchmarked using CESEC-Ill were the Dropped CEA, Main Steamline Break, and RCS Depressurization events. Those transients bonchmarked using CENTS were the Dropped CEA and RCS Depressurization events. These are two of the events that are typically analyzed on a cycle specific basis. The Main Steamline Break was not benchmarked using CENTS since the three-dimensional core neutronics model for OPPD-N A-8303-NP, Rov. 03 ,
59 of 125
0.0 TRANSIENT ANALYSIS CODE VERIFICATION (Continued) a 6.1 - IntroductlOD (Continued)
CENTS was not available during the benchmark effort. Since the MSLB event is not analyzed each cycle, CENTS will not be used in a licensing application for MSLB until a future benchmark is perfonned. Comparisons for the Dropped CEA and RCS Depressurization events are addressed below.
6.2 Comp.arison to Plant Data
- A prerequisite for beginning performance of transient analyses is verification that the code will stabilize with the correct system parameters when simulating steady state
- operation. This step was performed for both the CESEC-Ill and CENTS codes and correct results were obtained.
For plant transient benchmarking, the type of translents that have occurred and both the quality and quantity of data existing for each is very limited, in nearly all cases, operators take actions which reduce the consequences of the event, Introducing '
- complicated perturbations in system response which cannot be easily modeled, because the actions taken and the time at which they are performed are not always
- recorded. Strip chart recordings on an extremely compressed time scale are generally
. the only form of data available. This compressed time scale (with graduations typically of 10 minutes) does not permit adequate comparisons to CESEC-Ill or CENTS g
modellr g in which seconds are of major concem. Cyc!e 1 startup testing is the only
. source of plant transient data in which syst9m parameters were measured with high .
speed strip chart recorders and no operator action takert Good data exists for a -
nominal full power turbine-reactor trip and a 35% power total loss of RCS flow event.-
The CESEC-Ill and CENTS computer codes were both set up to model Cycle 1 in a - g best estimate mode to permit accurate comparisons to the actual measured plant responses for both of the above cases. A summary of each of these comparisons
- follows.-
6.2.1 Turbine-Reactor Trio For the turbine-reactor trip case, the plant comparison data were obtained-from the Cycle 1 startup testing performed May 10,1974. The event was -
initiatea from 97% of full power, all-rods-out, and equilibrium xenon. 'The plant response data used in the comparisons were obtained from vendor OPPD-NA-8303-NP, Rev. 03 60 of 125
_ _m__.___ _ . . _ _ _ _ _ _ _ , _ _
. . _ . , _ . qt_.._ ,, ,
6.0 TRANSIENT ANALYSIS CODE VERIFICATION (Continued) 6.2 Comoarlson to Plant Data (Continued) test recorders. No operator action was taken following the manual generator-turbine trip (whlen provided the RPS " loss of load" trip). Prior to the trip the main feodwater, the pressurizer pressure, and pressurizer level control systems were all in the automatic modo, and the letdown backpressure controi valve was in the manual mode. With the exception of adjusting the letdown backpressure control valve at 20 seconds, no l operator action was takeq for 60 seconds following the trip. Figures 6-1 through 6-9 show plots of the comparisons between the measured plant responses and the predicted CESEC-Ill and CENTS responses. It should i be noted that this test was performed based on a rated power level of 1420 MWt rather than the current limit of 1500 MWt (the design power extension for which licensing was obtained in Cycle 6).
Figure 6-1 shows the nuclear power responf,e following the turbine-reactor trip. The CESEC-Ill and CENTS predictions follow the same power decay rate. However, the endpoint residual power for both code predictions is slightly higher, which is conservative, it should be noted that trip delays locluded in the CESEC-Ill modeling provant the immediate power drop observed in the plant data; again this is conservative. Trip delays can also be added to CENTS, but were not for this case since it was intended as a best estimate analysis. The pressurizer pressure response predicted by both codes are shown in Figure 6-2. The figure shows very Good agreement of the codes' predictions with the plant response. The CESEC-Ill case was initiated 10 psla above the plant data and remained slightly above the plant response for the duration of the transient. The difference between the predicted and measured pressurizer pressures increased slightly due to the higher residual power after trip as shown in Figure 6-1. The CENTS case was initiated about 12 psia below the plant data and followed the plant response closely after the first 10 seconds.
This difference between pressurizer pressure predicted by CESEC-Ill and measured pressurizer pressures at 60 seconds is only 19 pela, a value which is less than the pressure measurement uncertainty. The difference between the CENTS prediction and the actual pressurizer pressure at 60 seconds is less than 3 psla. Figures 6-3 and 6-4 show the RCS cold-leg and hot-leg temperature responses, respectively, for each steam generator loop for the plant data and the CESEC-Ill and CENTS predicted average OPPD-NA-8303-NP, Rev. 03 61 of 125
6.0 TRANSIENT ANALYSIS CODE VERIFICATION (Continued) 6.2 Comoarlson to Plant Data (Continued) cold leg and hot leg temperatures. The differences in the transient response of the two steam generator loops for the plant data are attributable to the differences in the main feodwater flow rate rampdown after trip (see Figure 6-7). Both the CESEC-Ill and CENTS predictions g lead the actualloop measurements because of the measurement delays associated with the response time of the resistance temperature devices (RTDs) providing the temperature signals. Figures 6-5 and 6-6 show the g measured and predicted steam generator pressure responses for each steam generator. These plots show good agreement of the predictions from CENTS and CESEC-Ill with the Cycle 1 test data with only minor differences. The pressure predicted by both CESEC-lil and CENTS is higher than the measured pressure early in the event due to a combination of the greater heat residual as shown in Figure 6-1, a quicker turbine stop valve closure, and quicker steam dump-bypass operation assumed in the CESEC-Ill and CENTS analyses, Figure 6-7 shows the feedwater flow during and after feedwater rampdown. The feodwater rampdown function is an input to both CESEC-lit and CENTS. Figures 6-8 and 6-9 show the steam flow from both steam generators dving the event. The CENTS predictions 'ollow the measured steam flow closely. The discrepancies between the CESEC-Ill and CENTS predictions and the measured data are due mainly to a quicker turbine stop valve closure and quicker steam dump and bypass valve operation assumed in the CESEC-Ill and CENTS analyses.
In conclusion, both the CESEC-Ill and CENTS predicted parameters for the g turbine-reactor trip show very good agreement with those measured in the Cycle 1 startup testing performed at nominal full power condit!ons.
6.2.2 Four-Pumo Loss of Coolant Flow For the four-pump loss of coolant flow case, the plant comparison data were obtained from the Cycle 1 startup test performed March 6,1974. This event was initiated from 35% power by manually and simultaneously tripping all four reactor coolant pumps. At the time of trip the pressurizer pressure, pressurizer level, main feedwater, and steam oump and bypass controllers were in the automatic mode. At approximately 20 seconds after I
OPPD-NA-8303-NP, Rev. 03 62 of 125
l 8.0 TRANSIENT ANALYSIS CODE VERIFICATION (Continued) 0.2 .CDmpadEQ1to Plant Data (Continued) tfsa eip, the oporators took manual control of foodwater in order to precludo overtooding of the steam generators and too rapid of a cooldown for the following natural circulation test.
The behavior of the various RCS and secondary paramotors that were measured and the CESEC-Ill prodletions for the first 30 seconds following the RCP trips are shown in Figures 6-10 through 6-17. These g comparisons show excellent agroomont. The minor differences that exist are discussed below. Currently, CENTS does not have the capability of being initializ0d at powers as low as 35% which would have boon necessary to simulate this event. This featuro is being added to the code and will be availablo in the next release. However, the RCS flow coastdown after all four pumps are tripped is rotatively independent of the power lovel. Thoroforo, a CENTS case was run to prodlet the RCS flow coastdown. The predicted flow is plotted in Figuro 6-10.
Figuro 6-10 shows a plot of the measured total RCS flow versus timo and that prodicted by the CESEC-Ill and CENTS codos. Both codos incorporate explicit modeling of the reactor coolant pumps. This data shows excellent agreement with the predicted flow while being slightly conservativo.
Figures 6-11 and 6-12 show the pressurizer pressure ar,d lovel responso comparisons which also show excellent agroomont. Figuro 6-13 shows plots of core nuclear power versus time. As in the turbino-reactor trip caso, CESEC-lit shows a slightly higher residual power after trip. The predicted and measured steam generator pressure responses as plotted in Figure 6-14, also show very good a0rooment. The response of the hot-leg and cold-log temperatures, as shown in Figure 6-15, is consistent with the data obtained from the turbine-reactor trip case. Again the delay associated with the RTD responso causes the predicted temperatures to load those that were measured. F10uro 6-16 shows that the main g foodwater input function used in CESEC-Ill was acceptabio in terms of the actual feedwater systom response it should be noted that the operator action of assuming manual control of the main foodwater system at approximately 20 seconds had little offect on any of the other system OPPD NA-8303-NP, Rev. 03 63 of 125
6,0l . TRANSIENT ANALYSIS CODE VERIFICATION (Continued) s 6.2 Compadson to PlanLQata (Continued) parameters examined, and that following a soveral second reduction in flow the previous flow rate was reestablished. Figure 6-17 shows that g turbino stop valvo closure rate assumed in the CESEC-Ill analysis was 4
qulckor than the actual valve responso.- The figuro also shows a steam flow rato mismatch between the two steam generators for the plant data.
This la something one would not expect and raises the question of the validity of the measurement or its uncertainty for Inis steam generator
' steam rate flow, because the two conosponding feedwater flow ratos (in Figure 6-16) are consistent.
-g in conclusion, the CESEC-Ill predictions for the 35% power total loss of coolant flow, as well as the CENTS prodleted RCS flow coastdown, show l Very good agrooment with the parametors measured during Cycle 1 startup testing.
6.3 . Comoarlsons Between OPPD Analvses and Indcoendent Analvses Previously Ecrformed by the Fuel Vendors for Verification of CESEC-Ill
^
Of the transients ana1yzed by OPPD for reload core licensing (using CE methodology) no plant data existed, so comparison of the limiting events to previous independent analyses performed by either Advanced Nuclear Fuels, formerly Exxon Nuclear '
Cori gy ("NC), or Combustion Engineering (CE) was done. Since Exxon Nuclear .
. Company performed some analysos in this section (used for comparison) prior to -
becoming ANF, all references to this company will be to ENC, For the comparison cases, the assumptions used in the analyses were similar to those used by OPPD, l.0,, g the core physics parametert. did not vary significantly between fuel cycles.- The events '
chosen for comparison were:
- (1) The Dropped CEA event is dependent upon the initlal available overpower 4 margin to prevent exceeding the SAFDLs. The goal of the analysis is to detennine the DNBR required overpower margin (ROPM),
(2) The Hot Zero Power (HZP) Main Steamlino Break which determinos the .
minimum required shutdown margin.
- (3) The Hot Full Power (HFP) Main Stoamline Break which determines tno most negative moderator temperature coefficient of reactivity allowed. y p
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6.0 TRANSIENT ANALYSIS CODE VERIFICATION (Continued) 6.3 Comoarisons Between OPPD Analyses and Indcoendent AnalyscSAcybusly Performed by the Fuel Vendors for Verification of CESEC-Ill (Continued)
(4) The RCS Depressurization event which is used in the determination of the
[ ). The [ ] accounts for DNBR margin degradation in the thermal margin / low pressure (TM/LP) trip [
] l l
6.3.1 D. topped _CEA I The Cycle 8 Dropped CEA analysis portonned by OPPD was compared to the previous analysis, contained in the Updated Safety Analysis Report (USAR). The USAR analysis was performed by ENC for Cycle 6. Table 6.3-1 summarizes the parameters and their values for Cycles 6 onJ 8.
Plots of core power versus time for the OPPD (Cycle 8) and ENC (Cycle 6) analyses are found in Figure 6-18. The curves show a very similar prompt g drop, to 69% versus 70%, respectively, and both cases show a return to a nominal 100% power. Both cases assumed that the turbine admlSSlon valves opened to their full open position In an attempt to maintain full load during tho event (l.o., the turbine control system was placed in the load set mode which is not used at Fort Calhoun Station). The core heat flux plots are contained in Figure 6-19. Both are very similar, as was the case in the core power cases. Figure 6-20 contains plots of the coolant average temperature versus time. Both figures are in good agreement showing a drop in average coolant temperature to 567 'F. Plots of the inlet and outlet temperatures for Cycle 8 are also included. Figure 6-21 shows plots of g the pressurizer pressure versus time. The minimum pressures predicted at 160 seconds are 1957 psla and 1945 psia for Cycle 8 and Cycle 6, respectively. This efference is small enough to be less than the pressure measurement uncertainty.
In summary, the primary system responses between the ENC and OPPD analyses show excellent agreement with each other which is consistent with reload coros having similar core physics parameters.
6.3.2 Hot Zero Power Ma!n Steamline Break The hot zero power (HZP) Main Steamline Break, which is the basis for determination of the required shutdown margin, wac analyzed by OPPD for OPPD-NA-8303-NP, Rev. 03 65 of 125
6.0 TRANSIENT ANALYSIS CODE VERIFICATION (Continued) 6.3 Comoarisons Between.QEPD Analys.es and indegendent Analyses Prevl0USly E0dQUnciby the Fuel Vendors for Verification of CESEC-Ill (Continued)
Cycle 6. The results of this analysis have boon compared to those of ENC in their Cyclo 6 analysis and to those obtained by CE in their Cyclo 6 control grade auxillary feedwat ar (AFW) system analysis. Table 6.3-2 shows comparisons of the pertinent input values for each of the analyses.
Figure 6-22 shows plots of core power for the Cyclo 8 OPPD analysis and g Cycle 6 ENC analysis, respectivuy. The maximum return-to-power is less for Cycle 8 than for Cycle 6 and occurs lator due to the use of a higher shutdown margin. The Cycle 6 CE AFW analys!s power is not includod because thero was no retum-to-critical and no return-to-power. Figure 6-23 shows plots of the coro average hoat flux for OPPD, ENC and CE, g respectively. Both the OPPD and CE analyses, which were portormed using CESEC Ill and CESEC-1, respectively, show a slight heat flux increase at approximately 12 seconds. This is due to subcritical multiplication. Otherwise, the heat flux curves within the specific analyses are essentially the same as the core power curves with a slight decay.
Figure 6-24 shows the total reactivity versus time for each of the analyses, g With very similar moderator cooldown curves, the peak reactivillos occur chronologically with increasing shutdown margin as expected; i.e., for increased shutdown margin (C"As) It takes longer to be offset by the positive moderator cooldown reactivity insertion.
Figure 6-25 shows plots of RCS pressure versus time for Cycle 8 (OPPD) and Cycle 6 AFW (CE). Also included in Figure 6-25 is the Cycle 1 (CE) results. All three of these curves show excellent agreement. The Cycle 6 AFW (CE) analysis shows a lower end point pressure than the Cycle 1 (CE) and Cycle 8 (OPPD) analyses due to the assumption of auxiliary feedwater addition. The ENC data availabio did not include the RCS pressure response.
Figuro 6-26 shows plots of the steam generator pressures for Cycle 8 g (OPPD) and Cycle 6 AFW (CE), respectively. These plots show reasonable agreement between pressures and times. The increase in the intact steam generator's pressuro is due to MSIV closure; l.o., failure of the reverse flow check valve on the intact steam generator was chosen as the OPPD-NA-8303-NP, Rev. 03 66 of 125
l l
'j
.6.0- - TRANSIENT ANALYSIS CODE VERIFICATION (Continued) .
6.3 - ' Comoarisons Between OPPD Anafvses and Indeoendent Analyses Previousjy Performed bv the Fuel Vendors for Verifiention of CESEC-Ill (Continued) ;
most adverse single failure. Following dryout of the ruptured steam generator, the pressure drops to atmospheric. The times of dryout r '
slightly different due to the increased normal water level value used In the I l
Cycle 8 analysis. 1 In summary, the HZP Main Steamline Break analysis for Cycle 8 shows trends similar to those in Cycle 6 as analyzed by both CE and ENC.
t 6.3.3 Hot Full Power Main Steamline Break ,
The hot full power (HFP) Main Steamline Break provides an acceptance -
criteria for the most negative moderator temperature coefficient (MTC) of reactivity, if a return-to-critical occurs, the goal of the reload analysis is to show that the retum-to-power is bounded by the most limiting case which, !
for the Fort Calhoun Station, is the Cycle 1 analysis. The Cycle 8 HFP
, analysis of this event was compared to the previous analyses performed -
-by ENC in Cycle 6 and by CE in their Cycle 6 control grade AFW system 4 analysis. Table 6.3-3 shows a comparison of the important input -
parameters for each of the analyses.-
Figures 6-27,6-28, and 6-29 show plots of core power, core average heat g ]
flux, and total reactivity for Cycle 8 (OPPD), Cycle 6 (ENC),~ and Cycle 6 AFW (CE).' Within each cycle's analysis, the coro average heat flux slightly lags the core power which peaks at a time several seconds after the peak reactivity is reached (for the retum-to-critical cases) The retum-to-power peaks occur at different times due to the different scram worths used, as
- explained for the shutdown margin in the HZP. Steamline Broak analysis section.
Figure 6-30 shows plots of the RCS pressure versus time for the Cycle 8, g
- Cycle 6 AFW, and Cycle 1 analyses, These plots are very similar and :
show excellent agreement. Figures 6-31 and 6-32 show plots of the RCS {
- temperatures for Cycle 8 and Cycle 6 AFW. Again good agreement exists -
..to approximately 180 seconds. At this time, the Cycle 6 AFW arulysis assumed runout flow from both AFW pumps to the ruptured steam x
OPPD-NA-8303-NP, Rev. 03 67 of .125 '
l . , J . -s ,_ ,_
6.0 TRANSIENT ANALYSIS CODE VERIFICATION (Cuntinued) 0.3 Comoarlsons Between OPPD.Jcalyses and independent Analyses Previously Performed _by the Fuel Vendofs for Verification of CESEC-Ill (Continued) generator which resumed the RCS cooldown. This additional cooldown I
caused by the AFW system is prevented from occurring in Cycle 8 by the logic of the newer safety grade AFW system.
Figure 6-33 shows plots of steam generator pressures versus time for g Cycle 8 and Cycle 6 AFW (CE). These results are very similar except that the intact steam generator pressure, in the CE analysis, begins to drop after 180 seconds due to the AFW Induced RCS cooldowT1.
6.3.4 RCS Deoressurization The RCS Depressurization analysis is performed to calculate a [
] for the TM/LP trip which accounts for the DNBR margin degradation [
]
Because no figures from previous cycle analyses exist, comparison was made between the transient analysis training manual sample analysis and the figures generated by OPPD for Cycle 8. Pertinent input parameters are summarized in Table 6.3-4.
Figure 6-34 shows the plots of RCS pressure versus time for the initial g
case run without a trip which is used to determine the time manual trip is to be used.
A manual trip is next siniutated at the time of maximum margin degradation, i.e., at the time the maximum RCS Depressurization rate occurs. The maximum RCS Depressurization rate occurs in approximately the first 20 seconds and is constant. Therefore, the time at which a manual trip should occur is arbitrary but must be in the first 20 seconds. A trip time corresponding to a 100 psia drop is adoc, .ste to perform the analysis.
Figure 6-35 shows plots of core power versus time for the Cycle 8 analysis g and CE's example. The core average heat flux curves are found in Figure OPPD-NA-8303-NP, Rev. 03 68 of 125
6.0 TRANSlENT ANALYSIS CODE VERIFICATION (Continued) 6.3 COIDQadSQnS3_01 Ween OPPD Analyses and IndepCQdCDllilaly?,iC12IQV100 Sly Performed by the FucLvendors for vorlflCanon of CESEC.111 (Continued) 6-30. The RCS pressure versus time plots are shown in Figuro 0-37, in g the CE examplo, the initial pressure was 2300 psia, a value which corresponds to the maximum pressure before which the pressurizer sprays will be activated ln a 2700 MWt class plant (whose normal RCS pressuro is 2250 psla), in the Cyclo 8 analysis, a value of 2172 psia was used for the initial RCS pressuro, since the nomial oporating RCS pressure at Fort Calhoun is 2100 psia. The Fort Calhoun pressurizer sprays are fully closed at 2175 psia and fully open at 2225 pala.
The comparison of the figures show good agreement in the trends for the coro power, core average heat flux, and RCS pressure. Tho [
]
6.4 Comparisons Between OPPD Analyses and independent Analyses Previously Performed with CESEC-Ill for Verification of CENTS Of the transients analyzed by OPPD for rotoad coro licensing (using CE methodology) no plant data existed, so a comparison of timning events to previous analyses performed by OPPD using CESEC-Ill was done. The events chosen for comparison Woro:
(1) The Dropped CEA event is dependent upon the initial available overpower margin to prevent excooding the SAFDLs. The goal of the analysis is to dotormino the DNBR required overpower margin (ROPM).
(2) The RCS Depressurization event which is used in the determination of the
[ ). The [ ] accounto for DNDR margin degradation in tho thermal margin / low pressure (TM/LP) trip [
]
OPPD has been using CESEC-Ill to portonn reload safety analysos since 1982 and has successfully completed analysos for six cycles. Thorofore, the predictions from CESEC-Ill provido a reliable baseline against which CENTS can be benchmarked.
OPPD-NA-8303-NP, Rev. 03 69 of 125
6.0 TRANSIENT ANALYWS CODE VERIFICATION (Continued) 6.4 Comoarisons Between OPPD Analyses and indeoendent Analyses Previously Eerformed with CESEC-Illfor Verification of CENTS (Continued) 6.4.1 Drooped CEA The Cycle 12 Dropped CEA analysis performed by OPPD using CENTS was compared to the previous analyses performed by OPPD using CESEC-Ill. The Dropped CEA analyses performed for Cycle 12 using CESEC-lit was used for comparison. Table 6.4-1 summarizes the parameters and their values used for the Cycle 12 analysis. The response of various system parameters to the event are shown in Figures 6-38 to 6-41, All of the figures show good agreement between the CESEC-lit and CENTS predictions.
Figures 6-38 and 6-39 show the response of core power and heat flux as predicted by the CESEC-Ill and CENTS codes. The predicted responses from both codes are similar, with CENTS predicting a higher final power and heat flux, which is conservative. As described in the analysis methodology in Section 5.3, both cases (CENTS and CESEC-Ill) assumed that the turbine admission valves were placed in manual during the event in order to simulate the actual mode of operation of the turbine control system at Fort Calhoun Station. Figuie 6-40 contains plots of tha RCS hot and cold leg coolant temperatures versus timo. The response predicted by CESEC-Ill and CENTS are in good agreement. Figure 6-41 shows plots of the pressurizer pressure versus time. The pressures predicted at 200 seconcis are 2003 psia and 2006 psia by the CESEC-Ill and CENTS codes, respectively. This difference is small enough to be less than the pressure measurement uncertainty.
In summary, the primary system responses between the CESEC-Ill and CENTS predictions show excellent agreement with eacit other.
6A.2 RCS Deoressurization The Cycle 12 RCS Depressurization event was used as a basis for this comparison. This event was analyzed previously using CESEC-It!. An analysis was performed using CENTS in order to compare the predictions from the two codes. Table 6.4-2 summarizes the parameters and their I
values used for the Cycle 12 analysis.
OPPD-NA-8303-NP, Rev. 03 70 of 125
6.0 TRANSIENT ANALYSIS CODE VERIFICATION (Continued) l 6.4 Comparisons Between OPPD Analyses and Indeoendent Analyses Previously -
Performed with CESEC-Ill for Verification of CENTS (Continued) i The CESEC-Ill code does not simulate a thermal margin / low pressure (TM/LP) trip, so a TM/LP trip must be simulated using a manual trip.
CENTS, however, does simulate a TM/LP trip (by virtue of CENTS control system modelling language), so the TM/LP trip was credited in the CENTS simulation of the event, A manual trip is simulated in CESEC-lit at the time of maximum margin degradation, l.e., at the time the maximum RCS Depressurization rate occurs. The maximum RCS depressurl2ation rate occurs in approximately the first 20 seconds and is constant. Therefore, the time at which a manual trip should occur is arbitrary but must be in the first 20 seconds. A trip time corresponding to a 100 psla drop is adequate to perform the analysis.
Figure 6-42 shows plots of core power versus time for the CENTS and CESEC-Ill analyses. As can be seen in the figure, the two codes gave very similar predictions for core power versus time, except for the time of trip, However, as stated previously, the trip in CENTS was based on the actual TM/LP control system, while the time of trip in the CESEC-Ill case was chosen arbitrarily.
Figure 6-43 shows the RCS pressure response from the CESEC-Ill and CENTS analyses. The results show very good agreement post-trip.
The pressurizer pressure versus time plots are shown in Figure 6-44. In both the CESEC-Ill and CENTS analyses, a value of 2172 pala (2150 psia
+ 22 psia measurement uncertainty) was used for the initial RCS pressure.
The CENTS analysis shows an initial depressurization rate slightly higher than that of CESEC-Ill. However, after about 3 seconds both cases depressurized at approximately equal rates.
The comparison of the figures show good agreement in the trends for the core power and pressurizer pressime.
OPPD-NA-8303-NP, Rev. 03 71 of 125
l l
6.0 TRANSIENT ANALYSIS CODE VERIFICATION (Continued) 6.5 Summarv The CESEC-Ill computer code was developed for the analysis of FSAR transient and accident events for the two-by-four loop Combustion Engineering plants. The CENTS code was developed for the simulation of the transient response for a PWR during normal and abnormal conditions including accidents, it is a flexible nodal code that can be used for resolving safoty and operability issues, as well as analyzing FSAR transient and accident events. OPPD's engineering staff was trained for the propor use of CESEC-lit and CENTS and a closo working relationship has boon maintained with both of the CE development teams. To ensure accurato prediction capabilities, both CESEC-Ill and CENTS plant models were developed, inputs tested, and the output verified and validated against both plant data and other SAR analyses.
Benchmarking against Cycle 1 plant data for the Turbine-Reactor Trip and the Four-Pump Loss of Coolant Flow was performed and excellent agreemont between the predicted and observed responses was obtained for both CENTS and CESEC 111. g Verification of CESEC-ill for transients for which plant data was not available was accomplished by performing comparisons between the OPPD Cycle 8 analyses of tne limiting transients and the Cyclo 6 analysos by the fuel vendors (CE and ENC) and, in one case, tne transtont analysis training manual example, in all cases, these benchmarking comparisons showed very good agreement.
Verification of CENTS for transients for which plant data was not available was accomplisht My performing comparisons between OPPD Cycle 12 CESEC-Ill analyses of limiting transients and analysos of the same events using CENTS. In all casos, these benchmarking comparisons showed very good agreement.
OPPD continues to maintain a quality assured model for CESEC and plans to QA the basedeck for CENTS beforo it is used for any licensing analyses. OPPD wlil also continue to provide an update basedeck for each cycle, contained In an Engineering Analysis (EA).
l l
OPPD-NA-8303-NP, Rev. 03 72 of 125
6,0 TRANSIENT ANALYSIS CODE VERIFICATION (Continued)
TABLE 6.3-1 COMPARISON OF PARAMETERS INCLUDING UNCERTAINTIES USED IN THE CEA DROP ANALYSES FOR CYCLES 6 AND 8 Parameter Unlls Cycic_6 Cycle 8 in!tlal Coro Power Lovel MWt 102% of 1500 102% of 1500 Ccro inlet Temperaturo *F S47 547 Pressurizer Prossuro psia 2053 2053 RCS Flow Rato Opm 100,000 197,000 Moderator Temperaturo Cootf. 10 * *F/Ap -2.3 -2.7 Doppler Coeff, Multipiler 1.20 1.15 CEA Insortion at Full Power % insortion 0.0 25.0 Dropped CEA Wortti % d.p- -0.34 -0.28 OPPD-NA-8303-NP, Rov. 03 73 of 125
.6.0 TRANSIENT ANALYSIS CODEVERIFICATION (Continued) ;
TABLE 6.3-2 i COMPARISON OF PARAMETERS INCLUDING UNCERTAINTIES
-USED IN THE HZP MAIN STEAMLINE BREAK ANALYSIS FOR CYCLES 6 AND 8 !
Cycle 6 lbrnmeter. Units Cycle 6 .W Cycle 8 L< initial Core Power Level. MWt 0.0 1.0 1.0 1 Core inlet Temperature '
-F' 532 532 532 l Pressurizer Pressure . psia 2053 2175 2172 RCS Flow Rate -gpm 190,000 190,000 197,000 Effective Moderator Temperature :10 4 Ap/'F -2.3 -2.3 -2.5 Coefficient
-1 Doppler Coeff, Multiplier - 0.8 '1,15 1.15. I
- .\'
= Minimum CEA Scram Wortti : ~ % op" -3.0 -4.2 . -4.0 -
Inilla! Staam Generator Pressure psia N/A 906 895 initial Steam Generator Mass. % Narrow. 63 : 63 70
. inventory (Level)' '
Range Scale
+
t
.+- !
4 L
.7
-i i
OPPD2NA-8303-NP, Rev. 03 I
~
74 L of 125 i
.,,,...,_.,,,,,,,,,.R , , . . . . , , ,
6.0 TRANSIENT ANALYSIS CODE VERIFICATION (Continued)
TABLE 6.* 3 COMPARISON OF PARAMETERS INCLUDING UNCERTAINTIES USED IN THE HFP MAIN STEAMLINE BREAK ANALYSIS FOR CYCLES 6 AND 8 Cyclo 0 l
Mamciti .Unlis DyclLQ .AD6L, Cycitta initial Core Power Level MM 102% of 1500 102% of 1500 102% of 1500 Coro intet Temperaturo 'F '17 547 547 Pr4 urtzar Pressure pois ;vr8 P175 2172 RCS Flow Rato gpm 100,000 100,000 197,000 Moderator Temperatuto 10d op/'F -2.3 -2.3 - 2.5 Coeff6clent Doppler Coeff. Multipilot 0.0 1.15 1.t5 Minimum CEA Scram Worth % Ap -5.81 -5.81 -0.08
- Initial Steam Generator psia N/A 880.5 890 Pressuro initial Steam Generator % Narrow G3 63 70 Mass inventcry (Loyol) Range Scale
- Roduced to -0.57 to account for axial shape.
OPPD-N A-8303-NP, Rev. 03 75 of 125
6.0 TRANSIENT Af.. J's SIS CODE VERIFICATION (Continued)
TABLE 6.3-4 COMPARISON OF PARAMETERS INCLUDING UNCERTAINTIES USED IN THE RCS DEPRESSURIZATION ANALYSES FOR CYCLE B AND EXAMPLE CASE Partmelet untig Examole Catgi Cycip.11 initial Core Power Level MWt 102% of 1000 102% of 1500 Core iniot Temperaturo 'F 547 547 Pressurizer Proscure psia 2300 2172 RCS Flow Rate gpm N/A 209,796 Moderator Temperature 10 4 Ap/'F -2.5 -2.7 Coeffielent Doppler Coeff, Multipiler 1,15 1.15
- Example caso input data consistent with 2'<00 MWI plant operating characteristics.
. OPPD-NA-8303-NR Rev. 03 76 of 125
0.0 TRANSIENT ANALYSIS CODE VERIFICATION (Continued)
TABLE 6.4-1 COMPARISON OF PARAMETERS INCLUDING UNCERTAINTIES USED IN THE CEA DROP ANALYSES FOR CYCLE 12 Ematnetet Unlis .CESECdll CENIS Initial Core Power Lovel MWt 100% of 1500 100% of 1500 Core inlet Temporaturo 'F 545 545 Pressurizer Pressure psla 2075 2075 RCS Flow Rato gpm 208,280 208.280 Modorator Temperature Coeff. 10-4 ap/'F -2.7 -2.7 Doppler Coeff. Multipilor 1.15 1.15 CEA Insertion at Full Power % insortion 25.0 25.0 Dropped CEA Worth %op -0.23 -0.23 t
l OPPD-NA-8303-NP, Rev. 03 l
77 of 125
6.0 TRANSIENT ANALYSIS CODE VERIFICATION (Continued)
TABLE 6.4-2 COMPARISON OF PARAMETERS INCLUDING UNCERTAINTIES USED IN THE RCS DEPRESSURIZATION ANALYSES FOR CYCLE 12 Parameter Untts .CEb11.S. CESEC-Ill Initial Core Power Lovel tAYt 102% of 1500 102% of 1500 Coro inlet Temperature 'F 547 547 Pressurizor Pressure psia 2172 2172 ACS Flow Gato gpm 211,000 211,000 Moderator Tomporaturo 10-4 A p!'F -2.7 -2.7 Coefficlont Doppler Coeff. Muttiplier 1.15 1.15 TM/LP Trip Unit Coefficients a 29 73 N/A 6 18.44 y -11350 A1(Y)* -0.35294Y + 1.08824. Y s .25 N/A 0.57143 Y + 0.875, Y > .25 l PF(B)" 1.0, B .2.100% N/A i
.008B + 1.8, 50% < B < 100%
1.4, B s 50%
- Y = internal ASI g l-
"B = coro power E OPPD-NA-8303-NP, Rev. 03 l
l 78 of 125 l
l
J 3
7.0 REFERENCES
'$.getion 2 References i- 1 2-1 Westinghouse Proprietary Document, ' Report on the Analysis Methods and Evaluation i Models to be Employed in the Large Break and Small Break LOCA Analyses for Fort l Calhoun Unit 1 ", February 1991. !
-l Section 4 References 4-1 CESEC, Digital Simulation of a Combustion Engineering Nuclear Steam Supply System, December,1981, transmitted as Enclosure 1.-P to LD-82-001, January 6, 1982.
4-2 CEN-234(C)-P, Loulslana Power and Light Company, Waterford Unit 3, Docket 50-382, Rotponse to Questions on CESEC, December,1982.
4-3 Letter from A. E. Scherer (CE) to F. J. Miraglia (NRC), *Applicabmty of CESEC-Ill to the Fort Calhoun Station," February 27,1987.
4-4 CENPD-161-P, TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core," Joey,1975.
4-5_ . CEN-191(B)-P, "CETOP-D Code Structure and Modeling Methods for Calvert Cliffs 1 and 2," December,1981. - ;
4-6 CENPD-162-P-A, "CE Critical Heat Flux, Critical Heat Flux Correlation for CE Fuel Assemblies with Standard Spacer Grids Part 1 Un! form Axial Power Distributions,"
September,1976.
'4-7 CENPD-207-P, "CE Critical Heat Flux, Critical Heat Flux Correlation for CE Fuel Assemblies with Standard Spacer Grids Part 2 Non-uniform Axial Power Distributions,"
June,1978.
4-8 Letter from E. G. Tourigny (NRC) to W. C. Jones (OPPD) dated March 15,1983,
'4-9 .OPPD-NA-8301, Rev. 03," Reload Core Analysis Overview", April,1988.
4-10 CEN-257(0)-P, " Statistical Combination of Uncertaintles" November,1983.
'4-11 CE-NPD-282-P, " Technical Manual for the CENTS Code" Volumes 1 and 2, February .
1991.
.4-12 T. A, Porsching, et al.,
- FLASH-4: A Fully implicit FORTRAN IV Program for the Digital -
Simulation of Transients in a Reactor Plan", WAPD-TM-840, March 1969.
4-13 . CE-NPD-282-P, "Techn! cal Manual for the CENTS Code" Volume 3, April 1991, OPPD-NA-8303-NP, Rev. 03 79 of 125=
7.0 REFERENCES
(Continued)
Section 5 References 5-1 CEN-347(0)-P, Rev. 01,
- Omaha Batch M Roload Fuel Design Report", January,1987.
5-2 CEN-121(B)-P, "CEAW, Method of Analyzing Sequential Control Elemont Assembly Group Withdrawal Event for Analog Protected Systems", November,1979.
5-3 CENPD-199-P, Revision 1-P- A, *CE Setpoint Methodology", January 1986.
5-4 Fort Calhoun SER on Automatic Initiation of Auxillary Foodwater, contained in the letter to W. C. Jones from Robert A. Clark, dated February 20,1981.
5-5 Westinghouse Proprietary Document, " Control Element Assembly Ejection Accident Methodology Summary Report for Fort Calhoun Unit 1", April 1991. l l
5-6 OPPD-NA-8302, Rev. 02,
- Nuclear Design Methods and Verifications", April,1988.
5-7 Westinghouso Propriotary Document *Roport on the Analysis Methods and Evaluation Modols to be Employed in the Large Break and Small Break LOCA .inalyses for Fort Calhoun Unit 1 ", February 1991.
5-8 Fort Calhoun SER on Generic Letter 86-06 (TMI Action item II.K.3.5, " Automatic Trip Reactor Coolant Pumps During Loss-of-Coolant Accident"), contained in the letto,' to R. L. Andrews (OPPD) from Anthony Boumla (NRC), dated March 25,1988. E Section 6 References 6-1 *CESEC - Digital Simulation of a CE NSSS", Enclosure 1-P to LO-82-001, January 6, 10P2.
6-2 Letter from A. E. Senerer to F. J. Miraglia, LD-87-013 dated Februan/ 27,1987.
6-3 CE-NPD-282-P, " Technical Manual for the CENTS Code" Volumes 1 and 2, February 1991.
6-4 CE-NPD-282-P, ' Technical Manual for the CENTS Codo' Volume 3, April 1991.
OPPD-NA-8303-NP, Rev 03 80 of 125
l Excess Load Case Data l No Trip Transient Case '
l (CESEC/ CENTS)
Determine [ _ ]
Using Methods of Ref. 5-3 Enable High Power Trip CESEC/ CENTS Case with ( )
Run CETOP eterm ne Emo Maximum Power Determine Time MDNBR Calculate LHR FO "M Calculate DNBR ROPM Excess Load Event Omaha Public Power District Figure ROPM Methods Fort Calhoun Station-Unit No.1 51 l l
oPPD-NA-8303 NP, Rev. 03 81 of 125
100.0 , i i i i i i t Test D a t a -+--
I CENTS Predictions I
CESEC-III Predictions ---
1 I -
80.0 - 1 - l i
I I
Z O I H l P
o 60.0 - -
b I I
b I 3 I O l ;
40.0 - I -
N I st; l i
O U l I
4
< l 1
20.0 - i I
\
\
y W =- _- -r ---._ ___._ _
0.0 O 5 10 15 20 25 30 35 40 TIME, SECONDS l
t Full Power Turbine Trip Omaha Public Power District Figure Core Power vs Time Fort Calhoun Station Unit No.1 61 OPPD-NA-8303-NP, Rev. 03 f
82 of 125
i l
l l
i 2300 i T- i i i T e s t D a t a -+---
CENTS Predictions CESEC-III Predictions - --
2200 - -
m d
'n CL
- 2100 aw -
$ s D s
$ .n h .
2000 -
%'s s -
m ~ ___
s --
j ____.
H ,; --^ %~:
X v v ~
@ 1900 - -
u) ta 1800 - -
1700 0 10 20 30 40 50 60 TIME, SECONDS Full Power Turbine Trip Omaha Public Power District Figure '
Pressurizer Pressure vs Time Fort Calhoun Station Unit No.1 62 OPPD-NA-8303-NP, Rev. 03 83 of 125
IN 90-49 August 6, 1990 Page 2 of 5 The ECT/RFC inspections revealec 104 tubes with circumferential cracks at the expansion transition. The macrocracks, as defined by ECT/RPC, consisted of several discontinuous microcracks that were separated by small ligaments of sound material. The discontinuous nature of the array of microcracks was confirmed by the UT and examination of the removed tube specimens. As measu~ec by UT, the macrocracks ranged in circumference from 84 degrees to 329 degrees and ranged in depth up to 100-percent throughwall.
All tubes with crack indications were staked and plugged. In addition, the licensee evaluated the residual strength of the cracked tubes to assess their capability to sustain normal operating and postulated accident loadings before
, their removai from service. This structural evaluation considered the profiles for each crack obtained from the UT examination. This evaluation revealed one cracked tube which failed to meet the ASME Coce,Section III, NB-3225 and Appendix F stress limits for postulated accident conditions. (Regulatory Guide 1.121, " Bases for Plugging Degraded FWR Steam Generator Tubes," states that margins should be censistent with the stress limits in Section III of the code.) Based on these findings, the staff concludes that the integrity of the subject tube was not ensured under postulated accident conditions.
The staff has recently identified service induced, circumferential SCC, such as at Millstone Unit 2, to be a source of significant degradation to tubes in FWR steam generators. Such cracking is particularly noteworthy because it is generally not detectable with conve')tional bobbin probes used reutinely for inservice inspection. Such cracking is generally only detectable through the use of speciali:ed probes, such as the RPC probe.
Most circumferential cracking has been observed at tube expansion transitions at or near the top of the tubesheet, in addition to Millstone Unit 2, circumferential cracking at the expansion transition has recently been identified at one other Combustion Engineering (CE) plant (Maine Yankee), at three plants with Westinghouse Model 51 steam generators (North Anna Unit 1, Trojan Unit 1, and Sequoyah Unit 1), and at one plant with Westinghouse Model 0 steam generators (McGuireUnit1). Tubes in the affected CE and Westinghouse Model 51 steam generators were explosively expanded against the tubesheet. Tubes in the McGuire Model D steam generators were expanded against the tubesheet by mechanical rolling.
In addition to being found at the expansion transition location, widespread circumferential SCC has been observed at drilled-hole support plate locations atPalisades(CEsteamgenerators). Isolated instances of circumferential SCC have been reported at the uppermost support plate of a pre-replacement Westinghouse Model 44 steam generator of Indian Point Unit 3 ar.d at a row I U-bend of a Mocel 51 steam generator at Zion Unit 1. The circumferential SCC at Palisades and Indian Point Unit 3 appears to be associated with significant denting at the support plates.
4
,- , - , , - w y -
l i
1 560 , , , , , l Test D a t a - Loop 1 ">- )
Test Data - Loop 2 O tu 555 -
CENTS Predictions -
n CESEC-III Predictions --- I W :
e 550 -
W Q
t ,45 -
a w 540 -
El h b
(
k -.
530 7' 1% ,,x gQ m
O
% 525 -
520 O 10 20 30 40 50 60 TIME, SECONDS Full Power Turbine Trip Omaha Public Power District Figure RCS Cold Leg Temperatures Fort Calhoun Station-Unit No.1 6-3 OPPD-NA-8303-NP, Rev. 03 84 of 125
590 , , , , ,
Te s t D a t a - Loop 1 -+-
Test Data - Loop 2 -0 W 580:
CENTS Predictions 1 CESEC-III Predictions -- ~
y ,'s
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\ \\ -
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y '-
x 530 -
520 O 10 20 30 40 50 60 TIME, SECONDS i
l l
l Full Power Turbine Trip Omaha Public Power District Figure RCS Hot Leg Temperatures _
Fort Calhoun Station Unit No.1 64 OPPD-NA-8303-NP, Rev. 03 85 of 125
i l
1 l
1 1000 , , , , ,
Test Data - +
950 - CENTS Predictions ,,
CESEC-III Predictions ---
900 *
n ad% ,
f %3nge w# #
5 850 -/ -
n 04
- 800 L ta <
cr.
D
$ 750 - -
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=
0 m 650 - -
600 - -
550 - -
500 O 10 20 30 40 50 60 TIME, SECONDS Full Power Turbine Trip Omaha Public Power District Figure Steam Generator #1 Pressure Fort Calhoun Station Unit No.1 65 OPPD-N A-8303-NP, Rev. 03 86 of 125
1000 , , 4 4 ,
Tect Data +
950 . CENTS Predictions _
CESEC-III Predictions - -
900 -
Cs , _. _ ,_ _ _ _ ,gd+;w ,
I 5
7 , %n ,~.x #""'
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$ 750 -
N W
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e m 650 -
600 -
550 -
500 ' '
O 10 20 30 40 50 60 TIME, SECONDS Full Power Turbine Trip Omaha Public Power District ' Figure Steam Generator #2 Pressure Fort Calhoun Station Unit No.1 66 OPPD-NA-8303-NP, Rev. 03 87 of 125
4.0 , , , , ,
Test Data - SG #1 Test Dat a - SG # 2 -
3.5 -
CENTS Predictions -
CESEC-III Predicti. ns ---
u (;1 C 3.0 -
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l l
l l
l 1
4.0 , , , , ,
1 Test D at a +-
CENTS Predictions 3.5 -
CESEC-III Predictions -- -
n 3.0 ,- -
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o H
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s ' -
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0 10 20 30 40 50 60 TIME, SECONDS Full Power Turbine Trip Omaha Public Power District Figure Steam Generator #1 Steam Flow Fort Calhoun Station Unit No.1 68 OPPD-NA-8303-NP, Rev. :33 89 of 125
l 4.0 . . . . .
Test Data "
CENTS Predictions l 3.5 -
CESEC-III Predictions -- - l l
l l
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0.0 -
0 10 20 30 40 50 60 TIME, SECONDS Full Power Turbine Trip Omaha Public Power District Figure Steam Generator #2 Steam Flow Fort Calhoun Station Unit No.1 69 OPPD-NA-8303-NP, Rev. 03 90 of 125
i
- 1. O < ,y i , ,
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g' Test Data -
\ CENTS Predictions 0.9 -
\\ CESEC-III Predictions -- -
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0.3 -
0.2 0 5 10 15 20 25 30 TIME, SECONDS Four Pump Loss of Flow Omaha Public Power District Figure Total RCS Flow Fraction vs Time Fort Calhoun Station Unit No.1 6 10 OPPD-N A-8303-NP, Rev. 03 91 of 125
i
- n. n. _
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=
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m - i w ;
C l C1.
1800 ' '
0 5 10 15 20 25 30 TIME, SEC0tlCE NOTE :
CYCLE i (FULL POWER = 1420 MWt)
IMITIAL POWER = 35%
PLANT DATA: TEST PERFORMED MARCH 5, 1974 Four Pump Loss of Flow Omaha Public Power District Figure Pressurizer Pressure vs Time Fort Calhoun Station-Unit No.1 6 11 OPPD-NA-8303-NP, Rev. 03 92 of 125
SO . . ,
l i i l
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a 40 --
i i 0 5 10 15 20 25 30 TIME, SECONDS ,
NOTE :
CYCLE i (FULL POWER = 1420 MWt)
INITIAL POWER = 35% '-
PLANT DATA: TEST FERFORMED MARCH 6, 1974 Four Pump Loss of Flow Omaha Public Power District Figure Pressurizer Level vs Time Fort Calhoun Station-Unit No.1 6 12 OPPD-NA-8303-NP, Rev. 03 93 of 125
50 .
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0 0 5 10 15 20 25 30 TDiE, SECONDS NOTE :
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PLANT DATA: TEET PE.:EORMED MAR.H E, . 1974 Four Pump Loss of Flow Omaha Public Power District Figure Core Power vs Time Fort Calhoun Station Unit No.1 6 13 OPPD-NA-8303-NP, Rev. 03 94 of 125
l n
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Four Pump Loss of Flow Omaha Public Power District Figure' Steam Generator Pressure vs Time Fort Calhoun Station Unit No.1 6 14 OPPD-NA-8303-NP. Rev. 03 95 of 125
-_$.4
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PL/ ,':T DAT ': TEET CERFORMED MARCH 6, 1974 Four Pump Loss of Flow Omaha Public Power District Figure i RCS Temperatures vs Time Fort Calhoun Station Unit No.1 6-15 l
OPPD-NA-8303-NP, Rev 03 90 of 125
- 2
=
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NOTE :
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ii a .
PLAtli DAT!: TE37 PERFORMED MARCH 6. 1974 Four Pump Loss of Flow Omaha Public Power District_
Figure Main Feedwater Flow vs Time Fort Calhoun Station Unit No.1 6 16 OPPD-NA-8303-NP, Rev. 03 97 of 125
1
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Four Pump Loss of Flow Omaha Public Power District Figure Steam Flow vs Time Fort Calhoun Station Unit No.1 6-17 OPPD-NA-8303-NP, Rev. 03 98 of 125
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CYCLE 5: ENC ANALYSIS CYCLE 8: OPPD ANALYSIS CEA Drop incident Omaha Public Power District Figure Core Power vs Time Fort Calhoun Station Unit No.1 6 18 OPPD-NA-8303-NP, Rev. 03 99 of 125
.... n. ,
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w 60 0 20 40 60 80 100 120 140 160 TIME, SECONDS MOTE :
CYCLE 5; ENC ANALYSIS CYCLE 2: OPPD ANALYSIS CEA Drop incident Omaha Public Power District Figure Core Average Heat Flux vs Time Fort Calhoun Station Unit No.1 6-19 OPPD-NA-8303-NP, Rev. 03 100 of 125
l I, =. e n.
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in 8
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C a
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500 ' ' i i 0 20 40 60 80 100 120 140 160 TIME, SECONDS NOTE .
CYCLE 6: ENC ANALYSIS CYCLE S: OPPO ANALYSIS CEA Drop incident Omaha Public Power District Figure Coolant Temperature vs Time Fort Calhoun Station Unit No.1 6-20 OPPD-NA-8303-NP, Rev. 03 101 of 125
k
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1500 - -
1850 -
0 20 40 60 80 100 120 140 160 TIME, SECONG7 NOTE :
CYCLE 5: EMC AN AL'f SIS CYCLE 8: OPPD ANAL'(SIS CEA Drop Incident Omahr Public Power District Figure Pressurizer Pressure vs Time Fort Ct. 4.oun Station Unit No.1 6-21 1
l OPPD-NA-8303-NP. Rev. 03 102 of 125
e s
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.-NOTE :
)
p- CYCLE 6: ENC lNALYSIS WITH SDH='3,0% A:-
u e CYCLEE8! 10P:D ANALYSIS;WITH SgMy la,og;
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.Zero Power-MSLB Accident : Omaha Public Power District Figure Core. Power vs Time - Fort Calhoun Station-Unit No.1 6-22 - . _ _
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CYCLE E: ENC ANALYSIE WITH 50M= 3.0% cf CYCLE 6 AFW:
CE ANALYSIS WITH 50M= 4.2%af CYCLE E: OFF0 ANALYSIS WITH 50M= 4.0%af Zero Power MSLB Ac::ident Omaha Public Power District Figure Core Average Heat Flux vs Time Fort Calhoun Station-Unit No.1 6 23 i OPPD-NA-8303-NP, Rev. 03 104 of 125
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CYCLE !: CELANALYSIS
' CYCLE 5 AFW: CE ANAL'/SIE CYCLE E; OPPD ANALYSIS Zero Power MSLB Accident Omaha Public. Power District Figure Coolant System Pressure vs Time Fort Calhoun-Station-Unit No.1, 6-25 1
-OPPD-NA-8303-NP, Rev. 03 100 'of.125 ,
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Zero Power MSLB Accident Omaha Public Power District Figure Steam Generator Pressure vs Tirne Fort Calhoun Station-Unit No.1 6-26 1 l
i OPPD-NA-8303-NP, Rev. 03 107 of 125
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CYCLE 6: ENC ANALYSIS (CEA WORTH = 5 ei, L g)
C,/C, L : o AFW: CE ANALYSIS (CEA WORT'"= *~ 'e* 4 ~V)
CYCL': c^: CF?D ANALYSIS (CEA WORTH = 6.57h,g)
Full Power MSLB Accident Omaha Public Power District Figure Core Power vs Time Fort Calhoun Station Unit No.1 6-27 OPPD-NA-8303-NP, Rev. 03 108 of 125
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CYCLE-E: OFF0 ANALYSIS ,
Full Power MSLB-Accident- ' Omaha Public Power District Figure Coolant System Pressure vs Time Fort Calhoun Station-Unit No.1 -
6 30 OPPD-NA-8303-NP, Rev. 03
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. OPPD-NA-8303-NP, Rev. 03 112 of 125-
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Full Power MSLB Accident Omaha Public Power District Figure RCS Coolant Temperatures vs Time Fort Calhoun Station-Unit No.1 6 32 oPPD-NA-8303-NP, Rev. 03 113 of 125
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CYCLELS: 'OPFD ANALYSIS 1
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CYCLE 2: CE ANALYSI5 50R:2700.MWt1 UNIT .i OPPD ANALYSIE 1
-i RCS'Depressurization incident Omaha Public Power District - . Figure RCS Pressure vs Time ' Fort Calhoun Station-Unit No.1 6 34 OPPD-NA-8303-NP, Rev. 03 115 of 125
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CYCLE S: OPPD ANALYSIS-- .
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CYCLE S: 0PPO ANALYSIS- "
RCS Depreslsurization incident Omaha Public Power Olstrict Figure RCS Pressure vs Thel Fort Calhoun Station-Unit No.1 6-37 OPPD-NA-6303-NP, Rev. 03 118 of 125
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