ML20195E485
| ML20195E485 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 10/25/1988 |
| From: | Lindquist T, Odell L SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML20195E483 | List: |
| References | |
| ANF-87-161-S01, ANF-87-161-S1, NUDOCS 8811080110 | |
| Download: ML20195E485 (200) | |
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ANF-87-161 SUPPLEMENT 1 ty ' h ,,
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& ADVANCED NUCLEAR FUELS CORPORATION MILLSTON$ UNIT 2 PLANT TRANSIENT ANALYSIS REPORT ANALYSIS OF CHAPTER 15 EVENTS OCTOBER 1988 1
ADVANCED NUCLEAR FUELS CORPORATION ANF-87-151 Supplement 1 Issue Date: 10/25/88 MILLSTONE UNIT 2 PLANT TRANSIENT ANALYSIS REPORT -
ANALYSIS OF CHAPTER 15 EVENTS Prepared by: /K ts T. R. Lin1quist. Engineer PWR Safety Analysis Licensing & Safety Engineering Fuel Engineering & Technical Services Prepared by: . /
L. D. O' Dell, Team Leader
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PWR Safety Analysis Licensing & Safety Engineering fuel Engineering & Technical Services October 1988 gf .
i CUSTOMER OlSCLAlWER ledPCMTANT NOTICE R(QARD'NO CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFUU,Y Aovenced Nucseat Fueis Corporanon's warrandes and reoresentatens con.
comeg the subrect matter of me document are those set forro e too Agreement between Aovenced Nuc'oer Fvem Corporata and the Cwstemer pu swant *o wrHch thir drent 4 40ued. Ac0T10egry, exceCl Es othermee enDressly pro-vided in suca AQ4emort ne@er Advanced NuC' ear Fuevs Corporaten not any EsrSon acteg on as con 4Jf manet Spy warrP.nfy or representaten exoressed or imohed, etn resoect to the accuracy, compieteness, or useNeess of tre in er.
meDon contamed in this coCument or that the use of any c%rmaton ao3aratus.
method or process ceciesed m trhe cocument edi not inmnge prwatory cweed ngats: or assumes any haomtes with resocct to the use of any c%rmaton. ao-paratus momed or process ciecosed n trns document.
The mermanon contaced herom e for the so.e use of Catomvr.
In oroer to evod anoerment of ngnes of Advanced Nue ear Fuois Corporaten e partnts or evermons encn may De cefuced in the informaton centamed e the document, the reccent, Dy as accoorance of mis cocument. agrees not to puchan or mame puesic use (e the patent use of tro term) of sucn ceresten untW so authnrued e entog Dy Advanced Nuc: ear Fue's Corporston or unte a*er sta (6) montns breweg termeaten or enoirsten of tre aerosa4 Agreement and any ortensen thereof. uness etnermse encrossty proviced e tre agreemoet. No ngnts or heensee e of to any parents are irved Dy me Nmisnieg of mis cocu-ment.
ANF-3145 47;A (12 $7)
! . ANF-87-lCl l Supplement 1 4
Page i i
j TABLE OF CONTEN11 Section EAga
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i.0 intr 00uCTiON . . . . . . . . . . . . . . . . . . . . . . . . . . i 2.0
SUMMARY
AND CONCtVSIONS .................... 2 3.0 ANALYSIS OF PLANT TRANSIENTS . . . . . . . . . . . . . . . . . . 11 15.0 ACCIDENT ANALYSES ....................... 12 15.0.1 CLASSIFICATION OF PLANT CONDITIONS . . . . . . . . . . . . . . 12 15.0.1.1 Acceptance Criteria ...................... 13 15.0.1.2 Classification Of Accident E*.1nts By Category ......... 14 15.0.2 PLANT CHARACTERISTICS AND INITIAL CONDITIONS . . . . . . . . . . 16 15.0.3 POWER DIS 1RIBUT10N . . . . .................. 19 15.0.4 RANGE OF PLANT OPERATING PARAMETERS AND STATES . . . . . . . . . Z2 15.0.5 RLtCTIVITY COEFFICIE775 USED IN THE SAFETY ANALYSIS ...... 24 15.0.6 SCRAM INSERTION CHARAC1 ERISTICS ................ 26 15.0.7 TRIP SETPOINT VERIFICATION . . . . . . . . . . . . . . . . . . . 27 4
15.0.7.1 Reactor Protection System ................... 27 15.0.7.2 Specified Acceptable Fuel Design Limits (SAFDLs) . . . . . . . . 28 15.0.7.3 Limiting Safety System Settingr. . ...... ....... 29 15.0.7.3.1 Local Power Distribution ................. 29 15.0.7.3.2 Thermal Hargin/ Low Pressure . . . . . . . . . . . . . . . . 29 15.0.7.3.3 Additional Trip Functions . . . . . . . . . . . . . . . . . 30 15.0.7.4 Limiting Conditionr for Operation ............... 30 15.0.7.4.1 DNB Monitoring ................ ..... 30
. ANF-87-161 l
i Supplement 1 Page 11 l TABLE OF CONTENTS 1
1 Section East l
l i 15.0.7.4.2 Linear Heat Rate Monitoring . . . . . . . . . . . . . . . . 31 15.0.7.5 Setpoint Analysis ....................... 31 15.0.7.5.1 Limiting Safety SysteA Settings . . . . . . . . . . . . . . 31 q
15.0.7.5.2 Limiting Canditions for Operation . . . . . . . . . . . . . 32 -
l 15.0.9 PLANT SYSTEMS AND COMP 0NENTS AVAILABLE FOR MITIGATION OF ACCIDENT EFFECTS . . . . . . . . . . . ............ 50 .
15.0.10 EFFECTS OF MIXE0 ASSEMBLY TYPES AND FUEL 900 BOWING ...... 56 15.0.11 PLANT LICENSING BASIS AND SINGLE FAILURE Ck!TERIA ....... 57 15.0.12 PLOT VARIABLE NOMENCLATURE . . . . . . . . . . . . . . . . . . 60 -
15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM . . . . . . . . 62 l
l 15.1.3 INCREASE IN STEAM FLOW . . . . . . . . . . . . . . . . . . . . . 62 l
15.2 DECREASE IN HEAT REMOVAL SY THE SECONDARY SYSTEM . . . . . . . . 73 15.2.1 LOSS Of EXTERNAL LOAD ..................... 73 15.2.4 CLOSURE OF A SINGLE MAIN STEAM ISOLATION VALVE (MSIV) ..... 89 15.2.7 LOSS OF NORMAL FEEDWATER FLOW ................. 98 l 15.! DECREASE IN REACTOR COOLANT SYSTEM FLOW . . . . . . . . . . . 110 f l
15.3.1 LOSS OF FORCED REACTOR COOLANT FLOW . . . . . . . . . . . . . 110 15.3.3 REACTOR CGOLANT PUMP ROTOR SElZURE , . . . . . . . . . . . . 120 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES . . . . . . . . . 130 15.4.1 UNCONTROLLED CONTROL R0D ASSEMBLY (CRA) WITHDRAWAL FROM A .
SUBCRITICAL OR LOW POWER STARTUP CONDITION . ......... 130
. s 15.4.2 UNCONTROLLED CONTROL R00 BANK WITHDRAWAL AT POWER ... . . . 139
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15.4.3 CONTROL R00 MISOPERATION , . . . . . . . . . . . . . . . . . . 148
ANF 87 161 Supplement 1 Page iii TABLE OF CONTENTS Section Elag 15.4.6 CVCS HALFUNCTION THAT RESULTS IN A DECPEASE IN THE BORON CONCENTRATION IN THE REACTOR COOLANT . . . . . . . . . . . . . 153 15.4.8 SPECTRUM 0F CONTROL R00 EJECTION ACCIDENTS . . . . . . . . . . 158 15.6 DECREASES IN REACTOR COOLANT INVENTORY . . . . . . . . . . . . 175 15.6.1 INADVERTENT OPENING OF A PWR PRESSURIZER PRESSURE REllEF VALVE . . . . . . . . . . . . . . . . . . . . . . . . . 175
4.0 REFERENCES
. . . . . . . . . . . . . . . . . . . . . . . . . . 185
Af'F 161 Supplement 1 Page iv LIST OF TABLES Table P_agt 2-1 Disposition of Events Sumary . . . . . . . . . . . . . . . 3 2-2 Sumary of Results .................... 9 15.0.1-1 Accident Category Used for Each Analyzed Event ...... 15 15.0,2-1 Plant Operating Conditions ................ 17 15.0.2-2 Nominal Fuel Design Parameters .............. 18 15.0.3-1 Core Power Distribution ................. 20 15.0.4-1 Range of Key Initial Condition Operating Parameters . . . . 23 15.0.5 1 Reactivity Parameters . . . . . . . . . . . . . . . . . . . 25 15.0.7 1 Trip Setpoints .................... .. 35 15.0.7-2 Uncertainties Applied in LSSS Calculations ........ '36 15.0.7-3 Uncertainties Applied in the TM/LP LSSS Calculations ... 37 15.0.7-4 Uncertainties Applied in the LC0 Calculations . . . . . . . 38 15.0.7-5 Uncertainties Applied in DNB LC0 Calculations . . . . . . . 39 15.0.3 1 Component Capacities and Setpoints ............ 49 15.0.9-1 Overview of Plant Systems and Equipment Available for Transient and Accident Conditions . . . . . . . . . . . . . 51 15.0.12-1 Nomenclature Used in Plotted Results ........ .. 61
, 15.1.3-1 Event Summary for Increase in Steam Flow (Excess load) .. 65 15.2.1 4 Event Sumary for the Loss of External Load Event . .... 76 15.2.4-1 Event Sumary for the MS!V Closure .... ...... 92 15.2.7-1 Event Sumary for the loss of Normal feedwater Event . . 101 i _ _ _ _ _ _ _
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. ANF-87-161 Supplement 1 Page v LIST OF TABLES Table Page 15.3.1-1 Event Sumary for the Loss of Forced Reactor Coolant Flow . . . . . . . . . . . . . . . . . . 112 15.3.3-1 Event Summary for a Reactor Coolant Pump Rotor Seizure . 122 15.4.1-1 Event Sumary for the Uncontrolled Bank Withdrawal from Low Power Event . . . . . . . . . . . . . . . . . . 132 15.4.2-1 Event Summary for the Uncontrolled Rod Bank Withdrawal Event for the Limiting 100% Power Case . . . . 141 15.4.3-1 Sumary of MONBRs for Control Rod Misoperation Events . . 152 15.4.6-1 Summary of Results for the Boron Dilution Event for All Modes Where Slug Flow Conditions May Exist . . . 157 15.4.8-1 Event Sumary for a Control Rod Ejection . . . . . . . . 161 15.4.8 2 Bounding BCC/EOC Ejected Rod Analysis . . . . . . . . . . 162 l 15.6.1-1 Event Sumary for an inadvertent Opening of a PWR Pressurizer Pressure Relief Valve . . . . . . . . . . 178 l
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1 ANF-87-161 Supplement 1 Page vi LIST OF FIGURES Fiaure Eagg 15.0.3 1 Total Radial Peaking Factor . . . . . . . . . . . . . . . . 21 15.0.7-1 Verification Local Power Density Limiting Safety System Setting . . . . . . . . . . . . . . . . . . . . . . . . . . 40 15.0.7-2 TM/LP Trip function Al .................. 41 15.0.7-3 TM/LP Trip Function QR1 . . . . . . . . . . . . . . . . . . 42 15.0.7-4 Verification of DNB Limiting Condition of Operation for the CEA Drop Transient ................ 43 15.0.7-5 Verification of DNB Limiting Cond.ition of Operation for the Loss of Forced Coolant riow transient . . . . . . . 44 15.0.7-6 Verification of Local Powe, Density Limiting Condition of Operation .................. 45 15.0.7-7 1.inear Heat Rate LCO Used in LPD LCO Verification . . . . . 46 15.0.7-8 Axial Peaking Augmentation Factor Used in LPD LSSS and LPD LCO Verification ................. 47 15.1.3-1 Reactor Power Level for Increase in Steam Flow (Rated Power) . . . . . . . . . . . . . . . . . . . . . . . 66 15.1.3 2 Core Average Heat Flux for increase in Steam Flow (Rated Pcwer) . ..................... 67 15.1.3-3 Reactor Coolant System Temperatures for increase in Steam flow (Rated Power) . . . . . . . . . . . . . . . 68 15.1.3-4 Pressurizer Pressure for increase in Steam Flow (Rated Power) . . . . . . . . . . . . . . . . ....... 69 15.1.3 5 Reactivities for increase in Steam Flow (Rated Power) . . . 70 15.1.3-6 Secondary Pressure for increase in Steam Flow (Rated Power) . . . . . . . . . . . . . . . . . . . . . . . 71
ANF-87-161 Supplement 1 Page vil LIST OF FIGURES Fiaure Eagg i
15.1.3-7 Secondary Steam and Feedwater Flow Rates for ;
increase in Steam Flow (Rated Power) ........... 72 i l
15.2.1-1 Reactor Power Level for loss of External Load (Pressurization Case) . . . . . . . . . . . . . . . . . . . 77 15.2.1-2 Core Average Heat Flux for loss of External Load i
(Pressurization Case) . . . . . . . . . . . . . . . . . . . 78 15.2.1-3 Reactor Coolant System Temperatures for Loss 1 of External Load (Pressurization Case) .......... 79 15.2.1-4 Pressurizer Pressure for loss of External Load (Pressurization Case) . . . . . . . . . . . . . . . . . . . 80 15.2.1-5 Reactivities for loss of External Load (Pressurization Case) . . . . . . . . . . . . . . . . . . . 81 l
15.2.1 6 Secondary Pressure for Loss of External Load (Pressurization Case) . . . . . . . . . . . . . . . . . . . 82 15.2.1-7 Reactor Power Level for Loss of External Load (MONBRCase) .................. .... 83 15.2.1-8 Core Average Heat Flux for loss of External Load (HDNBRCase) ....................... 84 15.2.1-9 Reactor Coolant System Temperatures for loss of External Load (MLNBR Case) . . . . . . . . . . . . . . . 85 l 15.2.1-10 Pressurizer Pressure for loss of External Load (MONBR Case) ....................... 86 15.2.1-11 Reactivities for loss of External load (MONBR Case) . . . . 87 15.2.1-12 Secondary Pressure for Loss of External Load (MDNBR Case) ........,.............. 88 15.2.4-1 Reactor Power Level fo. MSIV Clcsure . ... ... .. . 93 15.2.4-2 Reactor Coolant System Temperatures for MSlv Closure ... 94 15.2.4-3 Pressurizer Pressure for MSly Closure . . . . . . . . . . . 95 r
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ANF-87-161 Supplement 1 Page viii LIST OF FIGURES Fiqure Elag 15.2.4-4 Reactivities for MSIV Closure . . . . . . . . . . . . . . . 96 15.2.4-5 Secondary Prissure for MSIV Closure . . . . . . . . . . . . 97 15.2.7-1 Core Average Temperature for Loss of Normal Feedwater Flow (Steam Generator Inventory) . . . . , .
102 15.2.7-2 Secondary Pressure for Loss of Normal feedwater Flow (Steam Generator Inventory) . . . . . . . . . . . . 103 15.2.7-3 Pressurizer Liquid Volume for loss of Normal Feedwater Flow (Steam Generr'or Inventory) . . . . . . . 104 15.2.7-4 Secondary Side liquid Level for loss of Normal Feedwoter Flow (Steam Generator Inventory) . . . . . . . 105 15.2.7-5 Core Average Temperature for loss of Normal Feedwater Flow (Pressurizer Inventory) . . . . . . . . . 106 15.2.7-6 Secondary Pressure for loss of Normal Feedwater Flow (Pressurizer Inventory) . . . . . . . . . . . . . . . . . 107 15.2.7-7. Pressurizer liquid Volume for Loss of Normal Feedwater Flow (Pressurizer Inventory) . . . . . . . . . 108 15.2.7-9 Secondary Side Liquid Level for Loss of Normal feedwater Flow (Pressurizer inventory) . . . . . . . . . 109 15.3.1-1 Reactor Power Level for Loss of Forced Flow . . . . . . . 113 15.3.1-2 Core Average Hest Flux for loss of Forced Flow . . . . . 114 15.3.1-3 Reactor Coolant System Temperatures for Loss of Fceced Flow . . . . . . . . . . . . . . . . . . . . . , 115 15.3.1 4 Pressurizer Pressure for loss of Forced flow . . . . . , 116 15.3.1-5 Reactivities for Loss of Forced Flow . . . . . . . . . . , 117 15.3.I 6 Primary Coolant Flow Rate for loss of Forced Flow . . . , 118 15.3.1-7 Secondary Pressure for loss of Forced Flow . . . . . . . 119
ANF-87-161 Supplement 1 Page ix LIST 0F FIGURES Fiaure Elag 15.3.3-1 Reactor Power Level for Reactor Coolant Pump Rotor Seizure . . . . . . . . . . . . . . . . . . . 123 15.3.3-2 Core Average Heat Flux for Reactor Coolant Pump Rotor Seizure . . . . . . . . . . . . . . . . . . . 124 15.3.3-3 Reactor Coolant System Temperatures for Reactor Coolant Pump Rotor Seizure . . . . . . . . . . . 125 15.3.3 4 Pressurizer Pressure for Reactor Coolant Pump Rotor Seizure . . . . . . . . . . . . . . . . . . . . . . 126 15.3.3-5 Reactivities for Reactor Coola,nt Pump Reter Seizure . . . 127 15.3.3 6 Primary Coolant Flow Rate for Reactor Coolant Pump Rotor Seizure .................. . 128 15.3.3-7 Secondary Pressure for Reactor Coolant Pump i Rotor Seizure . . . . . . . . . . . . . . . . . . . . 129 l 15.4.1-1 Reactor Power Lovel for Low Power Bank Withdrawal . . . . 133 15.4.1-2 Core Average Heat Flux for low Power Bank Withdrawal . . 134 l
l 15.4.1-3 Reactor Coolant System Temperatures for Low Power Bank Withdrawal . . . . . . . . . . . . . . . . . . 135 15.4.1-4 Pressurizer Pressure for low Power Bank Withdrawal .. . 136 15.4.1 5 Reactivities for low Power Bank Withdrawal . . . . . . . 137 15.4.1-6 Secondary Pressure for low Power Bank Withdrawal . . . . 138 15.4.2-1 Reactor Power Level for an Uncontrolled Bank Withdrawal at Power . . . . . . . . , . . . . . . . . . . 142 15.4.2 2 Core Average Heat Flux for an Uncontrolled Bank Withdrawal at Power . . . . . . . . . . . . . . . . . . . 143 15.4.2-3 Reactor Coolant System Temperatures for an Uncontrolled Bank Withdrawal at Power . . . . . . . . . . 144
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ANF-87-161 Su'pplement 1
-Page x y
LIST OF FIGURES f.19Ert P.lat L 15.4.2.4 Pre .zer Pressure for an Uncontrolled Bank i '
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.W iths .wal e Powe r . . . . . . . . . . . . . . . . . . .
15.4.2-5 Reactivities for an Uncorarolled Bank Withdrawal !
at Power . . . . . . . . . . . . . . . . . . . . . . . . 140 15.4.2-6 Secondary Pressure for an Uncontrolled Bank Withdrawal r at Power ..................... . . 147 l 15.4.8-1 Reactor Power Level for a Control Rod Ejection ,
(MONBR Case) . . . . . . . . . . . . . . . . . . . . . . 163 l
15.4.0-2 Core Average Heat Flux for a Control Rod Ejection ,
(MONBR Case) . . . . . . . . . . . . . . . . . . . . . . 164 15.4.8-3 Reactor Coolant System Temperatures for a Control Rod Ejection (MDNBR Case) . . . . . . . . . . . . . . . . 105 4
15.4.8 4 Pressurizer Pressure for a Control Rod Ejsetion (MONBR Case) .................... . 166 :
, 15.4.8 5 Reactivitics for a Control Rod Ejection (MONBR Case) . 167 7 1
15.4.8 6 Secondary Pressure for a Control Rod Ejection (MDNBR Ca.e) . . . . . . . . . . . . . . . . . . . . . 168 -
15.4.8 7 Reactor Power Level for a Control Rod Ejection '
(Pressure Case) . . . . . . . . . . . . . . . . . . . . . 169 .
15.4.8-B Core Average Heat Flux for a Control Rod Ejection [
(Pressure Case) . . . . . . . . . . . . . . . . . . . . . 170 1r t.8 9 Reactor Coolant System Temperatures for a Control l Rod Ejection (Pressure Case) . . . . . . . . . . . . . .
171 Pressurizer Pressure for a Control Rod Ejection 15.4.8-10 (Pressure Case) . . . . . . . . . . . . . . . . . . . . . 172 i i
i 15.4.8 11 Reactivities for a Control Rod r.jection !
(Pressure Case) . . . . . . . . . . . . . . . . . . . 173
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ANF-87-161 L Supplement 1 j Page xi l
! LISL_0f FIGURES Fiaure Eigg l
l 15.4.8 12 Secondary Pressure for a Control Rod Ejection (Pressure Case) . . . . . . . . . . . . . . . . . . . . . 174 r
[ 15.6.1-1 Reactor Power Level for an Inadvertent Opening of a PWR L Pressurizer Presrure Relief Valve (Rated Power) . . . . . 179 15.6.1-2 Core Average Heat Flux for an Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve (Rated Power) . . 180 15.6.1-3 Reactor Coolant System Temperatures for an Inadu atent Openir.g of a FWR Pressurizer Pressure l Relief Valve (Rated Power) . . . . . . . . . . . . . . . 181 15.6.1-4 Pressurizer Pressure for an Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve (Rated Power) . . 182 15.6.1-5 Reactivities for an Inadvertent Opening of a PWP.
Pressurizer Pressure Relief Valve (Rated Power) . . . . . 183 15.6.1-6 Secondary Pressure i Inadvertent Opening of a PWR Pressurizer Pressure . lef Yalve (Rated Power) . . . . . 184 L
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ANF-87-161 Supplement 1 ,
Page 1
1.0 INTRODUCTION
This report documents the results of Standard Review Plan (SRP)(I) Chapter 15 event analyses performed in support of Hillstone Point Unit 2 Cycle 10 operation with up to 23.5% steam generator tube plugging and the first reload of Advanced Nuclear Fuels Corporation (ANF) fuel. The Chapter 15 events were selected in accordance with ANF methodology.(2) The basis for event selection is documented in Reference 3, the Disposition of Events report. The LOCA/ECCS analyses are documented in References 18 and 19.
Section 2.0 presents a summary of the results and review of SRP Chapter 15 events. Section 3.0 presents the conditions employed in the event analyses and the results of these event analyses. Events are numbered in accordance with the SRP to facilitate review. A tabular list of the disposition of events and analysis of record for Millstone Unit 2. Chapter 15 events, with a cross reference between SRP ev'ent numbers and the Hillstone Unit 2 Updated FSAR(5) is included.
4 4
ANF 87-161 Supplement 1 Page 2 2.0 SUMARY AND CrACLUSIONS A summary Disposition of Events for Hillstone Unit 2 Cycle 10 is given in Table 2-1. This table lists each SRP Chapter 15 event, indicates whether that event is reanalyzed for this submittal, and provides a reference to the bounding event or analysis of record for events not reanalyzed. The Disposition of Events is reported in greater detail in Referene.e 3.
The analyses considered the following revisions to acceptable plant operating conditions:
(1) Incret. sed maximum radial peaking factor.
(2) Extension of the cycle length to 18 months.
(a) Increased both positive and negative bounds on the moderator temperature coefficient.
(b) Increased shutdown margin requirement.
(3) Operation over 6 full power inlet temperature range from 537 'F to 549 'F. Greater temperature reduction is acceptable concurrent with reduced power and pressurizer level during an end of-cycle coastdown.
The results of Anticipated Operational Occurrences and Postulated Accidents reanalyzed for this submittal are listed in Table 2-2. Acceptance criteria are met for each event.
The results reported herein confirm that event acceptance criteria, defined in Section 15.0.1.1 of this document, are met for plant operation as defined by the operating parameter ranges in Sections 15.0.1 - 15.0.11 of this report.
These results support operation with up to 23.5% average steam generator tube plugging at a rated thermal power of 2700 MWt.
Table 2-1 Disposition of Events Summary SRP Event Event Bounding Updated Classifi- Desig- Event or FSAR cation nation Name Disposition Reference Designation 15.1 INCREASE IN llEAT REMOVAL BY Tile. SECONDARY SYSTEM 15.1.1 Decrease in Feedwater Temperature Bounded 15.1.3 14.1.4, 14.11.2 15.1.2 Increase in Feedwater Flow
- 1) Power Bounded 15.1.3 14.1.4, 14.8, 14.11.2
- 2) Startup Bounded 15.1.3 15.1.3 Increase in Steam Flow Analyze 14.8, 14.11.1 15.1.4 Inadvertent Opening of a Steam Generator Relief of Safety Valve
- 1) Power Bounded 15.1.3 14.3, 14.11.1
- 2) Scram Shutdcwn Margin Bounded 15.1.3 15.1.5 Steam System Pipin; Failures Inside and Outside of Containment Analyze 14.1.4, 14.12 15.2 DECREASE IN HEAT REMOVAL BY Tile SEC0iDARY SYSTEM 15.2.1 Loss of External Load Analyze 14.9 15.2.2 Turb.ne Trip Bounded 15.2.1 14.9 .
15.2.3 Loss of Condenser Vacuum Not in Licensing Basis %
$5%
15.2.4 Closure of the Main Steam
- E a, Isolation Valves (MSIVs) Analyze 14.8 "37
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15.2.5 Steam Pressure Regulator Failure Not applicable;
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) BWR Event
Tcblo 2-1 Disposition of Events Summary (Cont.)
SRP Event .
Event Bounding Updated Classift- Desig- Event or FSAR cation nation Em!!t Disposition Reference Desienation 15.2.6 Loss of Non-emergency A.C. Power Not in Licensing Basis to the Station Auxiliaries 15.2.7 Loss of Normal feedwater Flow Analyze 14.8,14.10,14.A ~
15.2.8 Feedwater System Pipe Breaks Not in Licensing Basis Inside and Outside Containment 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW 15.3.1 Loss of forced Reactor Coolant Flow Analyze 14.6 15.3.2 Flow Controller Nalfunction Not Applicable 15.3.3 Reactor Coolant Pump Rotor Seizure Analyze 14.6 15.3.4 Reactor Coolant Pump Shaft Break Bounded 15.3.3 14.6 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.1 Uncontrolled Control Rod Back Withdrawal from a Subcritical or Low Power Condition Analyze 14.2 15.4.2 Uncontrolled Control Rod Bank =8$
Withdrawal at Power Operation Conditions Analyze !4.2 E$[
e~ .i R l-. .i 15.4.3 Control Rod Misoperation AO
. 1) Dropped Control Bank / Rod Analyze' 14.4 ,,
- 2) Dropped Part-Length Control Rod Not Applicable 14.4
Table 2-1 Disposition of Events Sumnary (Cont.)
, SRP j Event Event Bounding Updated Classifi- Desig- Event or FSAR cation nation H3!!!t Disposition Reference Desienation
- 3) Malpositioning of the Part-Length Control Group Not Applicable 14.4.5
- 4) Statically Misaligned Control Rod / Bank Not in Licensing Basis
- 5) Single Control Rod Withdrawal Analyze 14.2
- 6) Reactivity Control Device Not Applicable 14.1.4 Removal Error During Refueling
- 7) Variations in Reactivity Load to be Compensated by Burnup or On-Line Refuelimj Not Applicable 14.1.4 15.4.4 Startup of an Inactive Loop Not Applicable 14.7 -
15.4.5 Flow Controller Malfunction Not applicable; No Flow Con-troller 15.4.6 CVCS Malfunction that Results Analyze 14.3 in a Decrease in the Boron Con- Modes 1-6 centration in the Reactor Coolant 15.4.7 Inaovertent Loading and Operation 2EE of a Fuel Assembly in an Improper %E7 Position Not in Licensing Basis wTS 2.L E E:
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Tabla 2-1 Disposition cf Events Sumary (Cont.)
SRP Event Bounding Updated Event Classiff- Desig- Eventfr FSAR cation nation Name Disposition Reference Desianation 15.4.8 Spectrum of Control Rod Ejectiec Analyze 14.13 Accidents ;
15.4.9 Spectrum of Rod Drop Accidents Not applicable; (BWR) BWR Event 15.5 INCREASES IN REACTOR COOLANT INVENTORY 15.5.1 Ir: advertent Operation of the Not in Licensing Basis ECCS that Increases Reactor Coolant Inventory 15.5.2 CVCS Malfunction that In- Not in Licensing Basis creases Reactor Coolant Inventory 15.6 DECREASES IN REACTOR COOLANT INVENTORY 15.6.1 Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve Analyze 14.5
. 15.6.2 Radiological Consequences of the flot applicable 14.1.4 Failure of Small Lines Carrying Primary Coolant Outside of Containment 15.6.3 Radiological Consequences of Bounded 14.14 14.14 Steam Generator Tube Failure y $.
ro u 15.6.4 Radiological Consequences of a Not applicable; cn 5" $
Main Steamline Failure Outside BWR Event 2L Containment EO
Table 2-1 Disposition of Events Summary (Cont.)
SRP Event Event Bounding Updated Classifi- Desig- Event or FSAR cation nation Name Disposition Reference Desionation 15.6.5 Loss of Coolant Accidents Analyze 14.15.3, Resulting from a Spectrum of 14.15.4 Postulated Piping Breaks within the Reactor Coolant Pressure Boundary 15.7 RdDI0 ACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7.1 Waste Gas System Failure Bounded 14.17 14.17 15.7.2 Radioactive Liquid Waste System Leak or failure (Release to Atmosphere) Not in Licensing Basis 15.7.3 Postulated Radioactive Releases Not in Licensing Basis due to Liquid-Containing Tank Failures 15.7.4 Radiological Consequences of Fuel Bounded 14.19 14.19 llandling Accidents 15.7.5 Spent Fuel Cask Drop Accidents Not in Licensing Basis
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Tablo 2-1 Disposition cf Events Sususary (Cont.)
SRP Event Event Bounding Updated Classifi- Desig- Event or FSAR cation nation !Last Di sposi ti_oi.t Reference Desianation FSAR EVENTS NOT CONTAINED IN THE STANDARD REVIEW PLAN (1) Effects of External Events Bounded 14.21 14.21 (2) Failures of Equipment Which Bounded 14.1.4 14.1.4 Provide Join'. Control / Safety functions (3) Containment Pressure Analysis Bounded 14.16 14.16 (4) Hydrogen Accumulation in Bounded 14.18 14.18 Containment (5) Radiological Consequences of Bounded 14.20 14.20 the Design Basis Incident (DBI) l m
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Table 2-2 Summary of Results Maximum Maximum Pressurizer MDNBR Anticipated Operational Occurrence LHR (kW/ft) Pressure (psia) (XNB) 15.1.3 Increase in Steam flow (4) 19.3 2200. 1.21 15.2.1 Loss of External Load 17.6 2604.(2) 3,39(1) 15.2.4 Single MSIV Closure 20.9 (11) >l.19(12) 15.2.7 Loss of flormal feedwater (5) (11) (5) 15.3.1 Loss of Forced Reactor Coolant flow 17.2 2218. >l.19(3) 15.4.1 Uncontrolled Control Bank '
Withdrawal at low Power 20.1 2304. 1.55 15.4.2 Uncontrolled Control Bank Withdrawal at Power 19.1 2378.(13) 1.21 15.4.3 Control Rod Misoperation
- Dropped Rod or Bank 17.7 2220. >l.19(6) 15.4.6 CVCS Malfunction resulting in Decreased Boron Concentration (Adequacy of Shutdown Margin is Demonstrated) 15.6.1 Inadvertent Opening of a Pressurizer Pressure Rellet Valve 18.3 2200. 1.20{4) , , ,,
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Tablo 2-2 Summary cf Res:lts (Cont.)
Maximum Maximum Pressurizer MDNBR Postulated Accident LHR (kW/ft) Pressure fosia) (XNB) 15.3.3 Reactor Coolant Pump Rotor Seizure 17.4 2325. >1.19 I7I 15.4.3 Control Rod Misoperation
- Single Rod Withdrawal 19.1 2379. 0.98 I9) 15.4.8 Control Rod Ejection (8) 2671. 0.81 IIOI (1) MDNBR case (2) Peak pressure case. Maximus RCS pressure for Event 15.2.1 is 2697 psia.
(3) DNB LCO was statistically shown to protect against penetration of MDNBR limits for this event. Deterministic MDNBR is 0.98.
(4) Rated power case (5) 8ounded by Event 15.3.1 (6) DNB LC0 was statistically shown to protect against penetration of MDNBR limits for this event. Deterministic MDNBR is 1.14 (7) DNB LCO was statistically shown to protect against penetration of MDNBR limits for Event 15.3.1. Deterministic MDNBR is 0.96.
(8) Deposited enthalpy is less than 280 cal /g (9) Fuel failure bounded by Control Rod Ejection event (15.4.8)
(10) Less than 11.5% of the core cladding fails (11) Bounded by Event 15.2.1 (12) DNB LCO was statistically shown to protect against penetration of MDNBR limits for Event-15.3.1. Deterministic MDNBR is 1.01. 2? E?3; (13) Maximum pressure was determined over the range of analysis ins (rtion rates. 7$ p,}[
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ANF-87 161 Supplement 1 Page 11 3.0 A ELYSIS OF PLANT TRANSIENTS This section provides the results of event analyses performed to support Millstone Unit 2 operation with ANF reload fuel. Event numbering and nomenclature are consistent with the SRP to facilitate review.
This section also provides information on the plant licensing basis as it affects the event analyses, including the classification of plav :onditions, event acceptance criteria, and single failure criteria. Plant operating mode and analysis initial conditions are listed. Neutronics data and core and fuel design parameters are provided. Listings of systems and components available for accident mitigation, trip setpoints, time delays and component capacities are also includcd. These data, together with the plant design parameters and the event specific input data, rapresent a summary of analysis inputs.
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ANF 87-161 Supplement 1-Page 12 15.0 ACCIDENT ANALYSES 15.0.1 (L&SS!FICATION OF PLANT CONDITIONS Plant operations are placed in one of four categories. These categories are those adopted by the American Nuclear Society (ANS). The categories are:
NORMAL OPERATION AND OPERATIONAL TRANSIENT
. Events which are expected to oscur frequent 1/ in the course of power operation, refueling, maintenance, or plant mineuvering.
FAULTS OF MODERATE FREQUENCY
. Events which are expected to occur on a frequency of once per year during plant operation. ,
INFREQUENT FAULTS
. Events which are expected to occur once during the lifetime of the plant.
LIMITING FAULTS
. Events which are not expected to occur but which are evaluated to demonstrate the adequacy of the design.
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l ANF-87-161 Supplement 1 Page 13 l
15.0.1.1 Accentance criteria l
Operational Events This condition describes the normal operational modes of the reactor. As such, occurrences in this category must maintain margin between operating conditions and the plant setpoints. The setpoints are established to assure maintenance of margin to design limits. The set of operating conditions, together with conservative allowances for uncertainties establish the set of-initial conditions for the other event categories.
Mod 6 rate Frecuency Events
- 1. The pressures in reactor coolant and main steam systems should be less than 110% of design values.
- 2. The fuel cladding integrity should be maintained by en3uring that fuel design limits are not exceeded. That is, the minimum calculated departure from nucleate boiling ratio is not less than the applicable limits of the DNBR correlation being used.
- 3 . The radiological consequentes should be less than 10 CFR 20 guideliner.
- 4. The event should not generate a more serious plant condition without other faults occurring independently.
Infrecuent Events
- 1. The pressures in reactor coolant and main steam systems should be less than 110% of design values.
- 2. A small fraction of fuel failures may occur, but these failures should not hinder the capability of the core to be cooled.
- 3. The radiological consequences should be a small fraction of 10 CFR 100 guidelines (generally <10%).
- 4. The event should not generate a limiting fault or result in the consequential loss of the reactor coolant or containment barriers,
l ANF-87 161 Supplement 1
- Page 14 Limitina Fault Evtatt
- 1. Radiological consequences should not exceed 10 CFR 100 guidelines.
- 2. The event should not cause a consequential loss of the required functions of systems needed to cope with the reactor coolant and containment systems transients.
- 3. Additional criteria to be satisfied by specific events are:
- a. LOCA - 10 CFR 50.46 and Appendix K.
- b. Rod Ejection - Radially 1.veraged fuel enthalpy <280 cal /gm.
15.0.1.2 classification of Accident Events By Catecorv i
Table 15.0.1-1 lists the accident category used for each event analyzed in
- this report. This classification is used in evaluating the acceptability of the results obtained from the analysis.
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ANF 87 161 Supplement 1 Page 15 Table 15.0.1-1 Accident Category Used for Each Analyzed Event Event Accident Cateoorv j 15.1.3 Increase in Sten 1 Flow Moderate 15.2.1 Loss of External Load Moderate 15.2.4 Single MSIV Closure Hoderate ;
15.2.7 Loss of Normal Feedwates Flow Moderate 15.3.1 Loss of Forced Reactor Coolant Flow hoderate >
13.3.3 Reactor Coolant Pump Rotor Seizure Limiting Fault 15.4.1 Uncontro11*4 Bank Withdrawal at i Suberitical or low Power Moderate j i 15.4.2 Uncontrolled Bank Withdrawal at Power Moderate 15.4.3 Control Rod Hisoperation
- 1) Dropped Control Rod / Bank Moderate
- 5) Single Coe. trol Rod Withdrawal Infrequent l 15.4.6 CVCS Halfunction Resulting in l Decreased Boron Cencentration Moderate ;
1 15.4.8 Control Rod Ejection Limiting Fault '
l 15.6.1 Inadvertent Opening of a Pressurizer l Pressure Relief Valve Moderate f i
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ANT 87 161 Supplement 1 Page 16 15.0.2 PLANT CHARACTERISTICS AND INITI AL Cy"d(LQfQ Six operational modes have been consid m d in the analysis and are characterized as follows:
REACTIVITY % RATED . AVERAGE COOLANT THERMAL POWER TEMPERATURE MODE CONDITION. K,ff
- 1. Power Operation 1 0.99 > 5% 1 300 'F
- 2. Startup ?,0.99 i 5% 1 300 'F
- 3. Hot Standby < 0.99 0 2 300 'F
- 4. Hot Shutdown < 0.99 0 300 'F 'F
> T"V9
> 200
- 5. Cold Shutdown < 0.98 0 1200 'F
- 6. Refueling s 0.95 0 s 140 'F
- Excluding decay heat.
These operational modes have been considered in establishing the subevents associated with each event initiator. A set of initial conditions is established for the events analyzed that is consistent with the conditions for each mode of operation.
The nominal plant rated operating conditions are presented in Table 15.0.21 and principal fuel design characteristics in Table 15.0.2 2. The uncertainties used in the accident analysis applicable to the operating conditions are:
Core Power 2%
Primary Coolant Temperature 1 2 'T Primary Coolant Pressure 1 22 psi Primary Coolant Flow 2 4 *.
ANF-87 161 Supplement 1 Page 17 Table 15.0.2-1 Plant Operating Conditions Core Thermal Power 2700 MWt Pump Thermal Power (total) 17,1 MWt System Pressure 2250 psia RCS Flow Rate 340,000 gpm Average Coolant Temperature 576.3 'F Core Inlet Coolant Temperature 549.0 'F Steam Generator Pressure 870 psia Steam Flow Rate 11.95 Mlbm/hr Feedwater Temperature 435 'F Steam Generator Tubes Plugged 23.5 %
ANF-87-161 ,
Supplement 1
- Page 18 Table 15.0.2 2 Nominal Fuel Design Parameters Total Number of Fuel Assemblies 217 ;
Fuel Assembly Design Type 14x14 i Assembly Pitch 8.180 in. ;
Fueled Rods per Assembly 176 Guide Tubes per Assembly -
4 Instrument Tubes oer Assembly 1 f
Rod Pitch 0.580 in.
Clad Outside Diameter 0,440 in. !
Guide and Instrument Tube OD (above dashpot) 1.115 in.
Active Fuel Length 136.70 in.
Fuel Rod Length 146.484 in. l Number of Spacers 9 l<
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- ANF 87-161 Supplement 1 Page 19 '
15.0.3 POWER DISTRIBUTION The radial and axial power peaking factors used in the analysis are presented ,
in Table 15.0.3 1. Tho Technical Specification (I) Limiting Conditions of (
Operation assure that the power distribution is maintained within these limits j during normal operation. However, some events analynd result in transient l redistribution of the radial power peaking factors. Transient radial power [
redistribution is' treated as described in Section 15.4.3.2 of Reference 2. [
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ANF-87-161 Supplement 1 Page 20
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t Table 15.0.3 l Core Power Distribution Radial Peaking Factor: Figure 15.0.3-1 Axial Peaking Factor:
- 100% power 1.70
- 50% power 2.24 Fraction of Power Deposited in the Fuel .975 t
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ANF-87 161 Supplement 1 Page 22 l 15.0.4 EANGE 07 PLANT OPERATING PARAMETERS AND STATES l
Table 15.0.4-1 presents the range of key plant operating parameters considered in the analysis. A broader range of power, core inlet temperature, and primary pressure is considered in establishing the trip setpoints verified by the analysis results presented in this document.
The range of operating states of the reactor is also considered in the analysis. The effect of exposure on fuel thermal performance and neutronics parameters is considered. State values are selected for the event analyzed to provide the greatest challenge to the acceptance criteria for that event.
Severci calculations may be required to bound the range of the state variable.
For example, a range of neutronic parameters is used in the analysis of rod withdrawal events in order to verify the range of protection of the challenged trip setpoir.ts.
The range of initiating events is also considered in formulating the analysis conditions f,or an event. The initiating conditions are examined to identify a set which conservatively challenge; the acceptance criteria. Where not obvious, sensitivity studies are performed. For example, analyses are performed for uncontrolled rod withdrawal events throughout the range of reactivity insertion rate possible from boron dilution to maximum withdrawal rate of the highest worth control banks.
' ANF 87-161 Supplement 1 Page 23 Table 15.0.4 1 Range of Key Initial Condition Operating Parameters Core thermal power Subcritical to 2754 MWt(I)
Core inlet temperature Programed 12'F II)
(Poweroperation)
Reactor coolant system pressure 2250 psia 1 50 psi (3)
Fressurizer water level Programed i 7.5 in.
Feedwater flow and temperature Range consistent with power level i
r 5
(1) 102% of 2700 MWt l (2) Analysis supports operation over a full power inlet temperature range
- from 537 *F to 549 'F (3) 122 psi measurement uncertainty and 28 psi operating range f
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ANF 87-161 Supplement 1 Page 24 15.0.5 REACTIVITY COEFFICIENTS USED IN THE SAFETY ANALYSIS Table 15.0.51 presents the reactivity coefficients used in the analysis.
The set of parameters used in each analysis is listed in the appropriate Section for that event, l
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- ANF-87-161 Supplement 1 Page 25 Table 15.0.5 1 Reactivity Parameters 111.1 BOC Bound,ing EOC Boundina Moderator Temp Coef, 10'4 Ap/'F 0.4 2.8 Moderator Pres Coef, 10 6 Ap/ psi -1.4 5.0 Doppler Temp Coef, 10-5 3,f.F 1.0 2.0 Delayed Neutron Fraction 0.0071 0.0045 Effective Neutron Lifetime, :
10 6 seconds 19.0 22.0 238 U Atoms Consumed per Total Atoms Fissioned 0.49 0.79 l
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- MTC is +0.7 x 10'4 ap/'F for power levels less than 70% of rated.
f
ANF 87 161 Supplement 1
- Page 26 I
i 15.0.6 SCRAM INSERTION CHARACTERISTICS.
Scram reactivity insertion as a function of ASI wu used in the analysis for reactor trip. The ii.sertion worth includes the most reactive rod stuck out.
The shutdown margin of 3.6% As and a control rod drop time of 2.75 seconds (to 90% insertion) have been sLpported by the transient anaiyst,.
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15.0.7 . TRIP SETPOINT-VERIFICAJ[QH Operating limits for the Millstone Unit 2 nuclear plant are summarized below.
Methods of analysis for determining or verifying the operating limits are detailed in Subsection 15.0.7.5 and Reference 4. Axial power distributions
.and other core neutronics related parameters used in the setpoint verification l analyses were generated. with ANF's approved core simulator code XTGPWR(6) ,
This data was generated on a three dimensional core basis, as described in References 12. and 20. With this methodology, the values 'of F g used in the 7
setpoint verificai. ion are calculated directly with a three-dimensional model and since operation within the Technical Specification on F r limits F q, the-need for an F xy limit is eliminated.
Results of the ANF satpoint. analyses indicate that operating limits previously established for Millstone Unit 2 are adequate for Cycle 10 operation with ANF 4 reload fuel present in the core.
15.0.7.1 Reactor Protection System The reactor protection system (RPS) is designed to assure that the reactor is operated in a safe and conservative manner. The input parameters for the RPS are denoted as limiting safety system settings (LSSS). The values or functional representation of the LSSSs are calculated to ensure adherence to the specified acceptable fuel design limits (SAFDLs) during steady state and anticipated operational occurrences (A00s). The safe operation of the reactor
, is also maintained by restricting reactor operation to conform with the limiting conditions for operation (LCOs), which are administrative 1y applied at the reacto plant. The LSSS and LCO parametric values are presented in the following seccions.
I
f ANF-87-161 t Supplement 1 Page 28 15.0.7.2 Soecified Accentable Fuel Desian Limits (SAFDLs)
The SAFDLs are limits on the fuel and claddir.g established in order to preclude fuel failure. These limits may not be exceeded during steady state operation or during A00s. The SAFDLs are used to establish the reactor setpoints to ensure safe operation of the reactor. The specific SAFDLs used to establish the setpoints are:
~1). The local power density (LPD) which coincides with fuel centerline melt.
2). The MDNBR corresponding to the accepted criterion which protects against the occurrence of DNB.
The LPD limit for Hillstone Unit 2 has been 21 kW/ft in prior cycles and this limit is being retained for Cycle 10 It is noted that reload fuel for Cycle 10 contains gadolinia-bearing fuel rods which, for a given LPD, will operate with a higher fuel temperature and will consequently have a lower LPD limit.
The neutronics design of the gadolinia-bearing fuel rods is such that the maximum LPD in the gadolinia-bearing fuel rod with a standard fuel rod at the 21 kW/ft limit will be sufficiently below 21 kW/ft to prevent centerline melt.
Therefore, the gadolinia-bearing fuel would not bec'ome limiting and the 21 j- kW/ft design limit would remain applicable.
The XNB critical heat flux correlation (8) was used in the thermal margin anal) ris with statistical parameters corresponding to an upper 95/95 value of 1.17 which is conservative relative to the 95/95 limit for XNB. Observance of the limiting conditions for operation will protect against DNB with 95%
probability at a 95% confidence level during an A00. The XNB correlation was genlified for application to 14x14 fuel for CE reactors in Reference 9.
i The Cycle 10 loading of the Millstone Unit 2 plant is composed of fuel from 1
ANF-87-161 Supplement 1 Page 29 two~ different fuel vendors. To account for any loss in DNB performance which -
may occur in Cycle 10 as a result of mixing different fuel types, the ANF mixed . core methodology (10) was used. This effectively elevates the limiting XNB critical heat flux correlation value for Cycle 10.
15.0.7.3 Limitino Safety System Settinas 15.0.7.3.1 Local Power Distribution The local power distribution (LPD) trip limit is the locus of the limiting values of core pos:ar level versus axial shape index that will produce a reactor trip to prevent exceeding the 21 kW/ft LPD limit. The correlation between allowed core power level and peripheral axial shape index (ASI) was determined using methods which take into account the total calculated nuclear peaking and the measurement and calculational uncertainties associated with power peaking. The LPD barn for operation at 2700 MWt is shown in Figure 15.0.7-1 as a locus of power and AS.I pairs which is conservatively bounded by the calculated power and ASI pairs. In this figure ASI is defined as the uifference between the core power in the bottom half of the core and the top half divided by the sum of the top and bottom halves.
15.0.7.3.2 Thermal Marcin/, Low Pressure The thermal margin / low pressure (TM/LP) trip protects against the occurrence of DNB during steady state operations and for many, but not all, A00s. This reactor trip system mon', tors primary system pressure, core inlet temperature, core power and ASI. A reactor trip occurs when primary system pressure falls below the computed limiting core pressure, P*/ ar. A statistical setpoint methodology (4) is used to verify the adequacy of the existing TM/LP trip for Cycle 10. The methodology for the TM/LP trip accounts for uncertainties in l core operating conditions, XNB DNB correlation uncertainties, and l
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ANF-87-161 Supplement 1 Page 30 uncertainties in power peaking. The existing TM/LP trip function for operation at 2700 MWt is given by:
P var - 2215 x A1 (ASI) x QR1 (Q) + 14.28 x Tin - 8240, where Q is the higher of the thermal power and the nuclear flux power, Tin is the inlet temperature in 'F and Al and QR1 are shown in Figures 15.0.7-2 and 15.0.7-3, respectively.
15.0.7.3.3 Additional Trio Functions In addition to the LPD and TM/LP trip functions, other reactor system trips have been determined to provide adherence to reactor system design criteria.
The setpoints for these trips, shown in Table 15.0.7-1, are unchanged from the Cycle 9 values.
15.0.7.4 Limitino Conditions for Ooeration 15.0.7.4.1 DNB Monitorina The validity of the existing 100 for allowable core power as a function of ASI was verified to ensure adherence to the SAFDL on DNB during postulated CEA drop and loss-of-flow operational occurrences. The statistical analysis accounted for the effects of uncertainties associated with incore operating parameters, the XNB critical heat flux correlation, and power peaking. The <
allowed core power as a function of ASI for the existing LCO is shown to be conservatively bounded by the present analysis in Figures 15.0.7-4 and 15.0.7-5.
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d ANF-87-161' Suppim2nt. 1 iage 31 1
- 15.0.7.4.2 linear Heat Rate Monitorina ~
In the event that the in-core detector system is not in operation for an' ,.
extended period of time, the linear heat rate will be monitored through the S
-use of an LPD LCO. The verification of this LCO was performed in a fashion similar to that used in verifying the LPD limiting safety system setting (Section 15.0.7.3.1). The verification plot is shown in Figure 15.0.7-6. The LPD LC0 limits core power so that the linear heat rate (LHR) LCO based on loss-of-coolant (LOCA) considerations is not exceeded. The LHR heat rate LCO protected by the LPD LCO verified for Cycle 10 is depicted in Figure 15.0.7-7.
15.0.7.5 Setooint Analysis 15.0.7.5.1 Limitina Safety System Settinas local Power Distribution The local power distribution (LPD) trip monitors core power and ASI in order to initiate a reactor scram which precludes exceeding fuel centerline melt conditions. In the analysis for this trip function a large number of axial power distribution cases typical of the cycle were examined to establish bounding values of total power peaking, Fq, versus ASI. These cases were generated in a manner consistent with that discussed in Reference 4.
Statistical methods were then employed to account for the uncertainties in the parameters that are given in Table 15.0.7-2.
The peak linear heat rate in the core occurs at the position of the maximum total peaking factor, F g, which is the ratio of the maximum linear heat ,
generation rate in the core to the average linear heat generation rate in the core. F g is the product of the peak core linear heat rate, as determined by the ANF three-dimensional neutronics methodology, and the axial augmentation
ip ,
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ANF-87-161 5]'Mo
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Supplement 1 Page 32
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sa- factor depicted'in Figure 15.0.7-8, which.is an adjustment to. account for any M ,,
.-flux peaking and fuel densification which may occur.
'j; The allowed power for each ASI was calculated statistically for the Cycle 10 core incorporating the uncertainties listed in Table 15.0.7-2 as described in Reference 4. The results in Figure 15.0.7-1 bound the existing Millstone Unit 2 LPD trip and thus verify the adequacy of the existing trip function.
Thermal Marcin/ Low Pressure LSSS The TM/LP trip -is designed to shut the reactor down should the reactor conditions (ASI, inlet temperature, core power and pressure) approach the point where ONB might occur during either normal operation or an A00. The present analysis uses the XNB critical heat flux correlation and the statistical setpoint methodology described in Reference 4. The analysis methodology is consistent with the NRC's Standard Review Plan in requiring DNB to be avoided with 95% probability at a 95% confidence level.
'The uncertainties shown in Tables 15.0.7-2 and 15.0.7-3 are included in the verification of the TM/LP trip as described in Reference 4. An excess margin of protection is provided by the existing trip for Cycle 10.
15.0.7.5.2 Limitina Conditions for Ooeration DNB Monitorina The TM/LP trip system does not monitor reactor coolant flow or consider
. changes in power peaking that insignificantly change ASI. Thus, the TM/LP trip generally does not provide DNB protection for the four pump coastdown and CEA drop A00s. The analyses of these transients are given in Sections 15.3.1
4 ANF-87-161 Supplement 1 Page 33 and 15.4.3. The LC0 presented here administratively protects the DNB SAFDL for these transients.
The method used to establish the DNB LCO involved simulations of the CEA drop e and the loss-of-flow transients using the core thermal hydraulic code XCOBRA-IIIC(11) to determine ~ the initial power, as a function of. ASI, which provides protection from DNB with 95% probability. The uncertainties listed in Tables 15.0.7-4 and 15.0.7-5 are applied using the methodology described in Reference 4 The results of the statistical inalyses for the CEA drop and the loss-of-flow transient are summarized by the points in Figures 15.0.7-4 and 15.0.7-5 respectively. The points bound and, thus, verify the adequacy of the existing DNB LCO for Millstone Unit 2, which is shown in these same figures by the striight line segments. .
LPD Monitorina The plant Technical Specifications allow plant operations for limited periods of time with the in core detectors out of service. In this situation, the LPD barn provides protection in steady-state operation against penetration of the LPD limit established by LOCA considerations. The statistical methodology for the LPD LCO is essentially the same as that for LPD LSSS except:
- 1) The peak LPD limit is reduced and may be a function of axial elevation, and
- 2) The uncertainties listed in Table 15.0.7-4 are used, as opposed to the values in Table 15.0.7-2.
The allowed power versus AS! was statistically analyzed to account for the appropriate uncertainties. The points in Figure 15.0.7 6 represent the statistical calculation of the LHR curve depicted in Figure 15.0.7-7. The LC0 curve is shown by the straight line segments in Figure 15.0.7 6, and
ANF-87-161 Supplement 1 Page 34 conservatively bounds the calculated verification points.
i:
ANF-87-161 Supplement 1 Page 35 Table 15.0.7-1 Trip Setpoints Parameter Set Point Uncertainty Low steam generator pressure 680 psia 22 psi Low steam generator water level 36% 3.7 in Variable high. power 9.6% of rated 5%
above current power
(<107% of rated)
Low reactor coolant flow 91.7% 2%
High pressurizer pressure 2400 psia 22 psi
_ _= _ ____ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _______ __.
i ANF-87-161:
- Supplement 1 Page 36 Table 15.0.7-2. Uncertainties Applied in LSSS Calculations Source Value*
Engineering tolerance 0.03 Peaking uncertainty (%) 8.5 Power:
measurement (%) 2.0 of rated decalibration (%) - 3.0 of rated **
ASI uncertainty i 0.06 i
- The distributions are treated as normal and the uncertainty range represents 2 o values.
- Not treated statistically t $
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ANF-87-161.
Supplement 1 Page 37.
Table 15.0.7-3 Uncertainties Applied in the TM/LP LSSS Calculations Source Value*
Pressure Measurement 22 psi Time delay allowance - 50 psi **-
Inlet coolant temperature 2'F
- XNB correlation 0.198 Flow measurement 0.05 of rated
-* The distributions are treated as normal and the uncertainty range represents 2 o values.
- Not treated statistically m
e ANF-87-161 Supplement 1 Page 38 Table 15.0.7-4 Uncertainties Applied in the LCO Calculations Source Value*
Engineering tolerance 0.03 Peaking uncertainty (%) 8.5 Power measurement (%) 2.0 of rated ASI uncertainty 0.06
- The distributions are treated as normal and the uncertainty range represents 2a.
e _ _ . __ _
s n.
ANF-87-161 Supplement l'
.Page 39 Table 15.0.7-5 Uncertainties Applied in DNB LCO Calculations .
Source Value*
CEA Dron Loss of Flow Pressure measurement 22 psi 22 psi Inlet Coolant temperature 2*F 2*F XNB correlation 0.198** 0.198 ,
Flow measurement 0.05 of rated 1-0.05 of rated Flow Coastdown -- 0.12 seconds **
Trip setpoint -- 0.02 of rated flow '
Coil hold time -- 0.20 seconds Scram rod insertion time -- 0.24 seconds Rod Worth -- 1 0.72% Ap
- The distributtor.s are treated as normal and the uncertainty range represents 12a. ,
- Time for flow to drop from 90% to 80% of rated flow.
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ANF-87-161 Supplement 1 Page 40 ;
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FIGURE 15.0.7-5 VERIFICATION OF DNB LIMITING CONDITION OF tS OPERATION FOR Tile LOSS OF FORCED COOLANT FLOW TRANSIENT
120.0 m
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A FIGURE 15.0.7-6 VERIFICATI0fl 0F LOCAL POWER DENSITY ui - -
LIMITIriG C0fiDITI0f4 0F OPERATI0ft
20 18-UNACCEPTABLE OPERATION . - - _ -
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l FRACTION OF ACTIVE FUEL HEIGHT aS FIGt!RE 15.0.7-7 I LitlEAR !! EAT l<AIE LCO USED 1.1 LPD LCO VERIFICATION
_ _ _ _ _ _ _ _ _ _ . _ - - - - ~
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ANF-87-161 Supplement 1 Page 48 15.0.8 (OMPONENT CAPACITIES AND SETPOINTS Table 15.0.8-1 presents the component setpoints and capacities used in the analysis.
l l
Table 15.0.8-1 Component Capacities and Setpoints Component Setooint Response Time Capacity Turbine main throttle valve NA NA 13,144,680 lbm/hr Turbine stop valve NA 0.020 sec NA Main steam line isolation valves NA 6.0 sec NA Feedwater flow regulating NA NA 7,169,760 lbm/hr/SG valves Pressure safety valves 2500 psia i 3% NA 294,000 lbm/hr/ valve Steam line safety valves 2 0 1000 psia i 3% NA '794,060 lbm/hr/ valve 2 0 1005 2 0 1015 2 0 1025 2 0 1035 ' ' '~
2 0 1045 4 0 1050 Auxiliary feedwater pumps NA 600 sec for Event 300 gpm/MDAFP 15.2.7 600 gpm for SDAFP '
feedwater flow control NA 20.0 ser. ?!A valves Pressurizer relief valves 2400 psia 22 psi 2.0 see 153,000 lbm/hr/ valve l Pressurizer sprays Off - 2300 psia 22 psi NA 375 gpm Full On - 2350 psia i 22 psi Pressurizer proportional Off - 2275 psia i 22 psi NA '300 kW 'g?
heaters full On - 2225 psia i 22 ps' ,
- g 3; oCT Pressurizer backup heaters Off - 2225 psia i 22 psi NA 1200 kW g g og On - 2200 psia i 22 psi ,,,
- g ;.
22
ANF-87-161 Supplement 1 Page 50 I
t 15.0.9 PLANT S"STEMS AND COMPONENTS AVAILABLE FOR MITIGATION OF ACCIDEN" EFFECTS Table 15.0.9-1 is a summary of trip functions, engineered safety features, and other equipment available for mitigation of accident effects. These are listed for all Chapter 15 SRP events. . - ;e detailed listing of available reactor protection for each event in each operating mode is given in Reference 3.
l s
Table 15.0.9-1 Overview of Plant Systems and Equipment Available for Transient and Accident Condittor:s
~
Event Reactor 'irio Functions Other Sianals and Eauinment 15.1 Increase in Heat Removal by the Secondary System
~
Feedwater System High Power Trip SteamGenerato'rWaterLeveiSignals Malfunctions Thermal Margin / Low Pressure Trip reedwater Isolation Valves Low Steam Generator Pressure Trip Main Steamline Isolation Valves Safety Injection Actuation Signal Turbine Trip on Reactor Trip Increase in Steam Flow Low Steam Generator Pressure Trip Steam Generator Water Level Signals Thermal Margin / Low Pressure Trip Main Steamline Isolation Valves High Power Trip Turbine Trip on Reactor Trip Safety Injection Actuation Signal Atmospheric Steam Dump Controller.
Steam Bypass to Condenser Controller Auxiliary Feedwater System l Inadvertent Opening of a low Steam Generator Pressure Trip Steam Generator Water Level Signals ,
l Steam Generator Relief or Thermal Margin / Low Pressure Trip Main Steamline. Isolation Valves '
Safety Valve High Power Trip Turbine Trip.on Reactor Trip Safety Ir.jection Actuation Signal Atmospheric Steam Dump Controller Steam Bypass to Condenser Controller Auxiliary Feedwater System St'eam System Piping Low Steam Generator Pressure Trip Steam Generator Water Level Signals Failure Thermal Margin / Low Pressure Trip Main Steamline Isolation Valves High Power Trip Turbine Trip on. Reactor Trip Safety Injection Actuation Signal Atmospheric Steam Dump Controller High Containment Pressure Trip Steam 8ypass to Condenser Controller Auxiliary Feedwater System Containment Spray Containment Isolation &
Containment Air Coolers 1R AC >
a; 2 % '
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t____._________._. _ . . . - . _ _ _ _ _ _ _ _ _ _ _ __. , , , ... _-
Table 15.0.9-1 Overview of Plant Systems and Equipment Available for Transient and Accident Conditions (Cont.)
Event Reactor Trio Functions Other Signals and Eculoment 15.2 Decrease in Heat Removal .
by the Secondary System Loss of External Load / High Pressurizer Pressure Trip Steam Generator Water Level Signals Turbine Trip / Loss of High Power Trip Turbine Trip on Reactor Trip Condenser Vacuum Thermal Margin / Low Pressure Trip Atmospheric Steam Dump Controller Low Steam Generator Water Level Steam Bypass to Condenser Controller Trip Steam Generator Safety Valves Pressurizer Safety Valves Pressurizer Sprays Loss of Nonenergency AC Low Reactor Coolant flow Trip Steam Generator Water Level Signals Power to the Station High Pressurizer Pressure Trip Stccm Generator Safety Valves Auxiliaries Thermal Margin / Low Pressure Trip Pressurizer Safety Valves o Lew Steam Generator Water Level Auxiliary Feedwater System Trip Loss of Normal feedwater low Steam Generator Water Level Steam Generator Water Level Signals Flow Trip Steam Generator Safety Valves High Pressurizer Pressure Trip Pressurizer Safety Valves Thermal Margin / Low Pres: are Trip Auxiliary Feedwater System Pressurizer Sprays and Level Control feedwater System Pipe High Pressurizer Pressure Trip Steam Generator Water Level Signals 8?
Break Thermal Margin / Low Pressure Trip Steam Generator Safety Valves l'j; Low Steam Generator Water Level Pressurizer Safety Valves ,, g' 7' Trip Auxili ry Feedwater System 3; 2 23 Low Steam Generator Pressure Trip Pressurezer Sprays and Level Control
- A 2.
1 Table 15.0.9-1 Overview of Plant Systems and Eouipment Available for Transient and Accident Conditions (Cont.) ,
Event Reactor Trio Functions Other Sianals and Ecutoment 15.3 Decrease in Reactor Coolant System Flow Rate Loss of Forced Reactor Low Reactor Coolant Flow Trip At' aspheric Steam Dump Controller Coolant Flow Thermal Ma gin / Low Pressure Trip Steer Bypass to Condenser Controller High Pressurizer Pressure Trip Steam Generator Safety Valves Pressurizer Safety Valves -
Reactor Coolant Pump Low Reactor Ceolant Flow Trip Atmospheric Steam Dump Controller Retor Seizure / Shaft Break High Pressurizer Pressure Trip Steam Bypass to Condenser Controller Steam Generator Safety Valves Pressurizer Safety Valves IS.4 Reactivity and Power Distribution Anomalies Uncontrolled Control Rod Thermal Margin / Low Pressure Trip Rod Withdrawal Prohibit Action on Bank Withdrawal from a High Power Trip Pre-Trip A1+ 2; Subcritical or Low Power High Pressurizer Pressure Trip Startup Condition Uncontrolled Control Rod High Power Trip Pressurizer Safety Valves Bank Withdrawal at Power Thereal Margin / Low Pressure Trip Steam Generator Safety Valves Operation Conditions High Pressurizer Pressure Trip Pressurizer Spray add Level Control Control Rod and Bank Ceviation Alarms which initiate Rod With- %?
drawal Prohibit l'!E
- Rod Withdrawal Prohibit on Pre-Trip ,3 5F 7' Alarms -
3;2"3
- ?. :.
. ... . . . . .. . t". . '2
Table 15.0.9-1 Overview of Plant Systems and Equipment Available for Transient and Accident Conditions (Cont.)
Event Reactor Trio Functions Other Signals and Eauipment Control Rod Misoperation Low Pressurizer Pressure irip Pressurizer Safety Valves Thermal Margin / Low Pressure Trip. Steam Generator Safety Valves Low Steam Generator Water Level Pressurizer Spray and Level Control Trip Control Rod and Bank Deviation Alarms Safety Injection Actuation Signal Startup of an Inactive High Power Trip Administrative Procedures for Loop Thermal Margin / Low Pressure Trip Startup of an Idle Pump Plant Operation with less than all four primary coolant pumps is con-trolled by Technical Specificationr.
Chemical Volume and High Power Trip Administrative Procedures
- Control System (CVCS) Thermal Margin / Low Pressure Trip Sufficient Operator Response Time, Malfunction that Results High Pressurizer Pressure Trip in a Decrease in the Baron Concentration in the Reactor Coolant Inadvertent Loading and (Technical Specification Operation of a fuel Measurement Requirement and Assembly in an Improper Administrative Procedures Position preclude occurrence)
Spectrum of Control Rod High Power Trip Ejection Accidents Thermal Margin / Low Pressure Trip 8?
Long Term, Safety Injection 3 !!
Actuation Signal ,,27 7' JE 2 23 LS l
- w.y 3 Table 15.0.9-1 Overview of Plant Systems and Equipment Available for Transient and Accident Conditions (Cont.)
Event Reactor Trio Functions Other Signals and Eautoment 15.5 Increase in Reactor Coolant Inventory Inadvertent Operation of High Power Trip Pressurir.er Safety Valves the ECCS/CVCS Malfunction Thermal Margin / Low Pressure Trip Overpressurization Mitigation System that incria:es Reactor High Pressurizer Pressure Trip Coolant Inventory 15.6 Decrease in Reactor ' ' '
Coolant Inventory Inadvertent Opening of a Thermal Margin / Low Pressure Trip Charging and Safety Injection System PWR Pressurizer Pressure Pressurizer Heaters Relief Valve Steam Gcnerator Tube Thermal Margin / Low Pressure Trip Steam Generator Safety Valves .
Failure Safety Injection Actuation Signal Main Stea.'!ine_ Isolation Valves (MSIVs)
Atmospheric Stcan Dump Controller
! Steam Bypass to Condenser Controllar Auxiliary feedwa'ar Systes loss of Coolant Accidents No credit taken for a reactor trip Emergency Core Cooling System (ECCS)
Resulting from a Spectrum by the Reactor Protection System Auxiliary Feedwater System of Postulated Piping (RPS) due to the rapid depletion Containment Isolation. _
Bieaks within the Reactor of the moderator which shuts Jown Containment Spray and Air Cooler g' Coolant Pressure Boundary the reactor cora almost immedi- . . . . .
E 3; ately, followed by ECCS injection y;;T T which contains sufficient. boron a g ",
to maintain the reactor core in ' __, .
- 31.
a subcritical configuration. L*
i
. - - ~ , . - . . , . . _ _ , . , - . . - - . , - - - . , . , - . _ , . - _ _ . - - -
ANF 87-161 Supplement 1 Page 56 15.0.10 EFFECTS OF MIXED ASSEMBLY TYPES AND FUEL ROD BOWING In accordance with che NRC Safety Evaluation Report on the ANF mixed core methodology (10) , a penalty is applied in a'il MONBR calculations presented in this report. The penalty is .in addition to the calculated cross flow penalty t obtained by modeling the actual mixed core cror,s flow affects. The impact of this penalty is to effectively increase the XNB correlation limit from the calculated 95/95 linit to 1.19. .
L In accordance with ANF rod bow methodology (I3) , the magnitude of rod bow for the ANF assemblies has been estimated. The calculations indicate that 50%
closure of the rod-to-rod gap occurs at an assembly exposure of about 85.000 mwd /MTU for the ANF 14x14 design. Significant impact to MDNBR due to rod bow ,
dees not occur until the gap clost.res exceed 50%. Since the maximum design l exposure for ANF reload fuel in Millstone Unit 2 is s.ignificantly less than that at which 50% closure occurs, rod bow does not significantly impact the MDNBR for ANF fuel. Also, total peaking is not significantly impacted. [
[
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ANF-87-161 Suppl ~ement 1 Page 57
). 5. 0.11 PLANT LICENSING BASIS'AND SINGLE FAILURE CRITERIA' ,
The licensing basis for Millstone Unit 2 is as stated in the Final Safety [
Analysis Report (5) . The event scenarios depend on single failure criteria f established by the plant licensing basis. Examination of the Millstone Unit 2 licensing basis yields the following single failure criteria: !
(1) The Reactor Protection System (RPS) is designed with redundancy and I- independence to assure that no single failure or removal from ,
service of any component or channel of a system will result in the "
loss of the protr.ction function. !
(2) Each Engineered Safety Feature (ESF) is designed to perform its intended safety function assuming a failure of a single active component.
(3) The onsite power system and the offsite power system are designed such that each shall independently be capable of providing power for the ESF assuming a failure of a single active component in' either !
power system.
The safety analysis is structured to demonstrate that the plant systems design f satisfies these single failt.re criteria. The following assumptions result: [,
i
[
(1) The ESF required to function in an event are assumed to suffer a !
worst single failure of an active comoonent.
(2) Reactor trips occur at the specified setpoint within the specified [
delay time assuming a worst single active failure. [
(3) The foll wing postulated accidents are considered assuming 5 k concurrent loss of offsite power: main steamline break, control rod l ejection and large break LOCA. L I
(4) The loss of normal feedwater, an anticipated operational !
occurrence, is also analyzed assuming a concurrent loss of offsite t 1 power.
ANF-87 161 Supplement 1
. Page 58 The requirements of 10 CFR 50, Appendix A, Criteria 10, 20, 25 and 29 require that the design and operation of the plant and the reactor protective system assure that the Specified Acceptable fuel Design Limits (SAFDLs) not be exceeded during Anticipated Operational Occurrences (A00s). As per the ,
definition of A00 in 10 CFR 50, Appendix A. "Anticipated Operational Occurrences mean those conditions of normal operation which are expected to occur one or more times during the life of the plant and include but are not limited to loss of power to all recirculation pumps, tripping of the turbine generator set, isolation of the main condenser, and loss of all offsite power". The ta7DLs are that: (1) the fuel shall not experience centerline melt (21 kW/t- and (2) the departure from nucleate boiling ratio (DNBR) shall have a minimum allowable limit such that there is a 95% probability with a 957. confidence interval thtt departure from nucleate boiling (DN8) has not occurred.
As indicated in Reference 3, three major revisicns to acceptable plant' operating conditions are planned. The three revisions are:
(1) The maximum Technical Specification radial peaking factor is being increased from the current limit of 1.537 to 1.61.
(2) Plant cycle length is being increased from 12 to 18 month cycles.
(a) The moderator temperature coefficient (MTC) in the Technical Specification is being changed in order to accommodate the increased cycle length. The most positive MTC change for power less than or i equal to 70% is from +5 to +7 pcm/'F. The most positive MTC '3r powet greater than 70% remains at +4 pcm/'F. The most negative MTC i change at full power is from -24 to -28 pcm/'F. l (b) The shutdown margin requirements are being changed to offset the more negative end-of-cycle MTC. The change in MTC results in the ,
l
ANF-87-161 Supplement 1 Page 59 Modes 1, 2, 3 and 4 : shutdown margin going.from the current limit 'of 2 2.9% to 2 3.6%, ,
(3) The analysis will also support plant operation at reduced inlet temperatures. The current nominal inlet temperature is 549'F and the analysis will support up to a 12'F inlet temperature reduction at full power. Greater temperature reduction is acceptable if concurrent with !
reduced power and pressurizer level during an EOC coastdown. I 1
The event analyses were performed to ensure that these revisions to acceptable plant operating conditions are supported, j t l l
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ANF-87-161 Supplement 1 Page 60 15.0.12 PLOT VARIABLE NOMENCLATtJBli.. ;
Plotted results presented in this report employ PTSPWR2@) and SLOTRAX(15) output variable nomenclature. Specific variables plotted are listed and defined in Table 15.0.12-1.
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ANF-87-161
. . Supplement 1 Page 61 i
i Table 15.0.12-1 Nomenclature Used in Plotted Results Variable Name Definition DCLEVAl Steam Generator Downcomer Level, Loop 1 -
DK Total Reactivity :
DKDOP Doppler Reactivity (
DKMDD Moderator Temperature Reactivity '
LEVPR Pressurizer Liquid Level l LEVSGI Steam Generator Liquid Level, Loop 1 PD01 Steam Generator Dome Pressure, Loop 1 PD02 Steam Generator Dome Pressure, Loop 2 l PL Core Power Level j PPR Pressurizer Pressurc, !
PSGSA1 Steam Generator Pressure. Loop 1
{
QOA Core Average Heat Flux ;
- TAVEC Core Average Coolant Temperature TAVG1 Average Coolant Temperature, Loop 1 [
TCIO Core Inlet Coolant Temperature [
TCLAD Average Clad Temperature j TCLI Cold Leg Temperature, Loop 1 ;
TFAVG Average Fuel Temperature f
THL1 Hot leg Temperature Loop 1 l VWPR Pressurizer Liquid Volume ;
WDOSLT Total Steamline Steam Flow Rate [
WFWT Total Feedwater Flow Rate l WLPCR Vessel Flow Rate i
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P ANF-87-161 i Supplement 1 ;
- Page 62 t
15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM i 15.1.3 INCREASE IN STEAM FLOW I
15.1.3.1 Event Descriotion This event is initiated by a failure of the main steam system that results in
. an increase in steam flow from the steam generators. The increased steam flow ;
creates a mismatch between the heat being generated in the core and that being extracted by the steam generators. As a result of this power mismatch, the ;
primary-to secondary heat transfer increases and the primary system cools i down. If the moderator temperature coefficient is negative, the cooldown of l
the primary system coolant would cause an insertion of positive reactivity and the potential erosion of thermal margin.
)
15.1.3.2 Definition of Events Analyzed This event is predominantly a depressurization event. Two cases are analyzed, i The first case is from full power rated conditions. At full power, the margin ,
to limits is the smallest. Therefore, the full power conditions bound operation at lower power levels. The second case is initiated from hot g!
shutdown conditions. For the hot shutdown case, both four pump and one pump !
operation are considered. This case bounds the event consequences for (
transients initiated from refueling shutdown, cold shutdown and refueling t i operation initial conditions. ;
- The end of cycle moderator and Doppler feedback coefficients were selected to I maximize the challenge to the specified fuel design limits for both cases.
The time in the cycle will determine the value of the moderator reactivity j temperature coefficient (HTC). If the HTC is negative, there will be a -
i I
ANF-87-161 -
1 -Supplement 1 -
- Page 63 .
i i
j positive reactivity insertion dependent upon the magnitude of the moderatori eti temperature coefficient. If MTC is positive, then negative reactivity will be '
inserted as the coolant temperatura decreases. Therefore, the consequences of ,
- this event are bounded at end .f-cycle conditions when the modsrator temperature coefficient is at its maximum negative value. !
15.1.3.3 Analysis Results The minimum DNBR for this event initiated from full power occurred for a steam ;
flow increase to about 110%. At this steam flow rate, the TM/LP and the
-c variable high power trips coincide to produce nearly simultaneous trip ;
signals. The junction of these two trips represents the worst possible DNB l conditions. That is, maximum core power is attained combined with a low pressurizer pressure. The calculated deterministic DNBR is 1.21. The peak pellet LHR is 19.3 kW/ft. l
' I .
For the hot shutdown case, the event is initiated by a rapid opening of the atmospheric dump valves or the turbine bypass valves resulting in a steam ;
I flow increa'.e of 41% of the nominal full power steam flow. A bounding value ;
l for the negative moderator temperature coefficient was assumed as was the f Technical Specification value of the shutdown margin. The results of this [
event for both the one pump and four pump case were found to be bounded by the l l full power, full flow event. I l !
l l l The responses of key system variables are given in Figuros 15.1.3 1 to j l
15.1.3-7 for the rated power case. The sequence of events is given in Table ;
15.1.3-1.
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ANF 87-161 l Supplement 1 Page 64 l
1 1
l 15.1.3.4 Conclusion l
The results of the analysis demonstrate that the event acceptance criteria are met since the minimum DNBR predicted for the full power case is greater than the safety limit. The correlation limit assures that with 95% probability and I
95% confidence, DNB is not expected to occur; therefore, no fuel is expected to fail. The fuel centerline melt threshold of 21 kW/ft is not violated !
during this event. l l
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4
ANF 87-161 Supplement'l Page 65 Table 15.1.3-1 Event Summary for Increase in Steam Flow (Excess' Load)
Rated Power Case Event Summary Event Time (sec) 10% Step Increase in Steam Flow 0.00 Peak Core Average Temperature 3.70 Reactnr trip (TM/LP trip) 32.87 ,
Turbi.ie Stop Valve closed 33.00 Peak Power 33.33 Minimum DNBR 33.37 Steam Line Safety Valves open 36.94 Peak Steam Dome Pressure 39.59 i
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- . . - -_ -_ .- . . . _ - - . .. - = _ . - . - . . . . ._ _.
ANF-87 161 Supplement 1 Page 69
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ANF 87-161 Suppleme.it 1 Page 73 15.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2.1 LOSS OF EX(ERNAL LOAD 15.2.1.1 Innt Descriotion A Loss of External Lo'ad event is initiated by either a loss of external electrical load or a turb ne trip. Upon either of these two conditions, the turbine stop valve is assumed to rapidly close (0.1 second). Normally, a reactor trip would occur on a turbine trip; however, to calculate a conservative system response, the reactor trip on turbine trip is disabled.
The steam dump system (atmospheric dump valves - ADVs) is assumed to be unavailable. These assumptions allow the Loss of External Load event to bound the consequences of Event 15.2.2 (Turbine Trip - steam dump system unavailable) and Event 11.2.4 (Closure of both MS!Vs - valve closure time is
> 0.1 second).
The loss of Exterr.41 Load event primarily challenges the acceptance criteria -
for both primary and secondary system overpressure and DNBR. The event rosults in an increase in the primary system Mmperatures due to an increase W the secondary side temperature. As the primary system temperatures increase, the coolant expands in the pressurizer causing an increase in the pressurizer pressure. The prir;ary system is protected against overpressurization by the ,
pressurizer safety and reitef valves. Pressure relief on the secondary side is afforded by the steam line safety / relief valves. Actuation of the primary and secondary system safety valves limits the magnitude of the primary system temperature and pressure increase.
With a positive moderator temperature coefficient, increasing primary system temperatures result in an increase in core power. The increasing primary side
ANF 87-161 Supplement 1 -
Page 74 temperatures and power reduces the margin to thermal limits (i.e , DNBR limits) and challenges the DNBR acceptance criteria.
15.2.1.2 Definition of Events Analyzed The objectives in analyzing this event are to demonstrate that: the primary pressure relief capacity is sufficient to limit the pressure to less than 110%
(2750 psia) of the design pressure, t,he secondary side pressure relief capacity is capable of limiting the pressure to less than 110% (1100 psia) of design pressure and the minimum DNBR remains above the safety limit. No credit is taken for direct reactor trip on turbine trip, the turbine bypass system or the steam dump system.
Two cases are analyzed for this event: one challenging the overpressurization criteria, and one challenging the fuel design limits. In both cases the input parameters are biased to maximize the increase in reactor power during the transient. However, for the system overpressurization event the parameters and equipment operational states are selected to maximize the system overpressure, while for the fuel design limit case the parameters and equipment are selected to reduce the system pressurization and thereby provide a conservative estimate of the minimum DNBR during the transient.
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The loss of ExNrnal Lead is credible only for rated power and power operation events because there is no load on the turbine at other reactor conditions. The rated power conditions bound the consequences for other reactor power operating conditions because of the increased stored energy.
The higher the stored energy in the primary system, the more severe the consequences of this event.
ANF-87-161 Supplement 1 Page 75 15.2.1.3 Analysis Results t, The maximum pressurization case initiates with a ramp closure of the turbine control valve in 0.1 seconds. The pressurization of the secondary side results in decreased primary.to secondary heat transfer, and a substantial rise in l primary system temperature. This results in an insurge into the pressurizer. l compressing the steam space and pressurizing the primary system. The reactor trips on high pressure. The capacity of the pressurizer safety valves contain the pressurizer pressure to a maximum of 2604 psia. The maximum RCS pressure at the bottom of the reactor vessel is 2697 psia. ,
The minimum DNBR case iJ initiated in the same manner. The transient proceeds in a similar fashion with the primary side pressure being limited by the
. pressurizer PORV setpoint. The resulting minimum DNBR is 1.39. The peak !
pellet LHR is 17.6 kW/ft.
The responses of key system variables are given in Figures 15.2.1-1 to 15.2.16 for the maximum pressure case and Figures 15.2.17 to 15.2.1-12 for the minimum DNBR case. The sequence of events is given in Table 15.2.1 1. l The secondary side pressure relief valves contain sufficient capacity to limit the pressure to less than 110% (1100 psia) of design pressure, ,
i 15.2.1.4 Conclusion The calculated minimum DNBR for the loss of Load event is above the heat flux correlation safety limit, so the DNB SAFDL is not penetrated in this event.
The peak pellet LHR is less than the limit of 21 kW/ft. The maximum pressure i
! remains below 110% of design pressure. Applicable acceptance criteria for the ,
event are therefore met.
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ANF 87-161 Supplement 1 Page 76 Table 15.2.1 1 Event Summary for the loss of External Load Event Maximum Pressurization Event Sumary Eyfat Time heel Turbine Trip 0.00 Pressurizer heaters on 0.00 Steam Line E)fety Valves Open 3.34 Peak Steam Dome Pressure 4.67 Pressurizer Safety Valves Open '5.27 Reactor Scram (Begin Rod Insertion) on High Pressurizer Pressure 5.s2 Peak Power 5.80 Peak Pressure 6.36 Peak Co*e Average Temperature 8.36 Minimum DNBR Event Sumary
[yani fire fsee) -
Turbine Trip 0.00 Minimum DNBR 0.72 Steam Line Relief Valves Open 3.25 Pressurizer PORY opens 4.91 Pressurizer Safety Valves Open 5.35 Peak Pressurizer Pressure 5.36 Reactor Scram (Begin Rod Insertion) on High Pressurizer Pressure 6.22 Peak Power 6.68 i
Peak Core Average Temperature 8.83 I .
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ANF-87-161 Supplement 1 Page 89 15.2.4 CLOSURE OF A SINGLE' MAIN STEAM ISOLATION VALVE (MSIV) 15.2.4.1 Event Descriotion Closure of the Main Steam isolation Valve event is initiated by the loss of control air to the MSIV operator. The vals es are swinging check valves designed to fail in the closed position. The closure of these valves in a PWR can drastically reduce the steam load.
15.2.4.2 Definition of Events Analyzed From Reference 3, the limiting case is obtained when the event is initiated from rated full power conditions. For simultaneous closure of both MSIVs, the event will progress very similar to Event 15.2.1. The turbine stop valve closure time employed in the 15.2.1 analysis (0.1 see) is much smaller than the HSIV closure time (6 sec). Thus, the consequences of Event 15.2.1 will bound those of the dual MSIV closure event.
The asyneetric conditions resulting frcm the closure of only one of the two NSIVs is similar to that predicted for a steam line break. That is, the primary coolant loop associated with the closed MSIV experiences a heatup due to the loss of heat sink and the primary coolant loop associated with the open MSIV experiences a cooldown due to the perceived load increase. The temperr.ture increase seen by the hot loop will be limited by the v.tuation of the steam generator safety valves. The temperature decrease seen by the cooling loop will continue until such time as a reactor trip is generated.
Since the loop experiencing the cooldown will see the larger temperature change, the limiting conditions for the event are at end-of cycle. The end-of-cycle moderator temperature coefficient (MTC) is larger in absolute magnitude than the beginning-of-cycle MTC. When the larger MTC is coupled with the larger temperature change in the cooling loop, a larger overall
ANF-87-161 i' Supplement 1 .
Page 90 increase in core power will be predicted. : This. larger increase in core power- '-
will produce the limiting DNB conditions for the event.
15.2.4.3 Analysis Results The analysis considered both BOC and EOC initial conditions and was performed assuming that one of the two MSIVs close initiating a heatup on the side of -
the core associated with the closed MSIV. The side of the core associated with the open MSIV experiences a cooldown corresponding to a factor of two initial increase in steam flow. That is, the steam generator connected to the open MSIV is assumed to initially provide 100% of the rated power steam flow.
As the pressure in the steam generator drops, the amount of steam which is supplied by the steam generator is also modeled to decrease. This decrease in steam flow with decreasing steam generator pressure is modeled based on end-of-cycle coastdown data.
Due to the event asymmetry and the fact that the event proceeds nuch like a steam line break event, the steam line break methodology and RELAP5/M002(21) input deck was used to perform this analysis. The neutronics input required <
to predict the radial power distribution between the cold and the hot side of tne core was, however, redeteloped based on event specific XTGPWR(6) calculations. These XTGPWR calculations differ from the steam line break calculations due to the difference in the power range of interest. The steam line break analysis requires power distribution data for powers less than 100%
of rated power, whereas the single MitV closure requires power distribution, data around 100% of rated power.
The results of the limiting E0C analysis are given in the event rummary, Table 15.2.4-1, and in Figures 15.2.A 1 through 15.2.4-5. As indicatcJ in the event ,
summary table the secondary safety valves open early in the transient limiting the temperature rise en the hot side of the core associated with the closed
~
ANF-87-161 Supplement 1 Page 91 MSIV. The reactor trips on low steam generator pressure which terminates the power rise. 1 The peak LHR and MDNBR is predicted to occur on the cold side of the core shortly after the reactor trip. The peak LHR is 20.9 kW/ft and the deterministic MDNBR was found to be 1.01. Whereas this is less than the '
95/95 DNBR limit, it is greater thar. the dete.rministic hDNBR calculated for the loss of flow event. The loss of ficw event has been shown to have margin with respect. to DNB limits based on statistical analyses. Thus it is concluded that the DNB limits will not be violated and that fuel failures are precluded during the single MSIV closure event.
The secondary side safety valves were modeled with a 3% drift allowancc. The
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plant Technical Specifications (7) limits the drift to 1%. Accounting for this difference in drift allowance, the maximum secondary side pressure is abouc 20 psi less than the maximum calculated of value of 1117 psia, or 1097 psia. Therefore, the maximum secondary side pressure is less than 110% (1100 psia) of design pressure.
l 1 "> . 2 . 4 . 4 ConclulinD The statistical setpoint analysis performed for th : loss of flow event has demonstrated that the minimum DNBR limit is not penetrated by the single MSIV closure event. The peak LHR is less than the 21 kW/ft limit to centerline melt. Maximum pressure is less than 110% of design pressure. Thus, the single MSIV closure event has been demonstrated to meet all required acceptance critt.ria.
s t
- ANF-87-161
, Supplement 1 Page 92 Table 15.2.4-1 Event Summary for the MSIV Closure Event Time (sec)
Reactor at full power 0.0 J' One MSIV closes instantaneously. Flow through 0.0 other MSIV: Initially doubles.
Secondary relief valves lift on the steam 4.0 generator with closed MSIV Scram signal on low steam' generator pressure 15.7 Rod insertion begins. Peak power of 115% and MONBR 16.6 of 1.01 d
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ANF-87 161 Supplement 1-
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15.2.7 LOSS OF NORMAL FEE 0 WATER FLOW .
15.2.7.1 Event Descriotion A Loss of Normal Feedwater Flow transient is initiated by the trip of the main i feedwater pumps or a malfunction in the feedwater control valves. Because the main feedwater system is supplying subcooled water to the steam generators, the loss of main feedwater flow will increase the secondary-side temperature and reduce the steam generator heat removal capability. The rise in the secondary-side temperature leads to a rise in the primary system coolant temperature. As the primary system temperatures increase, the coolant expands into th9 pressurizer which increases the pressure by compressing the steam volume.
The temperatures of the secondary sd. des and primary loops are controlled by the opening and closing of the safety valves on the steam lines. The long-term cooling of the primary system is assured by the secondary-side water inventory supplied by the Auxiliary Feedwater System. Two motor-driven auxiliary feedwater pumps are automatically started upon a steam generator low liquid level signal. If a loss of offsite power occurs, the motor-driven auxiliary feedwater pumps are powered by the emergency diesel. In addition, a turbine driven auxiliary feedwater pump can be manually actuated. '
15.2.7.2 Definition of Events Analyzed Two cases were analyzed for a loss of Normal Feedwater -Flow transient initiated from rated power operation:
- 1) Initial conditions and setpoints biased to maximize pressurizer liquid level.
E$
(t ANF-87-161 Supplement 1 ,
Page 99
- 2) Initial conditions and setpoints biased to minimize steam generator o liquid inventory.
The analysis is performed with the SLOTRAX-ML code (15) . The SLOTRAX-ML code includes relevant aspects of the mass and energy balance of the primary and :
secondary systems. The following events are assumed to occur at the loss of feedwater initiation.
- 1) Reactor trips on steam generator low level with specified time delay.
- 2) Turbine conservatively trips with simultaneous closure of' turbine stop valve at event initiation.
- 3) Main feedwater valves are assumed to instantaneously close at event initiation.
- 5) Start sequence for emergency diesel generators is initiated with 10
- t. minute delay for delivery of 600 gpm of auxiliary feedwater.
1 Additional conservative conditions are applied for analysis of each case to !
present the greatest challenge to the event acceptance criteria. Symmetric tube plugging is considered since ANF's methodology has shown that symmetric tube plugging produces the most limiting results.
13.2.7.3 Analysis Results <
l The event is initiated when each steam generator is at the low level trip setpoint. The steam generators are conservatively assumed to isolate at event initiation. For an assumed loss of offsite power, the primary coolant !
pumps are also tripped at event initiation. (
I
ANF-87-161 .I Supplement 1 Page 100 i
l o An event summary is presented in . Table 15.2.7-1.i The. transient responses are - 1 presented in Figures 15.2.7-1 through 15.2.7-4 for the minimum steam ,
generator inventory case and Figures 15.2.7-5 through 15.2.7-8 for the maximum pressurizer level case. In both cases, the transient execution time was 5000 seconds.
The maximum pressurizer liquid level is calculated to be 1301 ft 3 at 180 seconds. Sufficient steam volume remains to preclude the expulsion of licuid from the pressurizer safety valves.
The minimum steam generator level is calculated to be approximately 6.5 ft.,
occurring at about 1000 seconds after the event initiation. The steam a
generator level steadily recovers from this minimum level, thus ensuring continued heat removal.
15.2.7.4 Conclusions A loss of. normal feedwater event does act result in the violation of SAFDLs, l peak pressurizer pressu e does not exceed 110% of the design rating and
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primar/ liquid is not expelled through :he pressurd.zor safety valves.
Adequate cooling water is supplied by the auxiliary feedwater system to allow l a safe and orderly pl&nt shatdown and to prevent steam generator dryout.
Thus, the loss of normal feedwater event has been demonstrated to meet all required acceptance criteria, l I I
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ANF-87-161 Supplement 1 Page 101 Table 15.2.7-1 Event Summary for the loss of Normal feedwater Event Maximum Minimum Pressurizer S. G.
Level Level Event Case Case Reactor trips on low S.G. level AFil sequence initiated 0.0 sec 0.0 sec Primary coolant pump trip 0.0 sec N/A Turbine stop valve closed 0.2 sec 0.2 see Cessation of main feedwater flow 1.0 sec 1.0 see Pressurizer heaters actuate 2.0 sec 10.0 see S.G. safety valves of.en 12.0 sec 10.0 sec Maximum pressurizer liquid level 1301 ft3 N/A reached at 180 sec AF,W available to SGs 600 sec 600 sec Minimun S.G. level N/A 6.5 ft i reached at 1000 sec
ANF-87-161 Supplement 1 Page 102 i
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1 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW L 15.3.1 LOSS OF FORCED REACTOR C00lANT FLOW
)
15.3.1.1 Event Descriotion The Loss of Forced Reactor Coolant Flow transient is initiated by a loss of the electrical power supplied to or a mechanical failure in a reactor coolant system (RCS) pump. These failures may result in a :omplete or partial loss of forced coolant flow.
The impact of losing a RCS pump or pumps is a decrease in the active coolant flow rate in the reactor core and, consequently, an increase in core temperatures. Prior to reactor trip, the ccmbination of decreased flow and increased temperature poses a challenge to DNB limits. This event is terminated by the RCS low ficw tric.
15.3.1.2 Definition of Evult.epalyzed This event is analyzed from full power initial conditions. The core thermal margins are minimized at full power conditions resulting in this being the bounding mede of operation for this event. One case is analyzed for this event to assess the challenge to the DNB SAFDL.
The loss of coolant flow imediately results in a loss of system heat rejection capacity. This causes the primary system coolant temperature to increase. The objective of selecting input and blasing is to minimize ONBR.
The event analysis is, therefore, biased to rinimize pressure which minimizes ONBR. The steamline bypass and the atmospheric dump valves are both assumed not to operate, which again most challenges the DNB SAFDL.
[ , a
.. y ANF 87-161 Supplement 1 Page 111 15.3.1.3 Analysis Regn111 The transient is initiated by tripping all four primary coolant pumps. As the pumps coastdown, the core flow is reduced, causing a reactor scram on low flow. No credit was taken for the reactor coolant pump under-speed trip. As the flow' coasts down, primary temperatures increase. '
it s increase in
'emperature causes a subsequent power rise due to moderator reactivity feedback. The primary challenge to DNB is from the decreasing flow rate and resulting increase in coolant temperatures.
The ONBR unsequences of this event were evaluated using ANF's statistical f setpoint methodology.I4) The event MONBR was shown to be greater than thermal margin limits. For illustrative purposes, the deterministic minimum DNBR is 0.98. The peak pellet LHR is calculated to be 17.2 kN/ft. ,.
The responses of key system variables for the deterministic case are given in .
iigures 15.3.1-1 to 15.3.1-7. The sequence of events is given in Table :
15.3.1-1. !
15.3.1.4 C.gnglusion The statistical setpoint analysis demonstrates that the minimum DNBR limit is !
not penetrated by the loss of Forced Reactor Coolant Flow event. Maximum l
peak pellet LHR for this event is below the incipient fuel centerline melt cri mion of 21 kW/ft.
U J
-,, _..-u-_x----a.u --__----. . - - - - - - - - - - - - , - . , , _ . . - - - - _ _ _ _
ANF-87-161 Supplement 1 i Page 112 l
[
1 Table 15.3.1-1 Event Summary for the loss of Forced Reactor Coolant Flow t Eggni Time (sec) l Initiate Four Pump Coastdown 0.00
- i Letdown flow valve open 0.00
~7.eactor Scram (Begin Rod Insertion) 1.79 ft Turbine Stop Valve closed 1.90 I
Peak Power 2.27 j Minimum DJSR 2.80
{
. Pressurizer spray actuates 3.00 Peak Core Average Temperature 3.22 [
Peak Pressurizer Pressure 4.59 j Steam Line Safety Valves Open 5.65
[
Peak Steam Dome Prsssure 8.89 [
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. Page 120 l
. 15.3.3 REACTOR COOLANT PUMP ROTOR SEIZURE 7
15.3.3.1 Event Descriotion This evar.t is initiated by a seizure of a RCS pump rotor. The seizure causes ,
an immediate reduction in RCS flow rate. As in the Loss of Forced Coolant l Flow event (Event 15.3.1), the impact of losing a RCS pump is a decrease in l
' the , active flow rate in the reactor core, and, consequently, an increase in I core temperatures. Prior to reactor trip, the combination of decreased flow and increased temperature poses a challenge to DNB limits. The event is !
terminated by the RCS low flow trip. .
t i
15.3.3.2 Definition of Events Analyzed i r
One case is analyzed for this event to maximize the challenge to the ONB limit. The bounding operating mode for this event is full power initial l conditions.
15.3.3.3 Analysis Results t
i The locked rotor analysis assumes the locked pump loss coefficient given by i the homologous pump curves at zero pump speed. The calculated minimum DNBR is
- 0.96. The peak pellet LHR for this event is 17.4 kW/ft. The sequence of events is given in Table 15.3.3-1 and the responses of key system variables l are given in Figures 15.3.3 1 to 15.3.3-7 for the deterministic case, i f
I f
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i I
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is ANF-87-161 Supplement 1-Page 121
)
The minimum DNBR for the locked rotor event was calculated using a i deterministic application of uncertainties. The minimum DNBR is 0.02 less l than the deterministic minimum DNBR for the loss of flow event, Event 15.3.1 l (MDNBR = 0.98). A statistical application of uncertainties demonstrated that the minimum DNBR for the, loss of flow event is greater than the XNB thermal margin limit. Due to the presence of margin to the DNB LCO limits for a loss of flow event and the inherent similarities between a locked rotor and a loss of flow transient, fuel failures are precluded for the locked rotor event. .
15.3.? 4 Conclusion The statistical satpoint analysis for the loss of flow event demonstrates that the minimum DNBR limit is not penetrated by the locked rotor event.
Therefore, no fuel failures are expected for this infrequent event.
The peak LHR is less than the 21 kW/ft limit to centerline melt.
l l
l l
l
ANF-87 161 Supplement 1 Page 122 Table 15.3.3-1 Event Summary for a Reactor Cvolant Pump Rotor Seizure 4
Event Time (sec)
Reactor Coolant Pump Rotor Seizes 0.00 Reactor Scram (Begin Rod Insertion) 0.72 Turbine Stop Valve closed 0.80 Peak Power 1.19 Minimum DNBR 1.40 Peak Core Average Temperature 1.63 Pressurizer spray actuates 1.70 Peak Pressurizer Pressure 3.04 Steam Line Safety Valves Open 4.15 Peak Steam Dome Pressure 5.09 N
I t
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_________.____.________m___ ___ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ -
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l FIGURE 15.3.3-2 CORE AVERAGE HEAT FLUX FOR REACTOR ^~~
E COOLANT PUMP ROTOR SEIZURE
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MS COOLANT PUMP ROTOR SEIZURE
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ANF 87-101 ,
Supplement 1 Page 130 .
15.4 REACTIVITY AND POWER DISTRIBUTION ANOMAL.!ES :
15.4.1 UNCON"ROLLED ':0NTRO. RO) ASSEMBLY (CRO WITHDRAWAL' FROM A .
SURCR TICAL Olt LOW MER STARTUP COND TION t i
15.4.1.1 Event Descrintion J This event is initiated by the uncontrolled withdrawal of a control rod bank, i This withdrawal adds positive reactivity to the core which leads to a power ;
excursion. Event 15.4.1 considers the consequences of the control bank [
tsithdrawal at suberitical or low initial power levels. ;
i As the control bank is withdrawn, the positive reactivity insertion causes a i significant core power increase as the reactor approaches prompt criticality.
As core power increases, the core average and hot leg temperatures also l increase. Due to the increasing power and temperatures, the ONB limits are i challenged. The transient eventually terminates on a variable high power f reactor trip signal, ,.
15.4.1.2 Definition of Events Analyzed l
Whenever the rod control system is energized, Technical Specifications j require four operating reactor coolant pumps, although pressure is allowed to ,
be as low as 2000 psia in Mode 3(I) . The greatest power rise for this event l ts obtained when it is initiated from the lowest power. Therefore, the event !
was analyzed from a Icw power critical Mode 2 initial condition at 2000 psia, f These conditions will bound all other low power or suberitical cases. The -
l' event was analyzed to support increased radial peaking and a more positive moderator temperature coefficient.
?
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l ANF 87-161
.'. Supplement I h ,,
Page 131 g
I 15.4.1.3 Analysis Results i i
The event is initiated with the withdrawal of a control bank. The resultant prompt critical power excursion results in a fuel temperature increase and l
negative Doppler reactivity feedback which limits the peak power. The :
t transient is terminated when control rods are inserted upon a variable.high power trip. The responses of key system parameters are plotted in Figures i i 15.4.1-1 to 15.4.1-6. The sequence of events is given in Table 15.4.1-1. t i
The MDNBR is evaluated for conditions at the time of peak clad surface heat flux, and accuunts for elevated zero power peaking. The MDNBR was calculated to be 1.55. The peak LHR 15 calculated to be 20.1 kW/ft. Thus, no fuel i failures are predicted to occur. .
15.4.1.4 Conclusion No fuel failures are predicted for this event. Therefore, the event meets the {'
applicable acceptance criteria.
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ANF-87-161 Supplement 1 1 Page 132<
l Table 15.4.1-1 Event Summary for the Uncontrolled Bank Withdrawal from Low Power Event ,
l Event Time (sect Bank Withdrawal begins 0.00 Letdown Valve Open O.00 Scram Reactivity insertion begins 23.79 Peak Nuclear Power 24.30 [
Peak Core Heat Flux 24.70 Pressurizer Spray Actuates 31.80 L
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' Supplement 1 4 1T . Page 139 15.4.2 UFCONTN6 tug CONTROL- ROD BANK hlTHDRAWAL AT POWEB t !
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15.4.2.1 Event Descriotion :
As with Event 15.4.1, this event is initiated by an uncontrolled withorawal of a control rod bank. This withdrawal adds positive reactivity to the core which leads to potential power and temperature excursions. Event 15.4.2 considers the consequences of control bank withdrawals at rated ard operating initial power levels.
As the control bank is withdrawn, the positive reactivity insertion causes an increase in core power and primary coolant system temperatures. Due to the
? increasim power and temperatures, the DNB limits are challenged. In general, this transient wi?! terminate on either a variable high power (VHP) or a TM/LP trip.
15.4.2.2 Definition of Events Analyzed 6
The analysis evaluates the consequences of an uncontrolled control rod bank withdrawal from rat 9d power. A spectrum of reactivity insertion rates were evaluated in order to bound events ranging from a slow dilution of the primary
. -system boron concentration to the fauest allowed control bank withdrawals.
Specifically, the analysis encompasses reactivity insertion -ates from 4. x ,
10-6 to 4. x 10'4 Ap/sec.
15.4.2.3 Analysis Results The uncontrolled control bank withdrawai transients were analyzed for full -
power conditions (MO% of rated). The limiting uncontrolled control rod bank withdrawal at 100% power and EOC kinetics occurred at an insertion rate of 4.
. ANF-87-161 Supplement 1 . ,
Page 140 ;
l x 10-6 Ap/sec. . The minimum DNBR was calculated. as 1.21. This' transient '
tripped on a thermal margin / low pressure signal. The maximum peak pellet LHR 1 occurs in a 100~, power case which uses BOC kinetics.. The maximum peak pellet :
LHR.is calculated to be 19.1 kW/ft.
I The sequence of events for the thcontrolled Bank Withdrawal transient is given -
in Table 15.4.2-1. . The transient _ response of key system variables are given in Figures 15.4.2-1 to 15.4.2-6. ,
f
- 15.4.2.4 ~ Conclusion Reactivity insertion transient calculations demonstrate that tt XNB l
correlation limit will not be penetrated during any credible rear.sivity "
insertion transient at full power. The maximum peak pellet linear heat l
t generation rate for this event is less than the fuel centerline melt criterion of 21 kW/ft. Applicable acceptance criteria are therefore met, and the adequate functioning of the tnermal margin / low pressure trip. is
! demonstrated. l l
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ANF-87-161 Supplement 1 Page L41 Table 15.4.2-1 Event Summary for the Uncontrolled Rod Bank Withdrawal Event for the Limiting 100% Power Case Event Time (sec)
Start Rod Withdrawal 0.00 Letdown flow valve open 0.00 Peak Heat Flux 80.93 TM/LP Trip Signal 512.86 Turbine Stop Valve closed 513.00 Peak Power Level 513.32 Minimum DNBR 513.36 Peak Core Average Temperature 513.39 Steam Line Safety Valves open 514.94 Peak Steam Dome Pressure 516.96 S
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on Time, sec 3.34,
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we FIGURE 15.4.2-1 REACTOR CORE POWER FOR AN UNCONTROLLED CS j BANK WITHDRAWAL AT POWER l
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Supplement 1 : i
'Page.148
. a 15.4.3 CONTROL ROD MISOPERATION The control rod misoperation event considers a number of dif ferent event initiators. Those analyzed are:
(1) Dropped control rod or bank (2) . Single control rod withdrawal Both of the above events include a redistribution of power which leads to a local augmentation' of the peaking factor in the affected region of the core.
15.4.3.1 Event Descriotion (1) Droceed Control Rod / Bank A control rod / bank drop event is initiated by a de-energized control element drive mechanism (CEDM) or another failure in the control rod system. With the insertion of negative reactivity due to the dropped rod, the core power decreases. Moderator and Doppler temperature feedback, driven by a constant turbine generator load, cause the power to return to its initial state. A localized increase in the radial peaking results from the power redistribution due to the dropped rod / bank. This event is a challenge to DNB limits because of the radial peaking augmentation combined with near full power operating conditions.
e
ANF-87-161
~'
Supplement 1
. Page 149 (2) Sinole Control Rod Withdrawal The withdr'awal of a single control rod results in a reectivity insertion and a localized increase in radial peaking. The degradation of core conditions characteristic of a reactivity insertion trant.1ent, combined with an increase in local radial peaking, poses a challenge to DNBR limits.
15.4.3.2 Definition of Events Analvud (1) Droceed Control Rod /8ank The analysis evaluates the consequences of this event from rated power conditions. A control bank drop was fcund to produce a variable high power trip and, therefore, does not pose a challenge to DNB limits. The minimum DNBR for a control rod drop event from full power and full flow with increased radial peaking was analyzed.
(2) Sinole Control Red Withdrawal The overall system response to the withdrawal of a single CEA will be identical to the response of a CEA bank withdrawal. The only difference will be that the core will experience localized peaking in the 'ticinity of the t withdrawn CEA that is not present if an entire bank is withdrawn.
This event was analyzed at rated power conditions. Radial peaking augmentation factors to account for localized peaking redistribution were utilized in the assessment of the challenge to MDNBR limits. 1
, 1
H ::.
w.:A ANF-87-161 Supplement 1 Page 150 15.4.3.3 Analysis Results Results are given in Table 15.4.3-1 for the Control Rod Misoperation events.
Radial peaking augmentation factors for dropped control rod / bank events and i single control rod withdrawal events are calculated at full power for different exposure conditions. Bounding radial peaking augmentation factors were used in the analysis. In addition, bounding values of control rod and bank worth were used.
(1) Droceed Control Rod / Bank The largest RCS pressure drop experienced in the control rod drop transient is 60.3 psid. This input was used in the DNB LC0 verification analysis.
The DNBR consequences for the control rod drop event were evaluated using ANF's statistical setpoint methodology.(4) The statistical event MDNBR was shown to be greater than thermal margin limits. For illustrative purposes, the deterministic minimum DNBR is 1.14. The peak pellet LHR is calculated to be 17.7 kW/ft.
(2) Sinale Control Rod Withdrawal The deterministic minimum DNBR for the single rod withdrawal event is 0.98.
The amount of fuel failure for this event is less than 11.5%, the amount of fuel that fails as a consequence of a control rod ejection event (Event 15.4.8). The peak pellet LHR is calculated to be 19.1 kW/ft.
A
. . . , . . . . 3 ANF-87-161'
- _ Supplement 1 Page 151 15;4'3.4 . Conclusion (1) Drooned Control Rod / Bank
! The minimum DNBR for the control rod drop event was statistically shown to be.
. greater th an t hermal margin limits. Thus, the DNBR SAFDL is not penetrated.
The maximum peak linear heat rate for the rod drop event is such that fuel centerline 'mel.t is not expected. Applicable acceptance criteria for the rod drop event are therefore met for Millstone Unit 2.
(2) Sinale Control Rod Withdrawal The percentage of the core that experiences fuel failure during a single rod withdrawal event is bounded by the consequences of a control rod ejection accident. The maximum peak linear heat rate for the single rod withdrawal event is such that fuel centerline melt is not expected.
, .k.
lk f
4 ANF-87-161 t Supplement 1 s.
Page;152 1 Table 15.4.3-1 Summary of MONBRs for Control Rod'Hisoperation Eventsi Operating Maximum ',
Event (Power) Mode (11' MQHRH ,
,LHR (kW/ft) ,
Dropped: Control Rod _(100%) 1 (2) 17.7 Single Rod Withdrawal (100%) 1 0.98(3) 19.1 1
i L
(1) These modes are defined in Reference 7.
(2) DN8 LCO was statistically shown to protect against penetration of MDNBR ,
limits for this event. The deterministic MDNBR is 1.14. !
(3) Amount of fuel failure is bounded by the Control Rod Ejection fuel . !
failure calculation. r l
l s , .
ANF-87-161
.i . Supplement 1
. . Page 153 15.4.6 CVCS MALFUNCTION THAT RESULTS IN A DECREASE IN THE BORON
, CONCENTRATION IN THE REAC10R COOLANT 15.4.6.1 Event Descriotion A loron dilution event can occur when demineralized water is added to the primary coolant system via the Chemical Volume and Control System (CVCS) or the Precise Control of Reactivity System (PCRS), resulting in decreasing boron concentration in the primary system coolant. This dilution of the primary system coolant boron concentration results in the addition of positive reactivity to the core. This event can lead to an erosion of shutdown margin for subcritical initial conditions, or a slow power excursion for at-power conditions. A boron dilution at rated or power operating conditions behaves in a manner similar to a slow uncontrolled rod withdrawal transient (Event 15.4.2).
15.4.6.2 Definition of Events Analyzed The boron dilution analysis evaluates the time to criticality caused by the dilution of the primary system boron and the subsequent loss of shutdown margin. This analysis determines the shutdown cooling system flow rate needed to meet the operator response time criteria for Refueling (Mode 6), Cold Shutdown (Mode 5), Hot Shutdown (Mode 4), and Hot Standby (Mode 3). The systems that would be involved in the boron dilution event, depending upon the mode of operation are the reactor coolant system, the shutdown cooling system and the chemical and volume control system.
The major differences between the operating modes are the system parameters which affect the rate at which boron dilution occurs and the boron mixing model used once the demineralized water it injected into the lower plenum of
ANF-87-161 Supplement 1 ..
Page 154 the reactor vessel. Parameters such as charging , pump capacity and primary ,
system water volume affect the dilution rate. The mixing model used depends :
on whether the reactor coolant pumps are operating and whether the shutdown- .
coolant system discharge is to both cold legs when the dilution is postulated to occur.
For the six modes of operation, two mixing cases are considered: uniform mixing and slug flow. In the uniform mixing case, the diluting water is assumed to be uniformly mixed imediately upon injection into -the primary system. Uniform mixing is assumed to occur if one or more reactor coolant 4
pumps are in operation.
A slug flow type of dilution is assumed to occur during operation of the shutdown cooling system when the main reactor coolant pumps are not running.
It is assumed that the diluted water from the shutdown cooling system discharged to the lower plenum will 'not imediately mix with the reactor coolant due to the relatively low flow rate.
Symetric and asymmetric variations on the slug flow case are considered.
In the symmetric flow variation, diluting water is assumed to be injected into two or more cold legs located on opposite loops of the reactor coolant system.
In the asymmetric flow variation, diluting water is assumed to be injected .
into only one cold leg. The asymmetric flow variation is more limiting than the symmetric flow variation. In either case, the time to criticality is reduced if the shutdown cooling system flow is reduced.
The boron dilution analysis also includes a calculation to determine the rate -
of reactivity insertion due to inadvertent boron dilution during Startup ,
(Mode 2) and Full Power Operation (Mode 1).
u
1 ANF-87-161 Supplement 1 Page 155 l >
154.6.3 Analysis Results
, Table 15.4.6-1 presents the minimum shutdown cooling flow for Modes 4 to 6 required to avoid a complete erosion of shutdown margin within the required operator response time' II) . These results are based on the asymmetric flow model. For Modes 4 to 6, the reactor vessel is taken to be filled to the mid-plane of the coolant loop nozzles. This is the minimum expected reactor coolant system volume during shutdown cooling operations. The shutdown margin (SDM) corresponding to the SDM boron concentration is 5% (bounding of K,ff =
O.95) for Mode 6, 2% for Mode 5 and 3.6% for Mode 4.
For Mode 3, the response time to avert criticality is computed to be 141 minutes, which is well in excess of the acceptance criteria of 15 minutes.
This result is based on the use of the uniform mixing assumption, a coolant volume of 8755 ft 3, a critical baron concentration of 871 ppm and a SDM boron concentration of 1157 ppm. The SDM concentration corresponds to a shutdown margin of 3.6%.
For Modes 1 and 2, it was assumed that a slow power excursion occurs as a result of boron dilution. The power excursion is terminated by a reactor trip. The maximum reactivity insertion rate is -4.643 x 10 6 Ap/see for Mode 2 and -4.041 x 10 6 Ap/sec, for Mode 1. These reactivity insertion rates are bounded by the uncontrolled rod withdrawal events 15.4.1 and 15.4.2.
15.4.6.4 Conclusions The results of the boron di'.ution analysis show that there is adequate time for the operator to manually terminate the source of dilution flow during Modes 6, 5 and 4. For Mode 3, the results show that there is adequate time for the oprator to manually terminate the source of dilution flow. The
i ANF-87-161 1' Supplement 1 -
Page 156 i.
operator can initiate reboration to recover the shutdown margin.. Boron dilution during power ' operation is bounded by - the analyses presented in i Sections 15.4.1 and 15.4.2. The results presented here demonstrate that there -
is adequate time for the operator to manually terminate the source of dilution .
flow following reactor trip.
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ANF-87-161 Supplement I?
Page 157 c Table 15.4.6-1 Summary of Results for the Boron Dilution Event for All Modes Where Slug Flow Co.nditions May Exist .
Charging Coolant Response Min. 50C Operating Flow Flow Volu BoronConc(ppm) Time Flow Mode faomi HQdil Critical SDM (min) .foom) 6 132 A 3202 1260 1655 30 1300 5 132 A 3202 913 1056 15 2450 4 132 A 3202 892 1165 15 2450 I
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A Asymmetric dilution model
ANF-87-161 Supplement 1. !
Page 158 16.4.8 SPECTRUM OF CONTROL ROD EJECTION ACCIDENTS 15.4.8.1 Event Description This event is initiated by a failure in the control element drive mechanism $
(CEDM) pressure housing causing a rapid ejection of the affected control rod.
The ejection- of the control rod inserts positive reactivity causing an increase in core power. Because of the increase in core power, this event challenges deposited enthalpy DNBR and pressurization acceptance criteria, 1
15.4.8.2 Definition of Events Analyzed Due to the complex interaction of the ejected rod worth, ejected peaking factor and Doppler feedback effects, it is difficult to a priori bound the
! consequences of the event for either rated power or hot critical operating
- coriditions. Therefore, each of these conditions were evaluated at both BOC -
and EOC for deposited enthalpy, DNBR and pressurization concerns.
For the evaluation of the DNBR and pressurization consequences, concurrent loss of offsite power is assumed. No credit is taken for the variable high power trip in the analysis of the pressurization consequences of a control rod ejection.
15.4.8.3 Analysis Results The hot full power (HFP) control rod ejection event was determined to deposit more energy into the primary system than the event initiated from hot zero power (HZP). Therefore, in terms of the event acceptance criteria, the HFP
- event poses a greater challenge than the HZP event. For this analysis, the 2 event was assumed to initiate from HFP at 102% of rated full power.
ANF-87-161 Supplement 1 Page 159 4
l To assess the acceptability of the outcome of a HFP rod ejection event, two ,
cases were examined. 'The first case determines the maximum pressurization -
potential of the primary system during this event. The second case evaluates the MONBR. For both the maximum pressurization and minimum DNB case, B0C and _
EOC kinetics were employed to establish the respective limiting cases.
l The limiting minimum DNB case is calculated to occur for BOC kinetics. Core boundary conditions used to evaluate the ONBR conservatively account for l
depressurization due to the postulated breach in the CEDM housing. The deterministic HDNBR is calculated to be 0.81. As a consequence of this event, less than 11.57. of the fuel rods are calculated to fail due to penetration of DNBR limits. The responses of key system parameters are shown in Figures 15.4.8-1 to 15.4.8-6.
The maximum pressurization case also occurs for B0C kinetics. However,. the peak pressurizer pressure remains below 1107. of the pressure vessel design limit. The peak pressurizer pressure is conservatively calculated to be 2671 l
psia. Key system parameters for the overpressure case are plotted in Figures
! 15.4.8-7 to 15.4.8-10.
The sequence of events for the Control Rod Ejection transient is given in Table 15.4.8-1.
The deposited enthalpy portion of the rod ejection accident has been evaluated with the procedures developed in the Generic Rod Ejection Analysis (II) . The ejected rod worths and hot pellet peaking factors were calculated using the XTGPWR code. No credit was taken for the power flattening effects of Doppler or moderator feedback in the calculation of y ejected rod worths or resultant peaking factors. The calcu'ations performed
- I ANF-87-161 Supplement 1 Page 160 used a full core three-dimensional XTGPWR model. The pellet energy deposition resulting from an ejected rod was conservatively evaluated explicitly for B0C and E0C conditions. The HFP pellet energy deposited was calculated to be 3 240.6 cal /g. The HZP pellet energy deposition was calculated to be 171.1 cal /g. The rod ejection accident was found.to result in an energy deposition of less than the 280 cal /g limit as stated in Regulatory 1.77. The significant parameters for the analyses, along with the results, are -
summarized in Table 15.4.8-2.
15.4.8.4 Conclusion The maximum pressurizer pressure does not exceed 1107. of the design pressure.
Less than 11.5?. of the core will experience fuel failure due to penetration of DNBR limits. Deposited enthalpy is less than the limit of 280 cal /g.
ANF-87-l61 Supplement 1 Page 161 Table 15.4.8-1 Event Summary for a Control Rod Ejection Maximum Pressurization Case Event Summary
. Event Time (sec)
Control Rod Ejects 0.00 Makeup Pumps On 0.00 Pressurizer Heaters On 0.00 Reactor Scram (rods begin insertion) 5.14 Torbine Stop Valve Closed 5.20 Peak 7:wer 5.62 Pressurizer M ety Valves Open 5.85 l Peak Core Avg. Temperature 6.19 i Peak Pressure 6.48 Steam Line Safety Valves Open 10.09 Peak Steam Dome Pressure 13.41 Minimum DNBR Case Event Summary Event Time (sec)
Control Rod Ejects 0.00 Letdown Valve Open 0.00 Reactor Scram (rods begin insertion) 1.93 Turbine Stop Valve Closed 2.00 Peak Power 2.41 Peak Core Avg. Heat Flux 2.57 Minimum DNBR 2.68 i Peak Core Avg. Temperature 3.07 I
Pressurizer Spray Actuates 3.20 Peak Pressurizer Pressure 4.28 i Steam Line Safety Valves Open 5.81 l Peak Steam Dome Pressure 9.98 l
l l
l 1
l l
Table 15.4.8-2 Bounding B0C/E0C Ejected Rod Analysis HFP H7P ContributionI *I to Contribution (8) to Energy Deposition, Energy Deposition, Value (cal /a) Value feal/a) l A. Initial Fuel Enthalpy (cal /g) 103.4 ---
16.7 ---
- 8. Generic Initial Fuel Enthalpy (cal /g) 40.8 ---
16.7 ---
C. Delta Initial Fuel Enthalpy (cal /g) 62.6 62.6 0 0 D. Maximum Control Rod Worth (pcm) 200 131.4 300 102.6 l
l E. Doppler Coefficient (pcm/*F) -0.8 1.17 -0.8 1.31 F. Delayed Neutron fraction, 0.0045 1.06 0.0045 1:28 G. Power Peaking factor 6.0 -
--- 13.0 ---
Total Fuel Enthalpy (cal /g) 240.6(b) 373,3(b)
(a) The contribution to the total pellet energy deposition is a function of initial fuel enthalpy, maximum control rod worth, Doppler coefficient, and delayed neutron fraction. The energy deposition contribution values and factors are derived from data calculated in the generic analysis of the control rod ejection transient document XN-NF-78-44. E EN 2' r!'
(b) Total pellet energy deposition (cal /g) calculated by the equation: - -
%2$
Total (cal /g) - (C+D) (E) (F) a, '
ll il)1l1 ll;'
gTS:. 0 -
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Time, see a3e
- *87 FIGURE 15.4.8-6 SECONDARY PRESSURE FOR A CONTROL - -- $, $ .
R0D EJECTION (MDNBR CASE) l
}
\
4000 j PL 1
3500-1 I
3000-l i o f i
& 250joh ,
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- v. = -
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1000-i 500-I O , ,
E O 2 4 d j g'O l2 14 16 o->
l Time, see aB&
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-n-FIGURE 15.4.8-7 REACTOR CORE POWER FOR A CONTROL $~S R0D EJECTION (?RESSURE CASE)
)
225000 ,
QOA 200000-1 S a
y 175000-L
.C N
c a 150000-m v
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+
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25000 , , ,
0 2 4 6 8 1o g'2 14 16 8I Time, see T, !!! S i
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! FIGURE 15.4.8-8 CORE AVERAGE HEAT FLUX FOR A CONTROL ES i R00 EJECTION (PRESSURE CASE)
640
[
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- 9eg7---
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- Time, sec a, g =
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l FIGURE 15.4.8-9 REACTOR CO0laNT SYSTEM TEMPERATURES FOR A %S
- CONTROL R00 EJECTION (PRESSURE CASE)
--, - __ ____ _ m__ _ _-_____<---<>-e--- -
- = , , - - - - - -- ,-- , . - ,
2700 i
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l 2100 . . . . ,
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Time, sec FIGURE 15.4.8-10 PRESSURIZER PRESSURE FOR A CONTROL % ._. S
.; R00 EJECTION (PRESSURE CASE)
gTSew0 Eu I!! 3. er -
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ANF-87-161 Supplement 1 Page 174 6 2
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ANF-87-161 (
Supplement 1 Page 175 15.6 DECREASES IN REACTOR COOLANT INVENTORY 15.6.1 INADVERTENT OPENING OF A PWR PRESSURIZER PRESSURE RELIEF VALVE 15.6.1.1 Event Descriotion This event is initiated by the inadvertent opening of a pressurizer pressure relief valve or safety valve, which results in the blowdown of primary coolant as steam. The primary system pressure decreases rapidly until the pressurizer liquid is depleted, and then to a pressure determined by the hot leg saturation temperature. Reactor scram occurs on thermal margir:/ low i pressure well before the pressurizer liquid is depleted during the full power f case, thus terminating the challenge to SAFDLs. i The startup mode case is terminated with the boron injection following the 7 actuation of the SIAS on low pressure. The startup mode case is '
conservatively assumed to reach the hot leg saturation pressure at event termination. -
1 This event is primarily considered a depressurization event, but with a l negative moderator pressure coefficient and a positive moderator temperature coefficient, the thermal margin will be eroded with increased power, increased coolant inlet temperatures, and decreased pressures. In addition the event ;
can also uncover the core with a decrease in the primary coolant inventory. -
l 15.6.1.2 Definition of Events Analyzed in Reference 3, this event was disposed to be analyzed for MDNBR for both modes 1 (full power), and 2 (startup). The startup power case is disposed to l be analyzed because the TH/LP trip can be manually bypassed below 5% power. .
ANF-87-161 Supplement 1 Page 176 The system response for the full power case was evaluated by using !
PTSPWR2(I4) . The full power event minimum DNBR was calculated using XCOBRA-IIIC(11) ,
P The system response for the startup case was determined by conservative !
problem constraints. The maximum power was limited to 7% of the rated power.
Above this power the assumed TM/LP trip bypass is automatically removed. The system pressure is conservatively assumed to be at the hot leg saturation pressure. The coolant inlet tamperature is assumed to be at a level consistent with a maximum power rise of 7% and a conservative time delay before the safety injection system terminates the event. XCOBRA-IllC was used with these system responses to predict the hot channel mass flux required for the critical heat flux calculation. The thermal margin was determined by the Modified Barnett critical heat flux correlation (IO) .
15.6.1.3 Analysis Results The sequence of events for the full power analysis are given in Table ,
15.6.1-1. Figures 15.6.1-1 to 15.6.1-6 show the transient re=ponse for Key system variables. The minimum DNBR for this event initiated from full power is 1.20. Yhe peak LHR s calculated to be 18.3 kW/ft.
The startup mode case resulted in a minin.im critical heat flux ratio of above 10, as calculated by the Modified Barnett correlation. The peak pellet LHR ,
is less than the full power value. Thus, the startup mode is bounded by the [
full power mode. l The charging and safety injection systems have been shewn to have sufficient capacity to easily compensate for the loss of primary coolant mass threugh the i
i
4 ANF 87-161 Supplement 1 Page 177 i
inadvertent opening of a pressurizer pressure relief valve. Therefore, the core is not expected to uncover during this event.
15.6.1.2 Conclusions l f
The results of the analysis demonstrate that the event acceptance criteria are '
met since the minimum DNBR predicted for the full power case is greater than
, the XNB correlation safety limit and the minimum CHFR predicted for the startup mode case is greater than the Modified Barnett CHF limit. The [
correlation limits assure with 95% probability and 95% confidence, that DNB is not expected to occur; therefore, no fuel is expected to fail. The fuel centerline melt threshold of 21 kW/ft is not violated in this event, t
4
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L' Supplement 1 Page 178 r
Table 15.6.1-1 Event Summary for an Inadvertent Opening .
of a PWR Pressurizer Pressure Relief Valve l r
l [.ypal Time (sec)
Letdown Valve Open 0.00 :
Pressurizer Relief Valve Opens 0.01 l Reactor Scram (rods begin insertion) 9.11 l
Turbine Stop Valve Closed 9.25 ,
I l Peak Power 9.58 '
i MONBR 9.69 ;
l l Peak Core Avg. Temperature 9.72 i Steam line Safety Valves Open 12.35 [
l Peak Steam Dome Pressure 13.60 l l [
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i i
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{ i l ?
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U y p O 2.6 6 7.6 10 12.6 16 17.6 20 yJv Time, see a; 3 a,
- a?
-n-FIGURE 15.6.1-3 REACTOR COOLANT SYSTEM TEMPERATURES FOR AN 3S INADVERTENT OPENING Of A PWR PRESSURIZER PRESSURE RELIEF VALVE (RATED POWER)
PPR 2150-3.,2100-O b 2060-d ~
E o
y 2000-h v
tJ T 1960-
.3 .
sn 4 1900- .
D.
1860-1800 , , , , , , , ,5 3 0 2.6 6 7.5 10 12.5 t's 17.6 20 y v_, =
Time, see ajg
?. .' .
f!GURE 15.6.1-4 PRESSURIZER PRESSURE FOR AN INADVERTENT OPENING 2$
Of A PWR PRESSURIZER PRESSURE REllEF VALVE (RATED POWER)
1
~~~~,____.----- DK 0 -
. . - ~ '- 7) WOP 21010tr m m 8.e m
] ,
8 L g I T
Uo ~b~
M e
" -a-
_g. L
-9 i i i i e s a up 0 2.5 6 7.5 10 12.5 16 17.5 20 ug 1 Time, sec 3; 3 a,
- 8Y FIGURE 15.6.1-5 REACTIVITIES FOR AN INADVERTENT OPENING bb 0F A PWR PRESSURIZER PRESSURE RELIEF VALVE (RATED POWER)
.9
>ANF-87 161 Supplement 1 Page 184 2
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Supplement 1 Page 185
4.0 REFERENCES
l 1. "Standard Review Plan for the Review of Safety Analysis Reports' for' Nuclear Power Plants," NUREG-0800 LWR Edition, U.S. Nuclear Regulatory Comission. Office of Nuclear Reactor Regulation, July 1981,
- 2. "Advanced Nuclear Fuels Methodology for Pressurized Water Reactors:
Analysis of Chapter 15 Events," ANF-84-73f P). Rev. 3 Advanced Nuclear Fuels Corporation, Richland, WA 99352, May 1988.
- 3. "Millstone Unit 2 Cycle 10 Plant Transient Analysis Report Analysis of Chapter .15 Events," ANF-87-161 Advanced Nuclear Fuels Corporation, Richland, WA 99352, September 1988.
- 4. "ENC Setpoint Methodology for CE Reactors: Statistical Setpoint Methodology," XN NF 507( A). Sucolements 1 and 2, Exxon Nuclear Company, Richland, WA 99352, September 1986.
- 5. Millstone Nuclear Power Station Unit 2 Final Safety Analysis Report.
- 6. "XTG - A Two Group Three Dimensional Reactor Simulator Utilizing Coarse Mesh Spacing," XN CC-28(A). Revisi g l. Exxon Nuclear Company, Richland, WA 99352, October 1978.
- 7. Millstone Nuclear Power Station Unit 2 Technical Specifications, Appendix A. License No. OPR 65.
1 Exxon Nuclear Company, Richland, WA 99352, April 1982.
- 9. "Justification of XNB Departure from Nucleate Boiling Correlation for St. Lucie Unit 1," XN NF 83 08fP). Exxon Nuclear Company, Richland, WA 99352, February 1985,
- 10. "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," XN NF 82 21(A). Rev. 1, Exxon Nuclear Company, Richland, WA 99352, September 1983.
- 11. "XCOBRA-IllC: A Computer Code to Determine the Distribution of Coolant During Steady State and Transient Core Operation," XN NF-75-21f A). Rev.
- 2. Exxon Nuclear Company, Richland, WA 99352, January 1986.
ANF-87-161 :
Supplement 1 ,
Page 186 l.
1
- 12. "Advanced Nuclear Fuels Corporation Setpoint Methodology for CE Reactors:
Three Dimensional Axial Power Distribution Generation," ANF-507(P). [
Addendum 1. Advanced Nuclear Fuels Corporation, Richland, WA 99352, June .
1988.
- 13. "Computational Procedure for Evaluating Fuel Rod Bowing," XN-NF-75-32f A).
Sunes. 1. 2. 3 & 4 Exxon Nuclear Company Richland, WA 99352, October ;
1983. r
- 14. "Description of the Exxon Nuclear Plant Transient Simulation Model for ,
- Pressurized Water Reactors (PTS-PWR), XN NF-74 5fA). Rev. 2 and
. lucolements 3-6. Exxon Nuclear Company, Richland, WA 99352, October 1986. [
- 15. "SLOTRAX ML: A Computer Code for Analysis of Slow Transients in PWRs," ,
XN-NF-85 24fA), Exxon Nuclear Company, Richland, WA 99352, September '
1986. ;
- 16. E. Daniel Hughes. "A Correlation of Rod Bundle Critical Heat Flux for f Water in the Pressure Range 150 to 725 psia," IN 1412 (TID 4500), Idaho !
Nuclear Corporation, July 1970, f
- 17. "A Generic Analysis of the Coolant Rod Ejection Transient for Pressurized Water Reactors," XN-NF-78 44f A), Exxon Nuclear Company, Richland, WA i 99352, October 1983. [
- 18. "Millstone Unit 2 Small Break LOCA Analysis," ANF 88-129 Advanced l Nuclear fuels Corporation, Richland, WA 99352, October 1988. I
- 19. "Millstone Unit 2 Large Break LOCA/ECCS Analysis," ANF 88-ll8, Advanced !
Nuclear Fuels Corporation, Richland, WA 99352, August 1988.
- 20. Letter, Mr. M.W. Hodges (NRC) to Mr. R.A. Copeland (ANF), "ANF 3- f Olmensional Setpoint Methodology for Combustion Engineering Reactors "
July 8, 1988.
i
- 21. "RELAPS/M002 Code Manual, Volume 1: System Models and Numerical Methods; i Volume 2, Users Guide and input Requirements " NUREG/CR 4312. EGG 2396. :
Revision 1, EG1G Idaho, Inc., Idaho Falls, ID 83415, March 1987.
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