ML20154H260
| ML20154H260 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 08/02/1988 |
| From: | Lindquist T SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML18053A574 | List: |
| References | |
| ANF-88-108, NUDOCS 8809210242 | |
| Download: ML20154H260 (91) | |
Text
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}{;,,, ADVANCED NUCLEAR FUELS CORPORATION P ALIS ADES CYCLE 8: DISPOSITION AND AN ALYSIS OF ST AND ARD REVIEW PLAN CH APTER 15 EVENTS AUGUST 1988
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ADVANCEDNUCLEAR FUELS CORPORATION ANF 88 108 Issue Date: 8/2/88 PALISADES CYCLE 8: O!SPOSITION AND ANALYSIS OF STANDARD REVIEW PLAN CHAPTER 15 EVENTS Prepared by:
i T. R. L %dquist, Engineer PWR 3afety Analysis Licensing & Safety Engineering Fuel Engineering & Technical Services August 1988 j
gf
e CUSTOWER DISCLAlWER 48808tf AAT NOTx3 RtGAA04e4 CONTENTS AND USE OF THs4 OCCUt0ENT hCAltgE&L_f_
Advenged Nuche Pues Corporecorre eerretme eed recroosetatees con.
comeg the outHect marter of me document are tecos est larm e me Agreement N Aerenced Nuc>eer Puees Corcorecon and me Cmomer purswit to enecn me document 4 useued. 4.cordegry, enCoot as otl' A 'ee guersee!y pr1>
e@ed a suon Agreement, neeer Advanced Nuw Puees Ccroorston 4 / any pereen acting on te teneff mesee any werfenfy or recrementaten, encrossed or c9eed. wem roecect to the accuracy, comosetenees, or useMeees of the efor.
megen contained in mes document, of that the ues of any inermetson, apparatus, momed or pegcess - e this 00 current wdl not etnoge prWetery oweed ngnes; or tecumes any lleosistes witn roecect to me use of any ettmatt.i, ap.
Derefus, FNined or procese Sec80eed in mie document.
The iMermanen centaced Peron 4 tor me goes use of Customer in orGet to avoid umpestreont of ngnte of Advanced Nucieer P9e4 Corcoration in Detente or imennette anece may De encsuced Iri me c4rmenon ootitanned in mis document. me recouent. 39 4e accessence of me 00cument, agreet not to pushen or meme puosic use (in me pesent use of tN term) of san cermecon WetW to autherteed in writing Dy A&anced Nucteer Puces Corocrat49 or urtd efter sa (6) montne tonowog termeenen or enciraron of me aforeseed 6;reerr ent and any esteneson mereef. Wheets omerwee espressly pnWi6ed e ?e Aq'Serrent. "M nonte or iaconese e ce is any parente are coes ey tee tura sneg of this derv-i reent.
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ANF 88 108 l
Page i i
TABLE OF CONTENT,1 d
gigg Seetion e
.[
8
1.0 INTRODUCTION
3 2.0 SUMARY AND CONCLUSIONS 3.0 ANALYSIS OF PLANT TRANSIENTS..................
10 l
11 l
15.0 ACCIDENT ANALYSES 11
[
15.0.1 PLANT INITIAL CON 0!TIONS 13 i
15.0.2 POWER O!STRIBUTION.......................
15.0.3 REACTIVITY COEFFICIENTS USED IN THE SAFETY ANALYSIS 18 4
20 i
15.0.4 TRIP SETPOINTS.........................
20 l
15.0.4.1 Inlet Temperature L Miting Condition Of Operation a
i d
22 t
j 15.0.4.2 Thermal Mgrgin/ Low Pressure (TM/LP) Trip............
24 15.0.5 OISPOSITION AND ANALYSIS OF EVENTS...............
I 15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM........
25 l
l 25 l
15.1.1 OECREASE IN FEE 0 WATER TEMPERATURE,..............
25 1
15.1.2 INCREASE IN FEE 0 WATER FLOW...................
26 15.1.3 INCREASE IN STEAM FLOW.....................
15.1.4
!NADVERTENT OPENING OF A STEAM GENERATOR RELIEF 28 l
OR SAFETY VALVE i
15.1.5 STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE OF 29 CONTAINMENT
)
30 15.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM.,
t 30 15.2.1 LOSS OF EXTERNAL LOAD.....................
i 1
32 j
15.2.2 TURBINE TRIP
[
33 2
15.2.3 LOSS OF CONDENSER VACUUM...........
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[
ANF-SS.108 Page 11 TABLE OF CONTENTS ELqg Section 1
33 CLOSURE OF THE MAIN STEAM ISOLATION VALVES (MS!V) (BWR) 15.2.4 34 1
15.2.5 STEAM PRESSURE REGULATOR FAILURE................
34 LOSS OF NONEMERGENCY A.C. POWER TO THE STATION AUXILIARIE 15.2.6 35 15.2.7 LOSS OF NORMAL FEE 0 WATER FLOW.................
I 37 FEE 0 WATER SYSTEM PIPE BREAKS lhSIDE AND OUTSIDE CONTA 15.2.8 38 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW 39 l
15.3.1 LOSS OF FORCEO REACTOR COOLANT FLOW 40
+
15.3.2 FLOW CONTROLLER MALFUNCTION..................
40 15.3.3 REACTOR COOLANT PUMP ROTOR SElZURE..............
41 15.3.4 REACTOR COOLANT PUMP SHAFT BREAK................
42 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES I
UNCONTROLLED CONTROL R00 ASSEMBLY (CRA) WITH0RAWAL FROM 42 i
15.4.1 SUBCRITICAL OR LOW POWER STARTUP CONDITION...........
44 i
15.4.2 UNCONTROLLEO CONTROL R00 BANK WITH0RAWAL AT POWER f
46 15.4.3 CONTROL ROD MISOPERATION....................
54 15.4.4 STARTUP OF AN INACTIVE LOOP,.................
56 l
15.4.5 FLOW CONTROLLER MALFUNCTION..................
l 15.4.6 CVCS MALFUNCTION THAT RESULTS IN A DECREASE IN THE l
{
56 C0t.; CENTRATION IN THE REACTOR COOLANT INA0VEd. TENT LOADING AND OPERATION OF A FUEL. ASSEMBLf 3
62 15.4.7 IMPROPER POSITION.......................
f 62 15.4.8 SPECTRUM OF CONTROL R00 EJECTION ACCIDENTS...........
63 l
15.4.9 SPECTRUM OF R00 OROP ACCIDENTS (BWR)..............
63 l
INCREASES IN REACTOR COOLANT SYSTEM INVENTORY 15.5 1
l
ANF.88 108 Page til J.
i
. TABLE OF CONTENTS Section Egg l
15.5.1 INADVERTENT OPERATION OF THE ECCS THAT INCREASES REACTOR 64 COOLANT INVENTORY 15.5.2 CVCS MALFUNCTION THAT INCREASES REAC10R COOLANT INVENTORY 64 f
15.6 OECREASES IN REACTOR COOLANT INVENTORY.............
65 15.6.1 INADVERTENT OPENING OF A PWR PRESSURIZER PRESSURE RELIEF VALVE.
65 15.6.2 RADIOLOGICAL CONSEQUENCES OF THE FAILURE OF SMALL LINES CARRYING PRIMARY COOLANT OUTSIDE OF CONTAINMENT 66 15.6.3 RADIOLOGICAL CONSEQUENCES OF STEAM GENERATf,R TUBE FAILURE 66 15.6.4 RADIOLOGICAL CONSEQUENCES OF A MAIN STEAM L'NE FAILURE OUTSIDE 68 CONTAINMENT (BWR) l 15.6.5 LOSS OF COOLANT ACCIDENTS RESULTING FR084 A SPECTRUM OF POSTULATED P! PING BREAKS WITHIN THE REACTOR COOLANT PRESSURE i
BOUNDARY............................
68 15.7 RA010 ACTIVE RELEASES FROM A SUBSYSTEM OR COMPONENT 69 15.7.1 WASTE GAS SYSTEM FAlt.URE....................
69 15.7.2 RADIOACTIVE LIQUID WASTE SYSTEM l.EAK OR FAILURE (RELEASE TO 69 ATMOSPHERE) r t
15.7.3 POSTULATED RADIOACTIVE RELEASES DUE 10 LIQUID CONTAINING TANK L
69 FAILURES............................
15.7.4 RA0!0 LOGICAL CONSEQUENCES OF FUEL HANDLING ACCIDEN1 69 15.7.5 SPENf FUEL CASK OROP ACCIDENTS.................
70 l
72 l
4.0 THERMAL HYDRAULIC COMPAT!BILITY i
80 i
l
5.0 REFERENCES
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f ANF 88 108 Page iv i
i I
LIST OF TABLLS E
U 21 Disposition of Events Sumary for Palisades............
4 22 S uma ry o f Re s ul t s.....................................
9 15.0.1 1 Nominal Plant Operating Conditions.....................
12 15.0.2 1 Core Power Distributions...............................
15 3
4 15.0.3 1 Reactivity Parameters..................................
19 t
15.4.3-1 Sumary of Conditions for Control Rod
[
Misoperation Events....................................
53 l
1 15.4.6 1 Sumary of Results for the Boron Dilutiou l
Event..................................................
61 I
4-1 Fuel Design Parameters for Stainless 3 teel Shielding 79 l
Ass.emblies.............................................
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-. -.,. -. ~ -. - _.,,, _ _ - - _ _ -.,
ANF 88 108 Page v LIST OF FIGURES i
Ficure Elas 15.0.2 1 Axial Power Distribution at 2530 MWt (ASI =
0.139)...........................................
16 15.0.2 2 Axial Power Distribution at 1265 MWt (ASI =
0.342)...........................................
17
l ANF 88 108 Page 1
1.0 INTRODUCTION
This report documents the results of a Standard Review Plan (SRP)II) Chapter 15 disposition of events and analysis performed in support of Palisades Cycle 8 operation.
A modified reactor protection system (RPS),
including a variable. overpower trip and an improved thermal margin / low pressure (TM/LP) trip with axial monitoring, will be installed prior to Cycle 8 operation and is supported by the the analyses reported in References 2 and 3.
Additioni.;
changes that will be implemented into Palisades Cycle 8 are:
(1) An increase in Technical Specification radial peaking factor limits to accommodate a low radial leakage loading pattern for the purpose of reducing vessel fluence.
The radial peaking f actors will be increased by 3.57..
(2)
Insertion of four ANF lead assemblies with high thermal performance spacers.
(3) Reinsertion of sixteen previously burnt assemblies at locations along the core periphery to reduce neutron
'luence at critical vessel welds. Each of these assemblies will be reconstituted with 56 stainless steel rods replacing the fuel rods along the four outer rows on one side of the assembly.
l The Chapter 15 events were disposed and analyzed in accordance with Advanced Nuclear Fuels Corporation methodology.III The LOCA/ECCS analyses in support j
l of Palisades Cycle S are documented in Reference 10.
l l
of the results and review of SRP Chapter 15 l
Section 2.0 presents a summary events.
Section 3.0 presents tne conditions employed in the event analyses l
and the results of these event analyses.
Events are numeered in accordance with the SRP to f acilitate review.
A tabular list of the disposition of t
- I L l
l ANF 88 108 Page 2 Chapter !$ events and analysis of record for Palisades, with a cross II), is reference between SRP event numbers and the Palisades Updated FSAR included.
Section 4.0 presents the results of a
therma). hydraulic compatibility analysis for the four lead assembites and the sixteen stainless steel shielding assemblies.
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at ANF.88 108 Page 3 2.0
SUMMARY
AND CONCLUSIONS A sumary Disposition of Events for the changes proposed for Palisades Cycle 3 I
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.
i The changes listed in Section 1.0 for Cycle 8 do not alter the plant system f
response to a transient event relative to the analysis supporting modified RPS j
operation.(3)
The increase in radial peaking li: sits will, however, impact minimum Departure from Nucleate Boiling Ratio (ON8R).
Therefore, the analysis j
for the events disposed to be reanalyzed for Cycle 8 wiil consist of an evaluation of the minimum DNBR and DNBR related consequences (e.g.,
fuel failure) using :.he appropriate transient conditions in Reference 3.
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 accepttneo criteria are met for Cycle 8 operation.
These results support operation with up to 29.3*.
i average steam generator tube plugging at a rated thermal power of 2530 MWt, I
which is consistent with the Reference 3 analysis, i
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Table 2-1 Disposition of Events Sammary for Palisades
$Ri Sounding Updated Event F.ve, t Event or FSAR Classift-
& sig-cation pgign Magg Dissosition Reference Desienation 15.1 INCREAS'. IN HEAT ROIDWAL Bf THE SEC0 MARY SYSTEM 15.1.1 Decrease in feedwater Temperature Bourded 15.1.3 14.9.4 15.1.2 Increase in feedwater Flow 1)
Power Bounded 15.1.3 14.9.6 2)
Startup Bounded 15.1.3 14.9.5 15.1.3 Increase in Steam Flow Analyze 14.10 15.1.4 Inadvertent Opening of a Steam Gererator Relief of Safety Valve 1)
Power Bounded 15.1.3 2)
Scram Shutdown Margin Sounded 15.1.3 15.1.5 Steam System Piping Failures Inside and Outside of Containment Sounded Ref.!!,12&I3 14.14 15.2 DECREASE IN HEAT R[MOVAL BY THE SECOIGARY STEAM 15.2.1 Loss of External Load Analyze 14.12 15.2.2 Turbine Trip Sounded 15.2.1 15.2.3 Loss of Condenser vacuum Sounded 15.2.1 15.2.4 Closure of the Main Steam Isolation Valves (MSIVs)
Sounded 15.2.1 15.2.5 5 team Pressure Regulator Failure Not applicable;
- i ladR [ vent
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O Table 2-1 Disposition of Events Summary for Palisades (Cont.)
SRP Event Event Sounding updated Classift-Desig-Event or FSAR cation nation M
Disbosition Reference Desfanation 15.2.6 Loss of Monemergency A.C. Power Short term bounded 15.3.1 to the Station Auxillaries Long term bounded 15.2.7 15.2.7 Loss of Normal Feedwater Flow Bounded Ref. 3 14.13 l
15.2.8 feedwater System Pipe Breaks C.,oidown Sounded 15.1.5 leside and Outside Containment Heatup Sounded 15.2.7 15.3 OfCREAS[ IN REACIOR C00UWii SYSTEM FLOW
}
15.3.1 Loss of forced Reactor Coolant 1
Tiow Analyze 14.7 j
15.3.2 flow Controller Malfunction Not Applicable 14.7 15.3.3 Reactor Coolant Pump Rotor l
Seizure Analyze 14.7 15.3.4 Reactor Coolant Pump Shaf t Break Bounded 15.3.3 14.7
]
15.4 RIACTIVI1Y AND POWER DIS 1Ribui10N ANOM4tlES 1
)
15.4.1 Uncontrolled Control Rod Bank Withdraaval from a Subtritical or low Power Condition Analyze 14 2.2.2 15.4.2 Uncontrolled Control Rod Bank Withdrawal at Power Operation Conditions Analyze 14.2.2.3,
15.4.3 Control Rod Misoperation l
I)
Dropped Control Bank / Hod Analyze 14.4 2)
Dropped Part-tength Control Rod Bounded 15.4.3(1) 14.6
j 4
Table 2-1 Disposition of Events Sunnary for Palisades (Cont.)
1 i
SRP M ing Updated
[ vent Event Event or 15AR i
l Classifi-Desig-ration nation Em!g Dissesition Reference Desienatlos l
3) h1 positioning of the Part-Length Control Group Not Appilcable 14.6 4)
Statically Misaligseni Control Sod / Sank Analyze 5)
Single Control Rod Withdrawal Analyze Ref. 8 14.2.2.4 6)
Core Barrel failure Analyze 14.5 15.4.4 Startup of an inactive Looc Analyze 14.8 i
I 15.4.5 flow Controller Malfw.cs'on Not applicable; No flow Con-J troller 7
1 15.4.6 CVCS Malfunction that Results j
in a Decrease in the Soron Con-i centratton in the Reactor Coolant 1)
Rated and Power Analyze 14.3 i
)
Operation Conditions l
I 2)
Reacter Critical. Hot Analyze 14.3 l
Standby and Hot shutdown i
3)
Refueling Shutdown Con-7.nalyze 14.3 dition, Cold Shutdoun j
Condition and Refveliog I
Operation i
15.4.7 Inadvertent toading and Operation Administrative of a fuel Assembly in an Improper Procedores j
Position Preclude this Event
. 1
- - - - - - - - - - - - - - - -, - - - - -,, - - - = - - - - -
.o Table 2-1 Disposition of Events Summary for Palisades (Cont.)
[ vent Event Bounding Updated Event or FSAR Classifi-Desig-cation nation Nagg Disposition Reference Deslanation 15.4.8 Spectrum of Control Rod Ejection Analyze 14.16 Accidents 15.4.9 Spectrum of Rod Drop Accidents Not applicable; (BWR)
BWR Event 15.5 If4CRIASES IN RIACTOR COOLANT INVENIORY 15.5.1 Inadvertent Operation of the Overpressure ICCS that increases Reactor Bounded 15.2.1 Coolant Inventory Reactivity Bounded 15.4.6 15.5.2 CVCS Malfunction that in-Overpressure creases Reactor Coolant Bounded 15.2.1 Inventory Reactivity Bounded 15.4.6 15.6 D'CRI ASIS IN RIACIOR COOLANT INVENIORY 15.6.1 Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve Bounded 15.6.5 15.6.2 Radiological Consequences of the Bounded 15.6.5 lailure of Small Lines Carrying Primary Coolant Outside of Containment 15.6.3 Radiological Consequences of Bounded Ref. 8 14.15 Steam Generator tube failure 15.6.4 Radiological Consequences of a Not applicable; Main Steamline failure Outside BWR Event Containment
Iable 2-1 Dispositics of Eneras Summary for Palisades (Cont.)
SRP Bounding Updated
[ vent Event Event or f5AR Classifi-Desig-Disposition Reference Designation (at ion natiqn M192 15.6.5 Loss of Coolant Accidents Analyze Ref. 8,10, 14.17 20121 14.18 Resulting from a Spectrum of 14.22 Postulated Piping Breaks within the Reactor Coolant Pressure Boundary 15.7 RADIDAC11V[ R[ttASE fROM A SUB5YSTEM OR COMPONENT 15.1.1 Waste Gas System failure Deleted 14.21 15.7.2 Radioactive Liquid Waste System teak or f ailure (Release to Atmosphere)
Deleted 15.7.3 Postulated Radioactive Releases Bounded Ref. 8 14.20 due to Liquid-Containing Iank failures 15.7.4 Radiological Consequences of f uel Bounded Ref. 8 14.19 ilandling Accidents 15.1.5 Spent fuel Cask Drop Accidents Bounded Ref. 8 14.11
. this section of the Standard Review Plan has been deleted.
lhe results of the analysis of the large break LOCA are reported in Reference 10.
8
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ANF 88 108 l
Page 9 l
Table 2 : Sumary of Results MON 8R Event
,(Mil 15.1.3 Increase in Steam Flow (I) 1,46 i
15.2.1 Loss of External Load 1.71 15.3.1 Loss of Forced Reactor Coolant i
Flow 1.40
(
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15.3.3 Reactor Coolant Pump Rotor i
Seizure 1.28
\\
15.4.1 Uncontrolled Control Bank I
1.01(3)(5)
I Withdrawal at Subcritical or Low Power I
15.4.2UncontrolledContro};fank Withdr4wal at Power 1.25 f
15.4.3 Control Rod Misoperation(2) o Oropped Rod or Bank 1.25 o Single Rod Withdrawal (II 1.22 l
o Core Barre) Failure 1.25 15.4.6 CVCS Malfunction resulting in Decreased Boron Concentration (Adequacy of Shutdown Margin is i
Oemonstrated.)
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15.4.8 Control Rod Ejection
<1.17 'I
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(1) 100% power case j
(2) Results are based on conservative assumptions pertaining to control
{
rod / bank configurations.
(3) <2.9% of the core is calculated to experience ONB (4) <12.2% of the core is calculated to experience DNS j
j (5) Conservatively bounds Reactor Critical. Hot Star.dby and Hot Shutdown j
modes.
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,O ANF 88 108 Page 10 3.0 ANALYSIS OF PLANT TRANSIENTS l
This section provides the results of the event disposition and analyses performed to support the Palisades Cycle 8 operation.
Event numbering and nomenclature are consistent with the SRP to facilitate review.
l l
Reference 3 contains information on the plant licensing basis as it affects the event analyses including:
Classification of plant conditions Event acceptance criteria Single failure criteria Plant operating modes Analysis initial conditions Core and fuel design parameters Listings of systems and ccm,1onents available for accident mitigation, trip setpoints, time delays and component capacities.
These data, together with the design parametersM and the event specific input data given in Reference 3 and this report, represent a comprehensive sumary of analysis inputs.
The plant initial conditions, power distributions and neutronics data for Cycle 8 are given in Sections 15.0.1.
15.0.2 and 15.0.3, respectively.
Section 15.0.4 contains results of an analysis to verify the applicability of the TM/LP trip and the Inlet Temperature Limiting Condition of Operation LCO) given in Reference 3. to Cycle 8 operation.
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ANF 88 108 l
Page 11 l
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i 15.0 ACCIDENT ANALYSES J
f 15.0.1 PLANT INITIAL CONDITIONS l
The nominal plant rated operating conditions are presented in Table 15.0.1 I
1.
The uncertainties used in the accident analysis applicable to the l
operating conditions are:
)
Core Power t 2%
Primary Coalant Temperature 2 5'F l
Primary Coolant Pressure 50 psi Primary Coolant Flow 3%
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i ANF 88 108 Page 12 Table 15.0.1 1 Nominal Plant Operating Conditions j
l Core Thermal Power 2530 MWt f
Pump Thermal Power (total) 15 MWt j
System Pressure 2060 psia Vessel Coolant Flow Rate
- 120.3 M1bm/hr Core Coolant Flow Rate" 116.7 M1bm/hr Average Coolant Temperature 570.58'F j
I Core inlet Coolant Temperature 543.65'F
(
Steam Generator Pressure 730 psia l
Steam Flow Rate 10.97 M1bm/hr t
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Feedwater Temperature 435'F Number of Active Steam Generator Tubes
- j l
(per steam generator) 6023 l
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- Reflects a 3% bypass flow.
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_-._..____.-_._-_____.__________..______.__.-.-__-.__,__.1.-
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ANF 88 108 Page 13 15.0.2 POWER DISTRIBUTION The radial and axial power peaking factors used in the analysis are presented in Table 15.0.2 1.
Figures 15.0.21 and 15.0.2 2 show the limiting axial shapes for 100% power and 50% power, respectively.
These axial shapes have ASIS of 0.139 for 100% power and 0.342 for 50% power.
In this context. ASI is defined as:
I Plower Pupper Plower + Pupper P,,y corresponds to the power generated in the lower half of the core and g
P corresponds to the power generated in the upper half of the core.
Upper The Technical Specification (15) 1.imiting Condition of Operation. radial peaking limits are increased by 3.5% for Palisades Cycle S.
The increase in radial l
peaking is to accomodate a low radial leakage fuel loading pattern.
The The limiting DNBR occurs on an interior pin of an assembly with 208 rods.
Technical SpecificationO5) Limiting Conditions of Operation assure that the power distribution is maintained within these limits during normal operation.
However, some events analyzed result in transient redistribution of the radtal power peaking f actors.
Transient radial power redistribution is treated as described in Section 15.4.3.
The analyses in Reference 3 use an F factor that is 3% higher than that r
specified by the Technical Specifications.
This augmentation factor was used to account for the f act that the axial shapes eere derived from a one-For dimensional core physics reodel rather than a three dimensional model.
Cycle 8. minimum CNBR analyses were performed using axial shapes from both one dimensional and three dimensional core physics models.
Comparison of the minimum DNBRs indicates that the core averay axial shapes from the one-dimensional model are conservative relative to the het 4ssembly axial shapes i
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ANF 88 108 Page 14 from the three dimensional model, Thus, the F augmentation factor was r
unnecessarily conservative and is e tinated from the analyses supporting Cycle 8.
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- , R; ANF 88-108 Page 15 Table 15.0.2-1 Core Power Distribution
- Fuel Rods /Assembiv Radial Peaking Factor:
LQA 111 Peak interior rod 1.70 1.73 N
Engineering Uncertainty 1.d1
.LQ1 1.75 1.78 Total Radial, F r,T Axial Peaking Factor:
100% powar 1.39 50% power 1.67 Fraction of Power Deposited in Fuei 0.974 For power operation at less than
- rated, the radial peaking is
[
f r f<0.5, where f is the F
(1+0.3(1 f)] for 0.5sf11 and 1.15 Fr'T ffaltionalpowerof2530Mwt.
Proposed Technical Specification limit.
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.g ANF-88 108 Page 18 15.0.3 REACTIVITY CQEFFICIENTS USED IN THE SAFETY ANALYSIS Table 15.0.3-1 presents the reactivity coefficients for Cycle 8 and those used in the analysis in Reference 3.
As discussed in Reference 3, the set of parameters which most challenges the event acceptance criteria is used in each analysis.
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'0 ANF 88 108 Page 19 I
Table 15.0.3-1 Palisades Cycle 8 Reactivity Parameters IL13 BOC E0C Nominal Boundina Nominal Boundina Moderator Temp Coef, 10'4 as/'F 0.25 0.5
-2.81
-3.5 Doppler Temp Coef 10 5 ggf.F
-1.36 1.09 1.56 1.76 Moderator Pres Coef, 10-6 Ap/ psi 0.24
-1.0 2.66 7.0 Delayed Neutron Fraction 0.006 0.0075 0.0053 0.0045 Effective Neutron Lifetime, 10-6 seconds 21.6 41.9 24.6 19.9 238 U
Atoms Consumed per Total Atoms Fissioned
.665
.54
.695
.70 l
o k
l 6
e
--,-------,--e-n,,-n-gn
--.-+--~~-------,-e,~w r
I ANF 88-108 Page 20 I
I 15.0.4 TRIP SETPOINTS Reference 3 presents the trip setpoints, biases, and time delays used in the analysis.
The actual trip setpoints used in ev:h transient analysis were l
biased such that the acceptance criteria for each eunt is most challenged.
A new T LC0 and thermal margia/ low pressJre (TM/LP) trip were developed inlet for operation with the modified RPS.
Their development is presented in Reference 3.
The T LCO was used to dNelop the initial conditions used inlet in the transient analyses and the TM/LP trip was included in the transient analyses (3)
The following two sections cantain the results of an analysis to LCO and TM/LP are applicable to Cycle 8 operation.
verify that the Tinlet E
15.0.4.1 Inlet Temoerature Limitina Condition of Ooeration The T LC0 provides protection against penetriting DNB during limiting inlet The T LCO derived anticipated operational occurrence (A00) transients.
inlet I5 in Reference 3 is given below:
1 543.35 +.0575*(P-2060) + 5.0 x 10-5*(P 2060)2 T inlet
+ 1.173*(W-120)
.0102*(W 120)2 E
1800 s P s 2200 psia 100 s W s 130 Mlb/hr.
As shown in Table J-?, the most limiting A00 transient that does r.ot produce a g
The 5
reactor trip is the inadvertent drop of a full length control assembly.
T LCO must provide DNB protection for this transient assuming a return to inlet full power with enhanced peaking due to the anomalous control assembly insertion pattern.
The T LCO was verified for Cycle 8 using the XCOBRA.
inlet IllC computer code (6,16) with a conservative peaking augmentation factor.
i The XCOBRA l!!C calc.iations were run to demonstrate that the inlet tempera-I
ANF-88 108 Page 21 LCO results in a DNBR greater than 1.17 for the XNB ture allowed by the Tini t correlationO7,18) over a range of pressurizer pressures and primary coolant system flow rates.
These calculations were performed at 102 percent of rated power, i.e. 2530 MWt, and an axial shape with an axial shape index (ASI) of
.139.
Based on an analysis of axial shapes within the range of
.14 to
+.544, this was the limiting shape for full power transients for Cycle 8.
The derived T LCO supports operation at 100 percent of rated power for inlet measured plant Asis greater than
.08 and less than +.484.
This allows for a plant ASI measurement uncertainty of
.06.
The verification analysis includes the following 'incertainties and transient allowances:
2% power measurement uncertainty
.06 ASI measurement uncertainty 50 psia pressurizer pressure measuremer? uncertainty 7'F inlet temperature (5'F tilt allowance
+
2'F measurement uncertainty) i6% on the flow rate (3% bypass flow + 3% measurement uncertainty)
Transient allowances from Reference 3 for a dropped rod event: 65 psia decrease in the pressurizer pressure; a 4.7'F facrease in the inlet temperature; and an increase in the flow rate of 0.42 M1b/hr.
Applying these biases to the calculations resulted in a minimum DNBR greater LCO at than 1.17 for pressure and flow points within the range of the Tinlet full power.
In order that the plant can still operate should the measured ASI become less LCO equation was extended to a than 08 the applicability of the Tinlet measured ASI of
.30 at 70 percent of rated power.
This extended T L0 inlet range was verified to be applicable to Cycle 8 in the manner described above.
t ANF-88 108 Page 22 The limiting pait-power axial, shown in Figure 15.0.2-2 was used for these calculations.
15.0.4.2 Thermal Marcin/ Low Pressure (TM/LP) Trio The modified RpS includes the hardware for a new TM/LP trip which is to be installed at the Palisades reactor.
This new TM/LP is an improvement over the previous trip in that it allows monitoring of the core axial shape index.
The function of the TM/LP trip is to protect against slow heatup and depress'urization transient events.
In order to perform this function, the TM/LP trip must initiate a scram signal prior to exceeding the specified acceptable fuel design limits (SAFDLs) on departure from nucleate boiling (DNB) or before the average core exit temperature exceeds the saturation The SAFDL insures that there is no damage to the fuel rods and temperature.
the limit on core exit saturation is imposed to assure meaningful thermal power measurements.
The TM/LP trip works in conjunction with the other trips and the limiting conditions of operation (LCO) on control rod group position, radial peaking, and reactor coolant flow. The variable high power (VHP) trip is factored into the TM/LP development by limiting the maximum possible power that can be achieved at a particular radial peaking to 10% above the power corresponding to that radial peaking.
The LCO on the control rod group position is included in the TM/LP through monitoring of the axial shapes and the LCO on radial in the peaking is factored in by including its variation with power level TM/LP development.
Finally, the 1.C0 on reactor coolant flow is built into the TM/LP through the use of conservative flows throughout its development.
The development of the TM/LP trip setpoints are documented in Reference 3.
From Reference 3, the TM/LP trip is given as:
P
= 1563.7 (0A) (QR;) + 12.3 (T in) 6503 A var
. g
.s ANF 88-108 Page 23 where:
QRi = 0.412 (Q) + 0.588 Q s 1.0
- QR1=Q Q 2 1.0
- and, QA = +.226 (ASI) +.964
+.162 s ASI s +.544 QA =.521 (ASI) + 1.085
.156 s ASI s +.162 QA =
.691 (ASI) + 1.058
.653 s ASI s.156 This TM/LP is applicable over a pressure range from 1700 psia to 2300 psia and to a minimum measured HZP primary coolant flow rate of 124.3 Mlb/hr.
The TM/LP trip function was verified for Cycle 8 by first determining a set of limiting axial shapes.
The limiting 4xial shapes were determined in.06 ASI i
LC0(3)
The limiting increments covering the ASI range defined by the Tinlet axial shapes were used in the XC0 BRA IIIC model to ensure that the minimum DNBR allowed by the TM/LP trip function is greater than the XNB correlation (17,18) 95/95 limit of 1.17.
Thus, the TM/LP trip (3) is verified J
to be applicable over the possible range of axial shapes for Cycle 8.
i 1
r
t ANF-88 108 Page 24 15.0.5 DISPOSITION AND ANALYSIS OF EVENTS The following sections discuss the disposition and analysis of each of the Each event is numbered according to the corresponding SRP Chapter 15 events.
SRP designation.
The plant licensing basis, single failure criteria and acceptance. criteria are outlined in Reference 3.
4 I
4 4
e v
J i
f
.i a
I ANF-88 108 Page 25 15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.1.1 DECREASE IN FEE 0 WATER TEMPERATURE 15.1.1.1 Event Descriotion A decrease in feedwater temperature event may initiate due to the loss of one of several of the feedwater heaters.
This loss may be due to the loss of extraction steam flow from the turbine generator or due to an accidental opening of a feedwater heater bypass line.
The event results in a decrease of the secondary side enthalpy leading to an increase in the primary-to-secondary side heat transfer.
The steam generator outlet temperature on the primary side decreases causing the core inlet temperature to also decrease.
In the presence of a negative moderator coefficient, reduced core inlet temperature results in an increase in the core power and a decrease in thermal margin.
1 15.1.1.2 Event Discosition and Justification Reference 2 disposed this event as being bounded by the Increase in Steam Flow event (Event 15.1.3).
The changes for Cycle 8 do not change this disposition. Therefore, no further analysis is required for Cycle 8.
15.1.2 INCREASE IN FEE 0 WATER FLOW 15.1.2.1 Event Descriotion The Increase in Feedwater Flow event is initiated by a failure in the feedwater system.
The failure may be a result of: (1) a complete opening of a feedwater regulating valve; (2) over speed of the feedwater pumps with the feedwater valve in the manual position; (3) inadvertent startup of the second
ANF-88-108 Page 26 feedwater pump at low power; (4) startup of the auxiliary feedwater system; or, (5) inadvertent opening of the feedwater control valve bypass line.
The event results in an increase in the primary-to-secondary side heat transfer due to increased feedwater flow.
The steam generator outlet temperature on the primary side decreases causing the core inlet temperature to also decrease.
In the presence of a negative moderator coefficient, reduced core inlet temperature results in an increase in the core power and a decrease in thermal margin.
15.1.2.2 Event Disoosition and Justification Reference 2 disposed this event as being bounded by the Increase in Steam Flow event (Event 15.1.3).
The changes for Cycle 8 do not change this disposition. Therefore, no further analysis is required for Cycle 8.
15.1.3 INCREASE IN STEAM FLOW 15.1.3.1 Event Descriotion This event is initiated by a failure or misoperation of the main steam system The that results in an increase in steam flow from the steam generators.
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 down.
If the moderator temperature coefficient is negative, the cooldown of the primary system coolant would cause an insertion of positive reactivity and the potential erosion of thermal margin.
l l
, - ~,,
O b
ANF-88-108 Page 27 15.1.3.2 Event Disoosition and Justification This event was disposed to be analyzed for modified RPS operation for both HZP and HFP conditions (2)
The system response for both^ cases was evaluated using PTSPWR2(5) and the event minimum DNBR was calculated using XCOBRA-I!!CIO)
For the HZP case, the control rods were initially inserted in the PTSPWR2 simulation (3)
This eliminates the insertion of shutdown reactivity due to activt. tion of the reactor trip system.
The system response will remain the same for Cycle 8 as for the modified RPS analysis.
The increased radial peaking for Cycle 8 will change the thermal margin for this event.
The thermal margin for the Increase in Steam Flow event from HZP is, therefore, disposed to be reanalyzed for Cycle 8.
As was the case for the modified RPS analysis, the thermal margin for the HZP case will be analyzed O) using the Modified Barnett critical heat flux correlation For the Increase in Steam flow event from HFP, the reactor trip system acts to terminate the event.
From Reference 3, the variable high power and the TM/LP trips protect the plant from penetrating DNBR limits.
For an increase in radial peaking for Cycle 8, the primary system responso to an increase in steam fluw event will not change for the HFP case.
As in the HZP case, the increase in radial peaking will impact minimum DNBR.
Therefore, the Increase in Steam Flow event from HFP for Cycle 8 will be analyzed to calculate the minimum DNBR for this event.
15.1.3.3 Analysis and Results The minimum DNBR for this event initiated from full power occurred for a steam flow increase of about 112*,(3)
At this steam flow rate, the TM/LP and the variable high power trips coincide producing nearly simultaneous trip signals.
The junction of these two trips represents the worst possible DNB conditions, that is, maximum core power is attained combined with a low pressurizer
i i -
ANF-88 108 Page 28 The calculated minimum DNBR for Cycle 8 is 1.46.
The peak LHGR is pressure.
calculated to be 14.9 kW/ft.
For the hot shutdown case, the event was initiated by a rapid opening of the atmospheric dump valves and the turbine bypass valves resulting in a steam flow increase of 285. of the nominal full power steam flow.
A bounding value for the negative moderator temperature coefficient (EOC conditions) was assumed.
Due to the csoldown of the primary coolant, coupled with a negative moderator temperature coefficient, the reactor becomes critical resulting in a significant return to-power.
The Doppler temperature coefficient eventually The minimum critical heat flux ratio (CHFR) computed terminates this event.
for this case, using the Modified Barnett correlation, is 2.05.
The peak pellet LHGR is calculated to be 8.0 kW/ft.
15.1.3.4 Conclusion 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 of 1.17 and the minimum CHFR predicted for the hot shutdown case is greater than the Modified Barnett CHFR limit of 1.135.
The correlation limit assures that with 95f. probability and 95Y.
confidence, ONB is not expected to occur; therefore, no fuel is expected to fail.
The fuel centerline melt threshold of 21 kW/ft is not approached in this event.
INADVERTENT OPENING OF A STEAM GENERATOR RELIEF OR SAFETY VALVE 15.1.4 15.1.4.1 Event Descri 1!9D 2
This event is initiated by an increase in steam flow caused by the inadvertent opening of a steam generator relief or safety valve.
The increase in steam flow rate causes a mismatch between the heat generation rate on the primary side and the heat removal rate on the secondary side.
ANF 88 108 Page 29
[
15.1.4.2 Event Disoosition and Justification The increase in steam flow due to opening a steam generator valve is less than tnat considered in the Increase in Steam Flow event (Event 15.1.3)(2)
Therefore, an inadvertent opening of a steam generator relief or ' safety valve i
is bounded by Event 15.1.3(2)
This conclusion will not change for Cycle 8.
t 15.1.5 STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE OF CONTAINMENT 15.1.5.1 Event Descriotion A steam line piping failure event, or steam line break (SLB), is initiated by' a rupture of a main steam line pipe causing an uncontrolled steam release from the secondary system.
As a result of the uncontrolled release of steam, the heat extraction rate from the primary side is no longer equal to the core heat generation rate.
This power mismatch increases the primary-to-secondary side heat transfer and, consequently, reduces the primary side temperatures.
When this overcooling on the primary side is coupled with a negative moderator temperature coefficient, the shutdown margin after scram can potentially be eroded.
Such an erosion of shutdown margin may result in a return to power l
which, in turn, challenges thermal margin.
The consequences of this event are governed by the steam flow rate out of the ruptured steam line, the primary I
pump operating assumptions (i.e.,
with or without offsite power),
the magnitude of the moderator coefficient and the initial primary side operating state.
15.1.5.2 Event Diseosition and Justification For a steam generator tube plugging level of 29?., the SLB event was disposed as being bounded by previous analyses (2)
The SLB event for Cycle 8 is disposed to be bounded by the current analysis of record.
The conservatisms inherent in the SLB analysis with regard to the stuck rod and bounding i
h, e
1 4
ANF-88-108 Page 30 4
reactivity feedback are not significantly affected' t,y the cha....s for Cycle 8.
15.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2.1' LOSS OF EXTERNAL LOAD I
I 15.2.1.1 Event Descriotion A Loss of External Load event is initiated by either a loss of external electrical load or a turbine trip.
Upon either of these two conditions, the i
turbine stop valve is assumed to rapidly close (0.1 second).
Normally, a I
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 available);
Event 15.2.3 (Loss of Condenser Vacuum steam dump system unavailable); and, j
Event 15.2.4 (Closure of the MSIV-valve closure time is > 0.1 second),
i The Loss of External Load event primarily challenges the acceptance criteria on primary system overpressurization and DNBR.
The event results in an j
increase in the primary system temperatures due to an increase in the j
secondary side terrporature.
As the primary system temperatures increase, the l
i I
coolant expands into the pressurizer causing an increase in the pressurizer The primary system is protected against overpressurization by the pressure.
pressurizer safety and relief valves.
Pressure relief on the secondary side is afforded by the steam line safety / relief valves.
Actuation of the primary i
and secondary system safety valves limits the magnitude of the primary system l
temperature and pressure increase.
i With a positive moderator temperature coefficient, increasing primary system f
temperature results in an increase in core power.
The increasing primary side
ANF 88-108 Page'31 temperatures and power reduces the margin to ' thermal limits (i.e.,
ON8R limits) and challenges the ONBR acceptance critv la.
15.2.1.2 Event Disoosition and Justification The Loss of External Load from HFP was disposed to be analyzed for modified RPS operation (2)
The event initiated from full power bounds all other operating modes.
The system response for the ONBR and pressurization cases was evaluated using PTSPWR2(5) and the event minimum DNBR was calculated using XCOBRA-!!!C(0)
In the modified RPS analysis of the loss of External load pressurization case, the reactor trip system acts to terminate the event by activating a high pressurizer pressure trip signal (3)
For an increase in radial peaking for Cycle 8, the primary system pressure response to a loss of load will not change for the pressurization case.
Therefore, this case will not require reanalysis for Cycle 8 operation.
The increase in radial peaking for Cycle 8 will, however, impact minimum DNBR. Therefore, the loss of External Load eve.nt (minimum DNBR case) from HFP for Cycle 8 is disposed to be reanalyzed.
The event minimum DNBR will be calculated using XCOBRA IllC(0) with the core conditions taken from the limiting PTSPWR2(5) run for the modified RPS analysis.
15.2.1.3 Analysis and Results The transient response to a loss of External Load for the minimum DNBR case is given in Reference 3.
Using XCOBRA IllC(6), the minimum DNBR for Cycle 8 is computed to be 1.71. The peak pellet LHGR is calculated to be 13.5 kW/ft.
l 4
o ANF-88 108 Page 32 s
15.2.1.4 Conclusion The calculated minfmum DNBR for the event is above the XNB critical heat flux correlation safety limit, so the DNB SAFDL is not penetrated in this event.
Peak pellet LHGR for the event is well below the fuel centerline melt criterion of 21 kW/ft.
Applicable acceptance criteria for the event are therefore met.
15.2.2 TURBINE TRIR 15.2.2.1 Event Descriotion This event is initiated by a turbine trip which results in the rapid closure of the turbine stop valves.
A reactor trip would occur on a turbine trip and the steam dump system would operate to mitigate the consequences of this event.
The primary system is protected against overpressurization by the pressurizer safety and relief valves.
Pressure relief on the secondary side is afforded by the steam line safety / relief valves.
15.2.2.2 Event Discosition and Jostification The assumptions made in the Loss of External Load event (Event 15.2.1) bound the consequences of a Turbine Trip event.
Specifically, the loss of External load event considers the following:
a conservatively fast turbine stop valve closure time; reactor trip does not occur on a turbine trip; and, the atmospheric dump valves are assumed to be unavailable.
The Turbine Trip event was disposed as being bounded by the Loss of External l
load event (Event 15.2.1) ~ for modified RPS operation (2)
The changes for Cycle 8 will not invalidate this disposition.
ANF 88 108 Page 33 15.2.3 LOSS OF CONDENSER VACUUM 15.2.3.1 Event Descriotion l
This event is initiated by a reduction in the circulating water flow or an increase in the circulating water temperature which can impact the condenser back pressure.
This c.ondition can result in a turbine trip without the availability of steam bypass to the condenser.
The primary system is protected against overpressurization by the pressurizer safety and relief valves.
Pressure relief on the secondary side is afforded by the steam line safety / relief valves.
15.2.3.2 Event Discosition and Justification The assumptions made in the Loss of External Load event bound the consequences of a loss of Condenser Vacuum transient.
The Loss of Condenser Vacuum event was disposed as being bounded by the Loss of External Load event (Event 15.2.1) for rated power and power operating modes (2)
The scenario of this event from other operating modes allows sufficient time for the operator to control the primary and secondary system temperatures (2)
These conclusions will not change for Cycle 8.
t 15.2.4 CLOSURE OF THE MAIN STEAM ISOLATION VALVES (MSIV) fBWR1 15.2.4 Event Descriotion Closure of the Main Steam Isolation Valve event is initiated by the loss of control air to the MSIV operator.
The valves are swinging check valves designed to fail in the closed position.
The inadvertent closure of the MSIVs is primarily a BWR event, however, the closure of these valves in a PWR can drastically reduce the steam load, j
i
?
ANF 88 108 Page 34
'8' 15.2.4.2 Event Discosition and Justification The closure time of the HSIVs is less than 5 seconds, but greater than the value used in Event 15.2.1 (0.1 seconds).
A MSIV closure event will progress in a similar, fashion as a loss of External Load (Event 15.2.1), but at a slower rate.
The consequences of Event 15.2.1 will bound those for Event 15.2.4 because of the more rapid valve closure time (2)
Sinss the changes made for Cycle 8 will not impact the system response, Event 15.2.4 will continue to be bounded by Event 15.2.1.
15.2.5 STEAM PRESSURE REGULATOR FAllVRE Palisades does not have steam pressure regulators.
Therefore, the Steam Pressure Regulator Failure event is not considered in this analysis.
15.2.6 LQ1LOF NONEMERGENCY A.C. POWER T0 THE STATION AUXILI ARIES 15.2.6.1 Event Descriotion A Loss of Nonemergency A.C. Power to Station Auxiliaries event may be caused by a complete loss of the offsite grid together with a turbine generator trip l
or by a failure in the onsite A.C. power distribution system.
The loss of A.C. power may result in the loss of power to the primary coolant pumps and the main feedwater purnps.
The combination of the decrease in primary coolant flow rate, the cessation of main feedwater flow and trip of l
The decrease of both the turbine generator compounds the event consequences.
primary coolant flow and main feedwater decreases the primary to secondary system heat transfer resulting in the heatup of the primary system coolant.
The increase in primary system coolant temperature increases the overpressurization potential and increases the threat of penetrating DNB.
r ANF 88 108 Page 35 The event is most limiting when initiated from full power conditions.
During this mode of operation the amount of stored heat in the fuel rods is the greatest and the margin to CNB is minimized.
15.2.6.2 Event Discosition and Justification This event can be separated into two distinct phases: the near-term and the long term.
The near-term phase is characterized by the loss of power resulting in the coastdown of the primary coolant pumps, the coastdown of the main feedwater pumps and the trip of the turbine generator.
The coastdown of the primary coolant pumps causes an immediate reduction in thermal margin.
The trip of the reactor and the subsequent insertion of control rods terminates the challenge to ONB limits.
The near term ph'ase of th event is similar to that of a loss of Forced Reactor Coolant Flow transient (Event 15.3.1).
The near term consequences of this event are addressed in the analysis of Event 15.3.1(3)
The long term consequences of a Loss of A.C. Power event are determined by the heat removal capacity of the auxiliary feedwater system.
The long term portion is similar to the Loss of Normal Feedwater Flow transient (Ever.t 15.2.7).
The long term effects are, therefore, addressed by the analysis of the Loss of Normal Feedwater Flow event (3)
The changes for Cycle 8 will not alter this conclusion.
15.2.7 LOSS OF NCRMAL FEE 0 WATER FLOW 15.2.7.1 Event Descriotion A Loss of Ncrmal Feedwater Flow transisnt is initiated by the trip of the main feedwater pumps or a malfunction in the feedwater contrni valves.
The loss of main feedwater flow decreases the amount of subcooling in the secondary side downcomer which diminishes the primary to secondary system
C ANF 88 108 Page 36 i
heat transfer and leads to ar. increase in the primary system coolant temperature.
As the primary system temperatures increase, the coolant
[
expands into the pressurizer which increases the pressure by compressing the l
t steam volume.
i The opening of the secondary side safety valves controls the heatup of the f
primary-side.
The long term cooling of the primary system is governed by the l
heat removal capacity of the auxiliary feedwater flow.
The auxiliary feedwater pumps are automatically started upon a steam generator low liquid level signal.
I 15.2.7.2 Event Diseosition and Justification A Loss of Normal Feedwater Flow event is only credible for rated power and power operating conditions (2)
The worst consequences occur when the fewdwater is lost during rated power operation since more stored heat is contained in the fuel than in other modes of operation.
The short term impacts of the t.oss of Normal Feedwater Flow event challenges The the ONB and the primary system overpressurization acceptance criteria.
ONB challenge is maximized when it is assumed' that offaite power is lost F
causing the primary coolant pumps to coastdown.
The Loss of Forced Reactor Coolant Flow event (Event 15.3.1) addresses the short term DNB consequences of f
a Loss of Normal Feedwater Flow transient.
After the reactor trip system is j
activated, the core power is drastically reduced alleviating the challenge to DNS.
I The long term effects of this event primarily challanges the pressurization limits of the primary system due to the filling of the pressurizer and steam generator dryout.
If the pressurizer were to fill c4 Ately solid with liquid, the primary system pressure control would be lost and primary liquid would be expelled through the pressurizer safety valves.
ANF 88 108 Page 37 The dryout of a steam generator causes the loss of a primary to-secondary system heat sink exacerbating the primary side heatup.
The long term consequences of a Loss of Normal Feedwater Flow event were analy.ted in f
Reference 2.
L The changes for Cycle 8, will not impact the system response to a loss of Normal Feedwater Flow.
The DNS challenge is addressed in the analysis of the loss of Forced Reactor Coolant Flow event (Event 15.3.1).
The primary system pressurization and pressurizer fill cases will not be impacted.
Therefore, this event is disposed as being bounded by the modified RPS analysis for the pressurization, steam generator dryout and pressurizer fill casesI3I.The ONB case is bounded by the loss of Forced Reactor Flow event (Event 15.3.1).
15.2.8 FEEDWATER SYSTEM Pipe BREAXS INSIDE AND OUTSIDE CONTAINMENT 15.2.8.1 Event Descriotiga A Feedwater System Pipe Break event occurs when a main feedwater system pipe is ruptured.
The ruptured pipe will cause a blowdown of the affected steam generator if the break occurs upstream of the feedline check valve. If the rupture occurs downstream of the check valve, the event would behave much like the loss of Normal Feedwater Flow transient.
Since the auxiliary feedwater flow is injected into the steam generators via a separate piping n6cnorc than the main feedwater, the delivery of auxiliary feedwater will not be interrupted by the pipe rupture.
The event results in both a primary system cooldown and a heatup.
Initially, the event results in a cooldown of the primary side coolant due to the energy removal during the blowdown stage of the event.
The eventual depletion of secondary side inventory and lack of main feedwater will cause the primary system to heatup much like a loss of Normal Feedwater Flow event.
ANF-88 108 Page 38 15.2.8.2 Event Discosition and Justification The event was disposed in Reference 2 as being bounded during rated power operation as follows:
1.
The cooldown aspect of the event is bounded by the Steam Line Break event (Event 15.1.5).
2.
The heatup effects are bounded by the loss of External Load event (Event 15.2.1) for the primary system overpressurization and the Loss of Normal Feedwater Flow event (Event 15.2.7) for the long term cooling requirements.
Feedwater pipe breaks from modes other than rated power result in a primary system cooldown and are bounded by the Steam Line Break accident (Event 15.2.8).
The changes for Cycle 8 will not impact the syste, response to a Feedwater System Pipe Break event.
Therefore, this event is disposed as being bounded as described above.
15.3 QE.REASE IN REACTOR COOLANT SYSTEM FLOW 15.3.1 LOSS OF FORCEO REACTOR COOLANT FLOW 15.3.1.1 Event Descriotion The Loss of Forced Reactor Coolant Flow transient is initiated by a disruption of the electrical power supplied to or a mechanical failure in a These f ailures may result in a complete primary coolant system (PCS) pump.
or partial loss of forced coolant flow.
The impact of losing a PCS pump or pumps is a decrease in the active flow rate in the reactor core and, consequently, an increase in core temperatures.
A
O i
ANF 88 108 Page 39 Prior to reactor trip, the combination of decreased flow and increased The event is terminated by the temperature poses a challenge to DNB limits.
PCS low flow trip.
15 3.1..
' 'iscosition and Justification De -
.sg scenario for this event is to initiate the loss of four PCS N)
Plant operation with a reduced low ot a
ated power condition trip setpoint (60?. of rated four PCS flow) for three PCS pump N)
This or reduced power (39?. of rated) has been justified v.
or g state is allowed for a limited period of time for repai r/ pump to provide for an orderly shutdown, or to provided for the conduct
- startup, of reactor internals noise monitoring test measurements.
For Cycle 8 operation, the increase in radial peaking will impact the minimum OflBR.
To assess the minimum DNBR for Cycle 8 operation, the minimum ONBR calculation will be reanalyzed for the loss of four PCS pumps from rated power.
from a The calculated minimum OfiBR for a loss of Forced Coolant Flow event three primary coolant pump initi:1 condition is bounded by the results of the rated power event.N) 15.3.1.3 Analysis and Results The transient is initiated by tripping all four primary coolant pumps.
As the pumps coast down, the core flow is reduced, causing a reactor scram on low This increase flow.
As the flow coasts down, primary temperatures increase.
in temperature causes a subsequent power rise due to moderator reactivity feedback.
The primary challenge to Of48 is from the decreasing flow rate and resultir.g increase in coolant temperatures.
Using XCOBRA.!!!C, the minimum Ot4BR for Cycle 8 is computed as 1.40.
The peak pellet LHGR is calculated to be 13 1 kW/ft.
ANF-88-108 Page 40 15.3.1.4 Conclusion The XNB critical heat flux safety correlation limit of 1.17 is not penetrated, so event results are acceptable with respect to the ONBR SAFDL. Maximum peak pellet LHGR for this event is below the incipient fuel centerline melt criterion of 21 kW/ft.
Applicable acceptance criteria for the event are therefore met for Cycle 8.
15.3.2 FLOW CONTROLLER MALFUNCTION There are no flow controllers on the PCS at Palisades.
Tnerefore, this event is not credible.
15.3.3 REACTOR COOLANT PUMP ROTOR SEIZURE 15.3.3.1 Event Descriotion The seizure causes This event is initiated by a seizure of a PCS pump rotor.
an immediate reduction in PCS flow rate.
As in the Loss of Forced Coolant is a decrease in Flow event (Event 15.3.1), the impact of losing a PCS pump the active flow rate in the reactor core and, consequently, an increase in Prior to reactor trip, the combination of decreased flow core temperatures.
and increased temperature poses a challenge to ONB limits.
The event is terminated by the PCS low flow trip.
15.3.3.2 Event Diseosition and Justification The most limiting scenario for a Reactor Coolant Pump Seizure event occurs for rat 2d power or power operating conditionsI2I.
Plant operation with a reduced low flow reactor trip setpoint (607. of rated four PCS flow) for three Results of PCS pump operation at reduced power was justified in Reference 7.
the three PCS pump case from reduced power were bounded by the event initiated
6 ANF 88-108 Page 41 from rated power (II.
For Cycle 8 operation, the increase in radial peaking impacts the minimum DNBR.
To assess the minimum ONBR for Cycle 8 operation, the minimum DNBR calculation will be reanalyzed for a pump rotor seizure from rated power conditions.
This event initiated from three PCS pump operation at reduced power will remain bounded by the full power event for Cycle 8.
15.3.3.3 Analysis and Results The first locked rotor case is analyzed using the calculated value of core flow.
Assuming the locked pump loss coefficient given by the homologous curves at zero pump speed, the core flow is 78% of the nominal full power, four pump operation value.
The second case is analyzed at 74.7% flow as specified in the Technical Specifications (Reference 15, page 2 7).
The XCOBRA !!!C calculated minimum ONBRs are 1.35 and 1.28 for Case 1 and Case 2, respectively. The peak pellet LHGR for each case is 13.1 kW/ft.
15.3.3.4 Conclusion The XNB critical heat flux correlation safety limit of 1.17 is not penetrated and no fuel failures are expected for this infrequent event.
- Thus, applicable acceptance criteria for this event are met for Cycle 8.
15.3.4 REACTOR COOLANT PUMP SHAFT BREAX 15.3.4.1 Event Descriotion
'his event is initiated by a failure of a PCS pump shaft resulting in a free-heeling impeller.
The impact of a :oolant pump shaf t break is a loss of pumping power from the affected pump and a reduction in the pCS flow rate.
The flow reduction due to the seizure of a pump rotor is more severe than that for a shaft breakt however, the potential for flow reversal is greater for the
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l ANF 88 108 Page 42
/
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I shaft break event.
The event is terminated by the low reactor coolant flow trip.
15.3.4.2 Event Discosition and Justification I
The event is most limiting at rated power conditions because of a minimum margin to DNBR limits.
The initial flow reduction for this event is bounded The by that for the Deactor Ccolant Pump Rotor Seizure event (Event 15.3.3).
potential for e,rsater reverse flow due to a shaft break is accounted for in the seized rotor analysis oy decreasing, internally in PTSPWR2(5), the rotor inertia to zero at the time of predicted reversed flow.
f The changes made for Cycle 8 will not impact the system response to a PCS pump shaft break.
The impact to minimum DNBR is bounded by the analysis of Event 15.3.3.
Therefore, this event is disposed as being bounded.
15 1 REACTIVITY AND POWER DISTRIBUTION ANOMAlfES 15.4.1 UNCONTROLLED CONTROL R00 ASSEMBLY (CRA) WITHORAWAL FROM A SUBCRITICAL OR LOW POWER STARTUP CONDITION I
15.4.1.1 Event Descriotion
(
This event is commenced by an uncontrolled withdrawal of a control rod bank.
j l
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 withdrawal at suberitical or low initial power levels.
As the control bank is withdrawn, the positive reactivity insertion causes a significant core power increase as the reactor approaches prompt criticality.
As the core power increases, the core average and hot leg temperatures also increase.
Due to the increasing power and temperatures, the ONB limits are challenged.
An additional assumption included in the event analysis for
s ANF 88 108 Page 43 I
I3) modified RPS operation is that the plant is operating with three PCS pumps The transient eventually terminates on an overpower reactor trip signal.
15.4.1.2 Event Discositten and Justification Fcr Cycle 8 operation, the changes to radial peaking will impact the minimum DN8R for this event.
The system response to this event will, however, not be affected.
To assess the minimum DNBR for Cycle 8 operation, the minimum DNBR calculation will be reanalyzed.
15.4.1.3 Analysis and Results This event was analyzed assuming three primary coolant pumps to be operating.
The event is initiated with control bank withdrawal.
The minimum DNBR calculated for the event is 1.01, which is below the 1.17 95/95 DNS safety iimit for the XNB critical heat flux correlation.
The percent of the core l
experiencing boiling transition was calculated to be less than 2.9% for Cycle 8,
as compared to less than 2.3% for the Reference 3 analysis.
Due to conservative assiimptions in the fuel failure calculation, the offsite radiological doses for the uncontrolled bank withdrawal from low power are less than 10% of the 10 CFR 100 limits for Cycle S.
15.4.1.4 Conclusions In this infrequent event, only a small fraction of the core is calculated to experience boiling transition.
Possible radiological releases are less than 10% of the 10 CFR 100 guidelines.
Therefore, this event meets the applicable ac eptance criteria for Cycle 8 operation.
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ANF 88 108 rage 44 l
UNCONTROLLED CONTROL R00 BANK WITH0RAWat AT POWER' 15.4.2 i
i 15.4.2.1 Event Descriotion As with Event 15.4.1, this event is initiated by an uncontrolled withdrawal 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 and operating initial power levels.
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As the control bank is withdrawn, the positive reactivity insertion causes an i
Due to the increase in core power and in primary coolant system temperatures.
In most increasing power and temperatures, the ONB limits are challenged.
cases, the transient will terminate on a variable high power, a TM/LP or a high pressurizer pressure trip; however.. some cases do not activate a reactor protection system trip.
t
)
15.4.0.2 Event Diseosition and Justification W
evaluates the f
The analysis performed for modified RPS operation l
consequences of an uncontrolled rod withdrawal from both rated power and 50f.
t of rated power initial states.
A spectrum of reactivity insertion rates were j
evaluated in order to bound events ranging from boron dilutions to fast
(
l control bank withdrawals.
1, impact DNBR for both the full and The changes for Cycle 8 operation will To assess the minimum DNBR for Cycle 8 ope,'ation, the part-power cases.
l respective limiting minimum DNBR point for 50f. and 1007. power conditions are t
reanalyzed for Cycle 8.
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j ANF 88 108 Page 45 15.4.2.3 Analysis and Results The uncontrolled rod withdrawal transients were analyzed for full power (100?.
of rated) and mid power (50% of rated).
The calculated minimum DNBR occurred for a rod withdrawal from 100% of rated thermal power.
The mid power case series was, in generil, less limiting than the full power cases.
The limiting rod withdrawal at 50% power and EOC kinetics occurred at an insertion rate of 3 x 10 5 Ap/sec.
The ininimum DNBR was calculated as 2.36.
This transient did not scram, but was ended when the rods were fully withdrawn.
The peak pellet LHGR for the 50*. power case is calculated to be 10.3 kW/ft.
The limiting uncontrolled control rod bank withdrawal at 100% power and EOC kinetics occurred at er, insertion rate of 17.0 x 10 5 Ap/sec.
The minimum DNBR was calculated at 1.25.
This transient tripped on a thermal margin / low pressure signal.
The peak pellet LHGR for the 100?. power case is calculated to be 14.8 kW/ft.
15.4.2.4 conclusion Reactivity insertion transient calculations demonstrate that tha XNB correlat'on limit of 1.17 will not be penetrated during any credible The maximum peck reactivity insertion transient at full power or mid power.
fuel pellet linear heat rate for these events is well below the incipient centerline melt criterion of 21 kw/f t.
Applicable acceptance criteria are therefore met for Cycle 8,
and the adequate functioning of the thermal margin / low pressure trip demonstrated.
l
ANF 88 108 Page 46 15.4.3 CONTROL R00 MIS 0pERATION The control rod misoperation event considers a number of different event initiators. These include:
(1) Dropped control rod or bank; (2) Dropped part length control rod; (3) Ma1 positioning of a part lingth control rod group; (4) Statically misaligned control rod or bank; (5) Single control rod withdrawal; (6) Core barrel failure.
Each of the above events includes a redistribution of power which leads to a l
local augmentation of the peaking factor in the affected region of the core.
15.4.3.1 Event Des.:cietion i
(1) Droceed Control Red / Bank i
A control rod drop event is initiated by a de energized control rod drive l
mechanism (CRDM) or another failure in the control W system.
With the f
insertion of negative reactivity due to the dropped rod, the coro power Moderator and Doppler temperature feedback, driven by a constant l
decreases.
A l
turbine generator load, cause the power to increase to its initial state.
localized increase in the radial peaking results from power redistribution due This event is a challenge to ONB limits because of radial to the dropped rod.
l peaking augmentation together with near full power operating conditions.
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ANF 88 108 Page 47 (2) Droceed Part Lenoth Control Rod Part iangth control rods are not used during power operation and are maintained in a withdrawn state.
A failure of the rod brake mechanism could result in a part-length control red drop.
(3) Maloositionina of a Part Lenath Control Rod Grouc Use of part length control rods is not allowed during power operation.
The part-length control rods are maintained in a fully withdrawn state therefore, this event is not credible.
(4)
Statically Misalianed Control Red / Bank A static misalignment occurs when a malfunction in the CROM causes a control rod to be out of alignment with its bank or a control group to be in violation of its Power Dependent Insertion limits (POILs).
In the case of a static misalignment of a control rod, one control rod is positioned out of the core while the balance of the control bank is inserted.
This situation causes a localized increase in radial peaking in the affected region of the core.
The increased radial peaking, together with the initial core power level, can significantly reduce the margin to DNB.
The reverse condition, i.e. one control red fully inserted with its bank fully withdrawn, is essentially the same as a dropned control rod event.
I (5)
Sinale Control Rod Withdrawal l
The withdrawal of a single control rod results in a reactivity insertion and a localized increase in radial peaking.
The degradation of core conditions characteristic of a reactivity insertion transient, combined with an increase i
in local radial peaking, poses a challenge to CNBR limits, t
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ANF-88 108 Page 48 (6) Core Barrel Failure This event is initiated by the circumferential rupture of the core support barrel.
The core stop supports serve to support the barrel and the reactor core by transmitting all loads directly to the vessel.
The clearance between the core barrel and the supports is approximately one half inch at operating The worst possible axial location of the barrel rupture is at temperatures.
the midplane of the vessel nozzle penetrt.tions so that a direct flow path is formed between the inlet and exit nozzles in parallel with the path that goes The core sustains a small reactivity transient induced by through the core.
the motion o' the core rehtive to the inserted rod bank (s).
Reactor protection for the Core Barrel Failure event during hot shutdown, refueling shutdown, cold shutdown, and refueling operating conditions is For the provided by Technical Specification Shutdown Margin requirements.
reactor critical and hot standby operating conditions, reactor protection is l
provided by the variable overpower trip and a nonsafety grade high rate of-change of power trip.
For the rated oower nd power operating conditions, reactor protection is afforded for the variable overpower and thermal margin / low pressure trip.
15.4.3.2 Event Diseosition and Justification (1) Droceed Control Red / Bank 1
I The analysis supporting modified RPS operation evaluates the consequences of i
from rated power conditions (3)
A control bank drop causes a this event variable high power trip and, therefore, does not Mse a challenge to DNB limits.
The minimum DNBR for a control rod drop event from full power was analyzed for modified RpS operation.
For Cycle 8 operation, the minimum DNBR for the control rod drop event is disposed to be analyzed at rated power and full flow with increased radial
ANF 88 108 Psge 49 L
The system response due to a control bank drop will not vary for peaking.
Cycle 8 as compared to the analysis suppcrting modified RPS opera (2) Droceed Part lenath Control Rod A dropped part-length control rod till not be as severe as a dropped full-15.a.3(1)(2)
This length control rod and is, therefore, bounded by Event conclusion will not change for Cycle 8.
(3) Malcositionica of the part Lenath Control Rod Groun The Use of part-length control rods is not allowed during power operation.
part length control rods are maintained in a
fully withdrawn state; therefore, this event is not credible.
(4) Statically Misalianed Control Rod / Bank Reference 2 disposed the misaligned control rod event to be analyzed for l
The modified RPS analysis considered this event at an modified RPS operation.
initial full power operating condition with one control rod fully withdrawn and its control bank inserted beyond the appropriate PDIL(3)
The modified f
RPS analysis consists of an XCOBRA-!!!C calculation at full power conditions with a limiting assembly radial peaking augmantation factor.
For the statically misaligned control bank at rated power, the statically c
conditions (2) l control rod reaches the same steady-state misaligneo Therefore, the results for the Cycle 8 reanalysis of a misaligned control rod l
also apply to the misaligned control bank event at rated power.
I 1
are inserted in the for power operating conditions, control banks 3 anc 4 The control bank misalignment j
core for power levels of 35". to 657. of rated.
event was disposed to be reanalyzed to support modified RPS opera 507, and 65". of rated r
The analysis consists of XCOBRA IIIC calculations at i
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I ANF 88 108 Page 50 I
power conditions.
Each calculation includes a limiting assembly radial peaking augmentation factor.
For Cycle 8 operation, the increase in radial peaking necessitates the reanalysis of minimum DNBR for both the 50% and 65% power cases with four PCP l
flow.
(5) 11ngig Control Rod Withdrawal This event was disposed to be analyzed for both rated power and power operating conditions (2)
The analysis performed for modified RPS operation evaluates the consequences of single rod withdrawal from both 50% and 100%
g rated power initial conditions.
A numbed of reactivity insertion rates were 3
The PTSPWR2 evaluated to bound the minimum insertion r ates for this event.
portion of the analysis of a single control rod wit
- drawal is a continuation of the respective reactivity insertion rate curves generated for Event 15.4.2(3)
For Cycle 8 operation, the increased radial peaking will impact DNBR for the To assess the minimum DNBR for Cycle 8 operation, 50*. and 100% power cases.
the limiting DNBR cases will be reanalyzed under Cycle 8 conditions.
(6) Core Barrel Failure The probability of a circumferential rupture of the core support barrel has the same low probability of occurrence as a major rupture of the primary Therefore, this event is classified as a Limiting Fault event system piping.
with the corresponding acceptance criteria.
The acceptance criteria are given in Reference 3.
Reference 2 disposed this event not to be credible during hot shutdown, refueling
- shutdown, cold shutdown and refueling operation due tu the Technical Specification shutdown margin requirements.
The event initiated I
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e ANF-88-108 Page 51 from rated power bounds the power operating, reactor critical and hot standby operating modes.
For rated power, the FSAR analysis (8) is bounding due to a conservatively high reactivity insertion.
?
For the conditions assumed in the analysis supporting modified RPS operation, the maximum reactivity insertion at rated power with the control rods at their PDILs is less than the rekctivity insertion for the FSAR analysis. Reference 3, therefore, disposed this event to be bounded by the FSAR analysis (8) i For Cycle 8,
however, the increase in radial peaking necessitates the reanalysis of the minimum DNBR for the Core Barrel Failure event at rated power.
15.4.3.4 Analysis and Results i
I Calculated minimum DNBRs and peak pellet LHGRs are given in Table 15.4.31 for the Control Rod Misoperation events.
l Radial peaking augmentation factors for dropped control rod / bank events, j
static misalignment events and single control rod withdrawal events are
[
calculated at full power for different exposure conditions.
The radial peaking augmentation factors used in the Reference 3 analysis were verified to remain conservatively applicable to Cycle 8.
Control red and bank worth for Cycle 8 were verified to be bounded by the values used in the Reference 3 analysis.
Out to the motion of the core relative to the control red positions, a small i
reactivity insertion is experienced for the Core Barrel Failure event.
The I
maximum distance the core barrel may fall is 0.547 inches (8) at hot full power.
A conservatively high reactivity insertion rate is used in the i
analysis of minimum ONBR.
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ANF-88 108 Page 52 The amount of coolant flow that bypasses the reactor core increases as a result cf a failure of the core barrel.
A parallel flow path between the inlet and exit nozzles can potentially occur.
To account for the increase in core bypass flow, the total PCS flow rate is reduced by 107.(8)
The minimum DNBR for the Core Barrel Failure event is 1.25 for Cycle 8, as calculated using the XNB correlation.
Therefors, because the minimum DN9R is greater than the 95/95 limit of 1.17, no fuel failures would be expected for this Limiting F a '. ' t event.
Overpressurization of the primary system is bounded by the results of the Control Rod Ejection event (Event 15.4.8).
15.4.3.5 Conclusion The moderate frecency events result in minimum DNBRs greater than the XNB critical heat flux correlation safety limit.
Thus, the ONBR SAFDL is not penetrated.
The maximum peak linear heat rate for these events is below the fuel centerline melt criterion of 21 kw/f t.
For the Core Barrel Failure event, the minimum DNBR is greater than the XNB critical heat flux correlation safety limit.
Thus, the DNBR SAFDL is not penetrated and no fuel failur-S are predicted to occur.
Applicable acceptance criteiia for these events are therefore met for Palisades Cycle 8 operation.
I
l-ANF 88 108 Page 53 Table 15.4.31 Summary of MONBRs for Control Rod Misoperation Events Operating Maximum hini (Power)
Mot 1988 LHGR (kW/ft)
Dropped Control Rod (100%)
1 1.25 15.6 Statically Misa11gned Control Rod (100%)
Bounded (Dropped Rod)
Statically Misaligned Bank (50%)
2 2.79 10.0 Statically Misaligned Bank (65%)
2 2.08 12.3 Rod Withdrawal (100%)
1 1.22 15.1 Rod Withdrawal (50%)
2 1.59 13.3 Rod Withdrawal (10'4 )
3 Bounded (15.4.1)
Rod Withdrawal (10'4 )
4 Bounded (15.4.1)
Rod Withdrawal (i 10'4 )
5 Subcritical Core Barrel Failure (100%)
1 1.25
- These modes are defined in Reference 3.
- The Core Barrel Failure transient is classified as a Limiting Fault event.
l ANF 88 108 Page 54 15.4.4 STARTUP OF AN INACTIVE LOOP 15.4.4.1 Event Descr,4 112D 9
u This event is initiated by the startup of an inactive primary coolant pump.
The startup of an inactive pump can lead to an introduction of colder primary coolant into the reactor core. The lower coolant temperature, together with a negative moderator temperature coefficient, can cause an increase in core power and a degradation of DNB margin.
Sufficient protection is available to reduce the consequences of this event.
15.4.4.2 Event Oiscosition and Justification A Startup of an Inactive Loop is classified as a Moderate Frequency event with l
l the corresponding acceptance criteria. The acceptance. criteria for this class l
of event are given in Reference 3.
Reference 3 disposed this event to be bounded by the FSAR analysisI8) for the l
[
analysis supporting modified RPS operation.
f For operation with one inoperative pump, the low flow trip setpoint and the variable overpower trip setpoint are simultaneously changed to the allowable l
values for the selected pump condition.
Under this arrangement, the variable overpower tr'p will terminate any transient resulting from the inadvertent j
activation of an idle pump before any significant. decrease in thermal margin.
For Palisades, this event is most limiting for an initial condition of three operating primary coolant pumps with the corresponding reduced power level and variable high power trip setpoint.
Continuous power operation with less than i
four primary coolant pumps is not allowed by the Technical Specifications.
Additionally, startup of an inactive primary coolant pump when operating above l
I hot shutdown is not allowed.
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ANF 88 108 page 55 Due to the changes for Cycle 8, the ONBR will be analyzed with an increase in radial peaking for this event.
15.4.4.3 Results of Analysis As part of the modified RPS, a variable high power trip is to be added.
This trip will cause a reactor trip when the reactor power increases to a power level 10% above the current power level.
This trip will provide the required protection to mitigate the consequences of an Idle Loop Startup transient.
For power operation with three pumps in service, the variable high power trip setpoint has a maximum value of 49% of rated power, which is 10% above the maximum allowed operating power level of 39% of rated.
When a primary pump is removed from service, the thermal power is reduced in accordance with the Technical Specifications.
Because of the reduced variable nigh power trip settings, the maximum nominal reactor power for threa pump operation without trip is less than 49% of rated, or 39% maximum operating l
power level plus a 10% margin to trip.
Including a trip uncertainty of 5.5%(3), the maximum attainable power for three pump operation is 54.5% of j
rated without causing a reactor trip.
Although a slight temperature drop due to the startup of the inactive pump is experienced, the effect on system pressure and hot channel minimum DNBR is covered by the large power margin to full power conditions.
Therefore, the
~
consequences of this event are bounded by the nominal full power minimum DNBR with four primary coolant pump flow.
ANF 88 108 Page 56 15.4.5 FLOW CONTROLLER MALFUNCTION There are no flow controllers on the PCS at Palisades.
Therefore, this event is not credible, 15.4.6 CVCS MALFUNCTION THAT RESULTS IN A
DECREASE IN THE BORON CONCENTRATION IN THE REACTOR COOLANT 15.4.6.1 Event Descriotion A boron dilution event can occur when primary grade water is added to the primary coolant system via the Chemical Volume and Control System (CVCS) or the accidental transfer of the contents of the iodine removal system duritig cold shutdown or refueling shutdown conditions.
The dilution of primary system boron adds positive reactivity to the core.
I This event can lead to an erosion of shutdown margin for suberitical 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 witMrawal transient (Event 15.4.2).
15.4.6.2 Event Discosition and Justification The boron dilution analysis to support modified RPS operationI3) evaluates the time to criticality caused by the dilution of the primary system baron and the subsequent loss of shutdown margin.
The modified RPS analysis addresses the l
I following modes of operation:
- 1) Refueling; 2) Startup; and, 3) Power f
operation.
The modified RPS boren dilution analysi: also includes a calculation to determine the time to criticality due to the failure to borate i
the core to compensate for reactivity changes after shutdown, t
l Out to changes in the initial and critical boron concentration for Cycle 8, j
the boren dilution event is reanalyzed for refueling, startup and failure to i
{
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I ANF 88 108 Page 57 reborate after shutdown cases.
The consequences for power operation are addressed by the reanalysis of Event 15.4.2 minimum DNBRs for Cycle 8, 15.4.6.3 Results of Analysis (1) Oilution Durina Refuelina For dilution to occur during refueling by primary makeup water, it is necessary to have at least one makeup water transfer pump operating, one charging pump operating, and the makeup controller set for dilution.
None of these conditions are required for refueling and would be in violation of operating procedures.
Nevertheless, such a dilution incident has been analyzed as follows:
1)
One shutdown cooling pump is running to remove decay heat.
2)
The valve in the bleed.off water header from the primary coolant pumps is closed.
3)
The makeup system is set for makeup at shutdown concentration.
4)
The boron concentration of the refueling water to maintain a shutdown margin of at least 5.0%(15) with all rods out of the core.
Periodic sampling insures that the concentration is maintained above the concentration corresponding to 5.0% shutdown margin.
5)
Minimum primary coolant volume for reactor vessel head removal 3
during refueling is considered (3300 ft ).
This is the volume necessary to fill the reactor vessel above the nozzles to insure cooling via the Shutdown Cooling System.
6)
The charging dilution flow is assumed to be 44 gem and the wave front / slug flow approach is utilized.
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The operator has adequate indication of any significant boron dilution from the audible count rate instrumentation.
High count rate is alarmed in the reactor containment and the main control room. The count rate is a measure of the effective multiplication factor.
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With all rods out of the core, the boron concentration must be reduced from the refueling to the critical boron concentratien before the reactor will become critical.
This would take approximately 110 minutes after arrival of the first wave front.
This is ample time for the operator to recognize the audible high count rate signal and isolate the reactor makeup water source by closing valves and/or stopping the primary makeup water transfer pumps.
(2) Oilution Durina Startuo After refueling and prior to hot standby, the primary coolant system may contain water having the boron concentration corresponding to shutdown margin of 2*,
Ap.
The maximum possible rate of introduction of unborated domineralized water is 133 gpm.
The volume of reactor coolant is about 8,628 ft, which is the total volume of the primary coolant system with 29.3?. steam generator tube plugging, excluding the pressurizer.
The primary coolant pumps are assumed to be running (i.e., perfect mixing is assumed).
Under these conditions the minimum time required to reduce the reactor coolant boron concentration to the critical concentration is about 44 minutes.
Boron dilution for start up will be performed under strict procedures and administrative controls.
During dilution at hot standby or reactor critical, the operating staff will be monitoring the nuclear instruments and the boronometer readings.
An abnormal change in the reading of these instruments will inform the operator that dilution is occurririg.
The operator will have further indication of the process from volume control tank level and from operation of the letdown
9, 4
ANF 88-108 Page 59 diverter valve.
Further, should the makeup controller fail to close the makeup stop valve, the operator has visual indication of makeup water flow and of makeup water transfer pump operation.
In any case, should continued dilution occur, the reactivity insertion rate would be less than that considered for uncontrolled rod / rod bank withdrawals.
The reactor protection provided for the rod withdrawai incident will also provide protection for the boren dilution incident.
When the primary system boron concentration is being changed, at least one shutdown cooling pump or one primary coolant pump must be functioning to provide sufficient heat removal capacity.
Under the condition of one operating shutdown ' cooling pump, imperfect mixing is conceivable.
With imperfect mixing, a shutdown cooling pump flow greater than or equal to 2810 gpm is required to ensure that the acceptance criteria for this event is not violated for 27, 40 Alternatively, a minimum shutdown cooling flow of 1500 gpm will not violate the event acceptance criteria for a shutdown margin of at least 3.5?. ao.
These values were calculated by evaluating the minimum shutdown cooling pump flow rate necessary to bring the plant to a critical state in at least 15 minutesIII, assuming a maximum charging flow rate of 133 3
gpm and a reactor coolant volume of about 8628 ft.
(3) Dilution burina power Ooeration inadvertent injection of primary makeup water into the primary coolant system while the reactor is at power would result in a reactivity addition initially causing a slow rise in power, temperature and possibly prersure.
Assuming that unborated water is injected at the maximum possible rate of 133 gpm, the ratt of reactivity addition would be about 6x 10 6 ac/s.
This is much slower than the maximum rate possible with a rod withdrawal.
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.o ANF-88 108 Page 60 Continued baron dilution after reactor trip, if the operator takes no corrective action, is addressed in Reference 3.
The assumptions used in the
[
Reference 3 analysis bound Cycle 8 operation.
(4)
Failure to Add Boron To Comoensate for Reactivity Chances After Shutdown The analysis of the boron dilution event for this case is presented in Reference 3.
The assumptions employed in the Reference 3 analysis remain valid for Cycle 8 operation.
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15.4.6.4 conclusion The results of the analysis for this event are sumarized in Table 15.4.61.
The results show that there is adequate time for the operator to manually terminate the source of dilution flow.
The ocerator can then initiate reboration to recover the shutdown margin.
Boron dilution during power operation is bounded by the analyses presented in Sections 15.4.1 and 15.4.2.
However, the results presented here demonstrate that there is adequate time i
for the operator to manually terminate the source of dilution flow following reactor trip.
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ANF 88 108 i
Page 61 Table 15.4.6 1 Summary of Results for the Boron Dilution Event Reactor Conditions Dilution By Time to Criticality Refueling Primary Water 110 minutes (Charging at 44 gpm)
Refueling and Startup with Primary Coolant System Filled Primary Vater 44 minutes (Charging at 133 gpm main reactor coolant pumps running)
Refueling and Startup Primary Water
>l5 minutes
- with Primary Coolant System Filled Hot Standby or Critical Primary Water Considered in the uncontrolled rod /
rod bank withdrawal analysis Following a trip from the Power Operation Condition Bounded by Ref. 3 Failure to add boron to compensate for Reactivity changes after Shutdown Bounded by Ref. 3 Charging flow is 133 gpm and RHR flow 12810 gem with 12f. 10 shutdown margin at RHR flow 21500 gem with 23.5*. Ao shutcown margin.
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15.4.7 INADVERTENT LOADING AND OPERATION OF A FUEL ASSEMBLY IN AN 3 PROPER POSITION 15.4.7.1 Event Descriotion l
An inadvertent loading of a fuel assembly in an improper position can result in an alteration of the power distribution in the core which can adversely affect thermal margin.
15.4.7.2 Event Olsoosition and Justification is disposed as bounded for modified RPS operation due to the The event administrative controls and proceduros that ensure a properly loaded core (2)
The changes for Cycle 8 will not invaildate this disposition; consequently, this event will not require analysis.
15.4.8 SPECTRUM OF CONTROL _ ROD EJECTION ACC10EN 1 15.4.8.1 Event Descriotion This event is initiated by a failure in the CROM pressure housing causing a rapid ejection of the affected control rod.
h ejection of the control red Because of the inserts positive reactivity causing an increase in core power.
increase in core power, this event challenges both DNBR and overpressurization i
acceptance criteria.
15.4.8.2 Event Diseosition and Justification r
The minimum DNBR and pressurization consequences of a control red ejection f
W event were analyzed for the analysis supporting modified RPS operation The HFP case was determined to be most challenging to the acceptance criteria.
For Cycle 8, the system response to an ejected control rod will not change l
i
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O o-ANF 88 108 Page 63 from that for the modified RPS analysis.
Therefore, the pressuritation results for the modified RPS analysis art applicable to Cycle 8.
The fuel failure evaluation must be reanalyzed for Cycle 8 using cycle specific post ejection radial peaking factors.
15.4.8.3 Analysis and Results The minimum DNB case is initiated by the rapid insertion of positive reactivity due to the ejection of a control red.
A minimum ONBR less than 1.17 is calculated to occur for this event.
With the core boundary conditions predicted at the time of minimum ONBR, along with an asymmetric core power distribution, the amount of fuel failure is calculated.
In Reference 3, it was determined that 12.2% of the fuel rods in the core will fail due to the penetration of CNB.
Due to conservative assumptions employed in the Reference 3 analysis, the amount of fuel that is predicted to fall for Cycle 8 is less than 12.2%.
The offsite radiological doses for this event were calculated in Reference 3 to be below the 10 CFR 100 dose limits for 12.2% fuel failure.
1 15.4.8.4 Cenelusion The radiological doses are conservatively calculated to be less than the 10 l
CFR 100 dose limits.
Applicable acceptance criteria are considered, therefore, to be met for Cycle 8.
15.4.g SPECTRUM OF ROD OROP ACCIDENTS fBWR1 This event is not applicable to Palisades since it is not a BWR, 15.5 INCREASES IN REACTOR COOLANT SYSTEM INVENTORY
'0 l
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t ANF 88 108
[
Page 64 l
15.5.1
' NADVRRTENT OPERATION OF THE ECCS THAT INCREASES REACTOR COOLANT i
.NVEN"0RY 15.5.1.1 Event Descrintion i
This event is caused by an inadvertent actuation of the ECCS that results in l
an increase in the primary system inventory.
The primary challenge is to the f
primary system overpressurization criteria.
For the case where the primary system boron concentration is reduced as a result of ECCS actuation, Event 15.4.6 is bounding.
l 15.5.1.2 Event Disnosition and Justification This event was disposed to be bounded by Events.15.4.6 and 15.2.1 for the
[
l The event initiators and I
analysis supporting modified RPS operation (2) significant parameters remain unchanged for Cycle 8 operation as compared to II'3)
Therefore, the event is not analyzed fer l
the modified RP3 analysis Cycle 8.
I i
15.5.2 CVCS MALFUNCTION THAT INCREASES REACTOR COOLANT INVENTORY I
t 15.5.2.1 Event Descrintion l
A malfunction in the CVCS could result in the insdvertent operation of the i
If the letdown system is not operating, the result charging system pumps.
leads to an increase in the primary system coolant inventory and, potentially, an overpressurization of the primary system and/or a dilution of the primary system boron concentration, l
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ANF.88.!C8 Page 65 l
15.5.2.2 Event Discosition and Justification Sufficient relief capacity exists to limi,t the overpressurization potential to less than the 110% design value of 2750 psia.
The potential for dilution of the primary system boron is addressed in Event 15.4.6.
Reference 2 disposed this event as being bounded by Events 15.4.6 and 15.2.1 l
for modified RPS operation.
The event initiators and significant parameters i
remain unchanged for Cycle 8 operation.
Therefore, the event is not analyzed l
for Cycle 8.
i i
15.6 DECREASES IN REACTOR COOLANT INVENTORY 15.6.1 INADVERTENT OPENING OF A PWR PRESSURIZER PRESSURE REllEF VALVE l
4 I
15.6.1.1 Event Descrietion 1
An inadvertent opening of a pressurizer pressure relief valve or safety valve j
causes a decrease in the primary system pressure resulting in a loss of both thermal rargin and primary coolant inventory.
l The pressurizer reitef valves at Palisades are blocked closed during power operation by downstream isolation valves.
Therefore, an inadvertent opening For l
of a reitef valve will not result in a loss of primary coolant inventory.
1 a stuck open safety valve after a transient, the loss of coolant accident j
i (LOCA) mitigating procedures will begin.
15.6.1.2 Event Diseosition and Justification For a Reference 2 disposed this eve,it as not being credible for Modes 1-5.
stuck open safety valva after a transient, the event is bounded by the small break LOCA (Event 15.6.5). Changes for Cycle 8 operation will not change this l
disposition, l
l
,1 ANF SS 108 Page 66 15.6.2 RADIOLOGICAL CONSEQUENCES OF THE FAILURE OF SMALL LINES CARRYING PRIMARY COOLANT OUTSIDE OF CONTAINMENT 15.6.2.1 Event Descriotion This event occurs when a small line carrying primary coolant outside of containment ruptures leading to a depletion of primary system coolant and a release of contaminated liquid.
The charging and HPSI systems provide sufficient coolant to replenish that which is lost.
Consequently, no fuel f ailures would be predicted assuming a reactor trip on low pressurizer
- pressure, TM/LP or Safety injection Signal (SIS).
The radiological consequences are limited by the maximum primary coolant activity level allowed by the Technical Specifications.
15.6.2.2 Event Otsoosition and Justifica1jst Reference 2 disposed this event as being bounded by the small break LOCA (Event 15.6.5).
Changes for Cycle 8 operation will not change this disposition.
15.6.3 RADIOLOGICAL CONSE00ENCES OF STEAM GENERATOR TUBE FAILURE 15.6.3.1 Event Descriotion This incident occurs when a steam generator tube fails causing a leakage of coolant from the primary system to the secondary system.
The leakage results in a depletion of primary coolant, a reduction of primary system pressure and a release of fission products to the main steam system.
The consequences of this event are maximized for a rated power initial condition due to the amount of stored energy and decay heat that must be removed prior to bringing the two systems to an equilibrium pressure state.
t AfiF.88 108 Page 67 15.6.3.2 Event Diseosition and Justificatign The FSAR analysis was performed at a reactor power level of 2650 MWt and a primary system pressure of 2100 psiaIOI.
For a complete severance of one steam generator tube with a suusequent leakage rate greater than the capacity of the charging pumps, the reactor would trio en a low pressurizer (TM/LP) pressure signal of 1750 psia.
The TM/LP trip acts to protect against fuel damage in this event.
The dose calculations in the FSAR significant analysis were perforr.ed with a source term based on 17. fuel rod f ailure(8)
For Cycle 8, the core power is 2530 MWt with a 3.5Y. increase in radial peaking limits relative to previous cycles.
The Cycle 8 core power is about 4.57. less than the FSAR analysis while the radial peaking f actor is 3.5*.
For the same assembly exposure and 17. fuel red f ailure, the primary higher.
coolant activity for the FSAR analysis is about 1". higher than would be the case for Cycle 8.
Therefore, the amount of radioactive fission products that leak from the primary to the secondary system is greater for the FSAR assumptions.
After the reactor has tripped, the decay heat and stored energy in the core is For the modified removed via the atmospheric dump valves and steam bypass.
RPS analysis and Cycle 8 operation, the reactor power is 2530 MWt and the pressurizor pressure is 2060 psia, as compared to 2650 MWt and 2100 psia for the FSAR analysis.
The time required to remove the primary system energy for Therefore, for a power level of 2530 MWt is less than that for 2650 MWt.
Cycle 8 coeration, the secondary system steam valves are open for a shorter period of time resulting in a smaller radioactive release to the atmosphere.
N Referenca 2 disposed this event as being bounded by the FSAR analysis This disposition will not change for Cycle 8.
,t ANF 88 103 Page 68 15.6.4 RADIOLOGICAL CONSEOUENCES OF A MAIN STEAM LINE FAILURE OUTSIDE CONTAINMENT (BWR1 This event pertains to BWRs and is, therefore, not applicable to Palisades.
15.6.5 LOSS OF COOLANT ACCIDENTS RESULTING FROM A SPECTRUM OF POSTULATED PIPING BREAKS WITHIN THE REACTOR COOLANT PRESSURE BOUNDARY 15.6.5.1 Event Descriotion This event is initiated by a breach in the primary system pressure boundary.
The event initiators vary from relatively small breaks for small break LOCAs (SBLOCA) to complete ruptures of the PCS piping for large break 1.0CAs (LBLOCA).
The primary concerns of LBLOCA and SBLOCA analyses are the peak clad temperature (PCT) and, the amount of localized and core wide metal water reaction.
15.6.5.2 Event Disoosition and Justification ANF has performed a t.0LOCA analysis for Palisades which supports operation with the radial peaking limits given in Reference 15.
The results of this analysis are orovided in Reference 8.
According to Reference 8, the LBLOCA results are more limiting than the SBLOCA results.
For Cycle 8, the LBLOCA is disposed to be analyzed to show that the increased radial peaking does not result in a violation of 10 CFR 50.46(b) acceptance criteria.
For Cycle 8, the radial peaking factors will increase by 3.5*.
The l
changes to the Cycle 8 core will not cause the SBLOCA to become more limiting l
than the 1.BLOCA.
Therefore, a LBLOCA analysis for Cycle 8 operation with increased radial peaking limits will bound the consequences of a SBLOCA.
t k
ANF.38 108 Page 69 15.6.5.3 Analysis and Results l
i The analysis and results of the LBLOCA performed for Palisades Cycle 8 are l
documented in Reference 10.
15.7 RADI0 ACTIVE RELEASES FROM A SUBSYSTEM OR COMPONENT 2
15.7.1 WASTE GAS SYSTEM FAILURE 15.7.2 RA 31mTIVE L10010 WASTE SYSTEM LEAK OR FAILURE (RELEASE TO I
AT90$PHERE) j 15.7.3 POSTULATED RADICACTIVE RELEASES DUE TO LIOUID-CONTAINING TANK FA : LURES The results of the three events above are not dependent on either fuel type, steam generator tube plugging, reactor coolant flow rate, reactor coolant inlet temperature, or reactor protection system modifications.
The reference analysis is therefore not affected by the current licensing action and remains f
the bounding analysis for this event.
The reference analysis is provided in the Updated Palisades FSAR, Reference 8.
15.7.4 RADIOLOGICAL CONSEQUENCES OF FUEL HANDLING ACCIDENT
}
p 15.7.4.1 Event Descrietion A fuel handling accident occurs when a fuel assembly is damaged during j
refueling operations such that fuel rods are ruptured resulting in a release
{
j cf radioactivity.
The inventory of radioactive fission products is determined by the exposure and power level of the assemblies or fuel rods, j
l
6 ANF 08 108 Page 70 15.7.4.2 Event Diseosition and Justification The FSAR analysis assumes that the affected assembly is resident in the core for three full power years with a power of 2650 MWt and a peak rod radial peaking factor of 1.65(8)
The effective power level of the peak assembly is about 21.4 MWt.
The fission product inventory for the assembly is conservatively calculated based on the fission products contained in the peak powered fuel rod.
1 For Cycle 8 operation, the core power is 2530 MWt and r.ie peak red radial y
peaking factor is increased 3. 5 *..
For this peaking, the effective peak e
assembly power is about 21.4 MWt.
The effective assembly powers for both the reference analysisI8) and Cycle 8 are essentially the sare.
For the given f
assembly exposure, the amount of fission products will be the same
^>r the Cycle 8 conditions as compared to the FSAR conditions.
Therefore, the
~
consequences of a fuel handling accident for Cycle 8 are addressed of the FSAR analysis (0) 15.7.5 SPENT FUEL CASK OROP ACCIDENTS 15.7.5.1 Errat Descrietien A spent fuel cask drop accidant can result in the damage of an irradiated fuel assembly and the subsequent release of radioactivity.
15.7.5.2 Event Diseosition and Justification Reference 8 contains an analysis of the radiological consequences of this event.
The FSAR analysis conservatively assumes that the asse-bly with the maximu'n exposure is damaged.
A radial peaking f actor of 2.0 is appliec to this assembly.
The disposition of this event for the analysis supporting the modified RPS
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operation states that the FSAR analysis is bounding.
The peaking factor used I
' in the FSAR analysis tounds that for Cycle 8.
Therefore, the FSAR analyiis y
bounds the consequences for Cycle 8 operation.
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ANF-88-108 Page 72 4.0 IHERMAL-HYORAUllC COMPATIBILITY Thir section describes th thermal-hydraulic analyses performed in support of the following for Palisades Cycle 8:
(1). Insertion of four ANF lead assemblies with high thermal performance (HTP) spacers. The HTP spacer lead assemblies t.re each composed of 216 fuel rods.
(2). For Cycle 8,16 assemblies will be inserted along the core periphery to reduce neutron fluence on critical vissel welds.
The outer four rows of rods (56 rod locations) along one side of each of these shielding assemblies will be replaced with stainless steel :ods.
The purpose of the analyses is to demonstrate hydraulic compatibility of the these assemblies with the existing Palisades core.
Discussed in this Section are analyses of the affect of the ANF lead assemblies and stainless steel assemblies on the minimum departure from nucleat6 boiling ratio (DNBR) for the Palisades core.
The lead assemblies and reconstituted stainless steel assemblies will have no adverse teoact on LOCA/ECCS performance.
4.1 Thermal Hydraulic Desian Criteria The primary thermal hydraulic design criteria for ANF reload fuel assure that fuel rod integrity is maintained during normal operation and Anticipated Operational Occurrences (A00s).
Specific criteria are:
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ANF-88 108 Page 73 (1). Avoidance of DNB for the limiting rod in the core with 95f.
probability at a 957, confidence level.
(2). Fuel centerline-temperatures remain below the melting I
point of the fuel pellets.
Observance of these criteria is considered conservative relative to the requirement that A00s not result in fuel rod failures or loss of functional capability.
4.2 Summary of Results Results of minimum ONBR calculations performed to support ANF HTP, spacer lead assemblies in the Palisades reactor for Cycle 8 show that the XNB 95/95 limit of 1.17 is not violated for a limiting A00 event.
Likewise, for a limiting assembly adjacent to a stainless steel shielding assembly, the minimum DNBR is well above the XNB 95/95 limit of 1.17.
The minimum DNBR performance of the core during A00s thus accords with the thermal hydraulic design criterion on i
ONBR.
The thermal hydraulic simulations employed to evaluate minimum DNBR were performed in accordance with ANF's NRC approved thermal hydraulics methodology for mixed cores (16)
The 2?. mixed core penalty of minimum DNBR has not been assessed in these calculations because the lead assemblies do not represent a significant fraction.of the core.
For standard ANF fuel assemblies, fuel centerline temperatures have been shown in the Chapter 15 evt.nt analysis of A00s to be less than the limit for
{
incipient melt of 21 kW/ft.
The centerline temperatures for the lead assemblies and shielding assemblies will also be less than this limit.
b These results adequately demonstrate the thermal hydraulic compatibility of j
the HTP spacer lead assemblies and stainless steel shielding assemblies with j
4
g ANF-88-108 Page 74 the co-resident ANF standard fuel at Palisades.
Thermal-hydraulic design criteria are met for these fuel types.
4.3 Analysis and Results The thermal-hydraulic analysis for the lead assemblies with HTP spacers and the stainless steel assemblies will be discussed in the following two sections.
4.3.1 Lead Assemblies with HTP Soacers The spacer loss coefficients for the ANF standard fuel are derived from pressure drop tests performed in ANF's portable loop hydraulic test facility (I9)
The HTP spacer loss coefficient is also based on pressure drop test data from ANF's portable loop hydraulic test facility.
The ANF standard assembly has ten bi metallic spacers.
The ANF HTP spacer assembly modelled has ten HTP spacers.
The loss coefficients for the other assembly components (i.e., upper and lower tie plates) are identical for both the lead and standard fuel designs.
The overall assembly loss coefficient for an ANF lead assembly exceeds that of the ANF standard fuel by about 10?..
A full core of ANF fuel with HTP spacers would slightly decrease the total vessel flow relative to the current Palisades core, due to the greater hydraulic resistance of the HTP spacers.
The core flow distribution (CFD) analysis is performed to assess crossflow between assemblies in the core for use in subsequent minimum DNBR subchannti analyses.
The core flow distribution analysis is particularly important for mixed fuel loadings where hydraulically different fuel types are co-resident The result of the CFO analysis is a set of axially varying in the core.
boundary conditions on heat, mass, and momentum fluxes through the vertical boundaries of the assen.blies of interest.
These boundary conditions are employed in the subsequent 1/8th assembly simulations in which minimum DNBR is i
<l i
i A.NF-88 108 Page 75 computed.
In the analysis each fuel assembly in an octant of the Palisades core is modeled as a hydraulic channel.
The calculations are performed with the XCOBRA-IIIC computer code (6)
Crossflow between adjacer.t assemblies in the open lattice core is directly modeled.
The single-phase loss coefficients are used in the analyses to hydraulically charactorize the assemblies in a mixed core.
The core flow and subchannel calculations are performed at conditions The lowest DER representative of the dropped rod A00 for Palisades Cycle 8.
for a dropped rod event is calculated at full power with a nominal pressure of LCO.
For the 2200 psia and a flow of 130 M1bm/hr, as allowed by the Tinlet standard fuel assembly design the minimum DNBR under these conditions is calculated to be 1.22.
The radial peaking factor for the lead assembly was set equal to the proposed increased Technical Spe.:ification limit of 1.73 for a 21C rod assembly.
The limiting standard foal design is a 208 rod assembly.
A Sr. inlet flow maldistribution is assumed for the limiting assembly and surrounding assemblies.
The axial power distribution employed in the calculations is the limiting full power axial with an ASI of 0.139.
To establish the limiting assembly boundary conditions for the subsequent minimum DNBR
- analyses, two separate calculations were made.
These calculations provide heat, mass and momentum flux boundary conditions as a function of axial position for the following cases:
(1)
Limiting ANF HTP spacer lead assembly loaded in an interior location.
(2) Limiting ANF HTP spacer lead assembly loaded on the core periphery.
c
i ANF-88-108 Page 76
~ Boundary conditions from these cases were passed to the 1/8 assembly analysis for the minimum DNBR calculations.
In the 1/8 assembly simulation, the XCOBRA IIIC computer code is employed to evaluate the pertinent thermal hydraulic variables in the inter rod flow channels of the fuel assembly of interest.
Heat, mass, and momentum fluxes between the inter rod flow channels are explicitly calculated.
Local values of mass velocity and enthalpy are determined, and us1d to calculate the ONBR via the XN8 critical heat flux correlation (17,18)
Axially varying boundary conditions on the vertical boundaries of the assembly are obtained from the
)
appropriate CFD calculation, discussed above.
The calculations include factors to account for manufacturing tolerances and l
densificatior effects.
Specifically, a 3% engineering factor is applied to the limiting red power to account for fabrication tolerances on pellet diameter, density, enrichment and cladding diameter.
These manufacturing tolerances potentially affect heat flux at the limiting DNBR location in the t
assembly.
I f
The XNB DNB correlation is demonstrated to be applicable to the ANF standard l
fuel assemblies in Reference 18.
The ANF HTP spacer is specifically designed Flow l
to yield improved DNB performance relative to the ANF standard spacer.
l mixing data for the similar 17x17 HTP spacer design demonstrate significantly f
improved mixing relative to the ANF standard
- spacer, supporting the 1xpectation of improved DNB performa,ce.
The XNB correlation may be conservatively applied to the ANF HTP spacer lead assemblies in this analysis.
j l
For Case 1, a minimum DNBR of 1.18 is conservatively calculated for the ANF j
For Case 2, a minimum 09RR of 1.28 is calculated.
Because HTP lead assembly.
)
of the higher spacer loss coefficient for the h'a le.f assembly, flow is diverted from these assemblies to surrounding assemblies with standard l
spacers.
Consequently, local mass velocity decreases and local enthalpy
. -.. ~.,.. - -. - -.., -. -. -.. - - -. - - - - - -
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ANF 88-108 Page 77 increases yielding a lower DNBR (about 3?.) relative to a standard ANF design.
DNBR benefit due to increased mixing in the HTP spacer assemblies has been conservatively neglected for this analysis.
With the lead assembly loaded on the core briphery, Case 2, less flow is diverted to adjacent assemblies due to the proximity of the core baffle pl ate.
Because less flow is diverted from an assembly loaded on the core periphery, as compared to the flow diversion of an interior assembly, the minimum DNBR conditions are less severe.
Therefore, the minimum DNBR for Case 2 is about 8% higher than that for Case I with the lead assemblies loaded in interior locations.
The results of this analysis show that the calculated minimum DNBRs for HTP spacer lead assemblies in the Palisades reactor meet the 95/95 DNBR limit for the limiting A00 transient event for Cycle 8.
Therefore, safety margin is not compromised for the Palisades Cycle 8 core with four HTP spacer lead assemblies.
4.3.2 Stainless Steel Shieldina Assemblies The shielding assemblies will be loaded along the core periphery to reduce the neutron fluence on critical vessel welds.
Because the shielding assemblies are previously t,urnt assemblies reconstituted with stainless steel rods, the assembly power level will be substantially lower than the surrounding conventional fuel ast,enblies.
Higher powered assemblies adjacent to the shieldit,g assemblies may potentially experience an increase in crossflow due to the thermal differences between the two fuel types.
This increase in crossflos could adversely impact minimum DNBR in the affected assemblies.
To assess the impact to minimum DNBR for Cycle 8, a thermal hydraulic analysis was performed.
The details of the analysis are similar to those discussed above for the HTP spacer lead assemblies.
S a
t ANF 88 108 Page 78 The core flow and subchannel calculations were performed using XCOBRA-IIIC.
The core flow model consists of an octant of the Palisades Cycle 8 core with each assembly modelled as a hydraulic channel.
The hydraulic characteristics of the shielding assemblies are similar to those for the standard fuel design.
The assembly design parameters for the stainless steel assemblies are given in Table 4 1.
The core conditions used in this analysis are the same as those used in the HTP spacer calculations.
The radial peaking factor of an assembly adjacent to a
stainless steel shielding assembly was increased to the Technical Specification limit for that fuel type.
Axially varying crossflow boundary conditions for the limiting assembly are generated by the core flow calculation.
Using the crossflow boundary conditions from the core flow calculation in the 1/8 assembly subchannel model, the thermal hydraulic conditions in the th limiting subchannel are evaluated.
These conditions in conjunction with the XNB critical heat flux correlation yields a minimum DNBR.
The minimum DNBR for an assembly located adjacent to a shielding assembly is 1.33 which is well above the XNB 95/95 correlation limit of 1.17.
The minimum DNBR for a standard fuel assembly under these conditions is 1.22.
This result indicates that the presence of stainless steel shielding assemblies will not impact thermal margin for Cycle 8.
Because of the relatively low assembly power level, the stainless steel shielding assemblies will not penetrate minimum DNBR limits.
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i ANF-88-108 Page 79
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Table 4-1 Fuel Design Parameters'for the Stainless Steel Shielding Assemblies Fuel Parameters 0.417 inches Fuel Rod 00 0.437 inches Stainless Steel Rod 00 1.115 inches Guide Tube 00 15x15 Rod Array 0.55 inches Rod Pitch Number of Fuel Rod Positions /
152 Assembly Number of Stainless Steel Rod Positions / Assembly 56 S
Number of Guide Bars 8
Number of Guide Tubes NumberofinstrumentTubes 1
t J
5, ll ANF-88-108 Page 80
5.0 REFERENCES
1.
"Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," NUREG-0800, LWR Edition, U.S. Nuclear Regulatory Commission, Offico of Nuclear Reactor Regulation, July 1981.
2.
"Palisades Modified Reactor 9rotection System Report Disposition of Standard Review Plan Chapter 15 Events," ANF-87-150(NP). Vol. 1, Advanced Nuclear Fuels Company, June 1988.
3.
ANF-87-150fNP). Volume 2, "Palisades Modified Reactor Protection System Report: Analysis of Chapter 15 Events", Advanced Nuclear Fuels Company, June 1988.
4.
E. Daniel Hughes, "A Correlation of Rod Bundle Critical Heat Flux for Water in the Pressure Range 150 to 725 psia", IN-1412 (TID-4.1Q9.1, Idaho Nuclear Corporation, July 1970.
5.
"Description of the Exxon Nuclear Plant Transient Simulation Model for Pressurized Water Reactors (PTS-PWR), XN NF-74-5(A). Rev. 2, Exxon Nuclear Company, October 1986, and Supplements 3-6.
6.
"XCOBRA IllC:
A Computer Code to Determine the Districution of Coolant During Steady State and Transient Core Operation", XN NF 75-21(A). Revision 2, Exxori Nuclear Company, January 1986.
XN NF-86-91(P), "Low Flow Trip Setpoint and Thermal Margin Analysis 7.
for Three Primary Coolant Pump Operation of the Palisades Reactor",
Exxon Nuclear Company, November 1986.
8.
Palisades Final Safety Analysis Report, Updated Version (as of July 1985), Consumers Power Company.
9.
"Advanced Nuclear Fuels Methodology for Pressurized Water Reactors:
Analysis of Chapter 15 Events," ANF-84 73(P). Rev. 3, Advanced Nuclear fuels Company, May 1988.
"Palisades LOCA ECCS Analysis for 2530 MWt Operation with Increased 10.
Radial Peaking and 29.3% Steam Generator Tube Plugging", ANF 88 107, Advanced Nuclear Fuels Company, August 1988.
11.
"Plant Transient Analysis of the Palisades Reactor for Operation at 2530 MWt", XN NF-7718, Exxon Nuclear Company, July 1977, "Systematic Evaluation Frogram Design Basis Transient Reanalysis for 12.
the Palisades Reactor", XN NF 81 25, Revision 1, Exxon Nuclear Company, May 1981.
/l i
ANF-88-108 Page 81 13.
"Palisades Cycle 5 Reload Fuel Safety Analysis Report," XN-NF 81-34(P). S000. 1, Exxon Nuclear Company, December 1981, 14.
"Palisades Primary Design Parameters for PTS Analysis, XN-NF 19.1.(21, Exxon Nuclear Company October 1985.
15.
Palisades Plant Technical Specifications, Consumers Power Comrany, Appendix A to Lic)nse No. OPR-20, Revision dated October 28, 1987.
16.
"Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," XN-NF-82 21(A). Rev. 1, Exxon Nuclear Company, September 1983.
17.
"E;:xon Nuclear DNB Correlation for PWR Fuel Design," XN-NF-621( A).
Revision 1, Exxon Nuclear Company, April 1982.
18.
"Justification of XNB Correlation for Palisades", XN-NF-709, Exxon Nuclear Company, May 1983.
"Single Phase Hydraulic Performance of Exxon Nuclear and Combustion 19.
Engineering Palisades Fuel Assemblies", XN 76-1. Revision 0, Exxon Nuclear Company, January 1976.
20.
Letter, 0.J. VandeWalle (CPCo) to Director, Nuclear Reactor 20, 1985, Docket 50-255, License OPR 20, Regulation, dated June Palisades Plant, "NUREG 0737 Item II.K.3.30 Small Break LOCA Models and Item II.K.3.31 Plant Specific Analysis".
Letter, John A. Zwolinski (NRC) to Mr. VandeWalle (CPCo), dated 21.
July 3, 1985, Docket 50-255, License DPR-20, Palisades Plant, "TMI Action Plan Item !!.K 3.30, Small Break LOCA Analysis and II.K.3.31, Plant Specific Analysis".
Ls l ;.
>,e ANF 88 108 Issue Date: 8/2/88 PALISADES CYCLE 8: DISPOSITION AND ANALYSIS OF STANDARD REVIEW PLAN CHAPTER 15 EVENTS Distribution RA Copeland RC Gottula JS Holm l
JW Hulsman JD Kahn J
1 TR Lindquist LA Nielsen i
LD O' Dell 1
FB Skogen HE Williamson CPCo/HG Shaw (20) l, I
Document Control (3) l
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ERRATA SHEET FOR ANF-88-108 DATED AUGUST 1988 1.
Section 1.0, Page 1, a fourth change is being implemented for Cycle 8 in 1
that 16 assemblies having 6 w/o Gd 0 and 12 assemblies having 4 w/o 2
Gd o are being introduced instead o the norma 1 20 assemblies with 4 w/o g3 Gd 0 '
23 2.
Table 2-1, 15.1.4, "Relief of Safety Valve" should be "Relief or Safety Valve" 3.
Table 2-1, 15.2.7, under disposition, should say "Short term bounded" by "15.3.1" and "long term bounded" by "Ref 3".
4.
Table 2-1, line 15.4.3(5), "Ref 8" should be deleted under Bounding Event or reference column.
5.
Page 23, definitions should be added for the variables PVar, QA, QR, and g
Q.
6.
Section 15.1.3.3, second line, "flow increase of about 112%" should be "flow increase to about'112%".
7.
Page 28, second paragraph, first sentence, should begin "For the Hot Zero Power Case" instead of "For the hot shutdown case".
8.
On page 37, first paragraph, "reference 2" should be "reference 3".
9.
Page 54, The last sentence starting with "additionally, startup of..."
should be deleted.
- 10. On page 58, the first paragraph should be revised to indicate that "during fuel moves the operator has adequate indication of any signif. cant boron dilution from the audible count rate instrumentation.
High count rate is indicated by an increased' Tick frequency in the reactor containment and the main control room."
11.
On page 58, last paragraph, first sentence, delete the last four words "and the boronometer readings."
12.
Page 59, third paragraph should be modified to indicate that event acceptance criteria are also met with the assumption of a 3.5% shutdown margin, shutdown cooling flow of at least 650 gpm and no more than one charging pump operating.
13.
Page 65, section 15.6.1.1, in the first sentence of the second paragraph the word "downstream" should be deleted.
14 Page 75, second full paragraph on the second and third lines the words "dropped rod" should be "rod withdrawal".
- 15. Table 4-1, Guide tube 0.D. should be 0.417 inches.
MIO888-0055B-OP03-NLO2