ML20236B835
| ML20236B835 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 07/21/1987 |
| From: | Adkisson D ADVANCED MEDICAL SYSTEMS, INC. |
| To: | |
| Shared Package | |
| ML20236B791 | List: |
| References | |
| ANF-87-95, ANF-87-95-R01, ANF-87-95-R1, TAC-66524, TAC-66525, NUDOCS 8710260376 | |
| Download: ML20236B835 (21) | |
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ENCLOSURE 4 ADVANCEDNUCLEARFUELS CORPORATION ANF-87-95 Revision 1 Issue Oate:' 7/21/87-l HATCH 9X9 LEAD FUEL ASSEMBLIES SAFETY ANALYSIS REPORT
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. Prepared By:
/i.'ue /,/ L h
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'O. A. Adkisson 1
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Reload Licensing Licensing and Safety Engineering Fuel Engineering and Technical Services i
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= muure nun.suv o Gawu
a CUSTOMER OtSCLAIMER W80HTANT NOTICE IWilAfWe8G C00ffENTS ANO USE OF THIS DOCURENT MEASE READ CAMEPUU.Y Advensed Nuoteer Fuste Cememnon's wonenese and represenmoens con.
eunung the suceset meest of see document are these est kom a the Agmement besseen Advenged Nussear Puses Cameresen and me Cusemer pumuent a useen mee assument is issued. Aenendngly, essent as senenues empreamy pn>
wided h euen Agreement, neener Advensed Nuoteer Fusie Comomeen ner any semen asens on a moned menos any warranty or rupessenmeen, eeressed or lagded, una resseet e Ins assumsy, commessenses, or useheness of me weer. -
mean seresened M ette deswnent, or me the use of any intermemen, angenaus, nesmed or legemes emelemed in Wils desument we not ininnge pnvessly owned denne: er ensumes any seendes een ressent a em use et any woormemen, ap.
semes, memed or specese demseed h ses denument.
The intermeen eareemed hereen to ter me sale use of Cussemer.
h ereer a meld impelmeens of rtgle of Advanced Nuelser Puste Corgemeen in palmas er meanelene unten may be Ipseluded h sie indennemen contamed in ins desument. Wie resument, by its assessense of Wee desument, eyees not a peamen er meme puses use en me count use er me termt of euen meermanen unst es sumertsed in ureme by Aduenced Nussear Puole Corporamen er unet aAur est M menen saamming temunemen er egnamen of sne esamees Aqpeement and any eenneen menset, weses amenmee emmesy ansvides a me Ayoomem. No stWils or toenese M or e any pannes are ingded Irr Wie furmeNng of Wide doou-mom.
XN NF P00 796 (1/87)
V
O o ANF-87-95; Revision 1 TABLE OF CONTENTS Section EAER -
1.0 INTR 000CTION.......................................................
1 2.0 FUEL MECHANICAL DESIGN ANALYSIS....................................
L3
- 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS..................................
4' 3.1 Hydraulic Compatibility............................................
4 3.2 Thermal Margin Performance.........................................
4 3.3 Single Loop Operation..............................................
4 4.0 NUCLEAR DESIGN ANALYSIS............................................-
5 4.1 Standby Liquid Control System..................................,...
5 4.2 Cold Shutdown Margin...............................................
5-4.3 Fuel Pool Criticality..............................................
5' 5.0 ANTICIPATED OPERATIONAL OCCURRENCES................................
'6 5.1 Overpower Events...................................................
6 5.2 Control Rod Wi thdrawal Error.......................................
6 5.3 Fuel Dislocation Error..............................................
6 5.4 Fuel Rotation Error................................................
7 6.0 POSTU LAT ED AC C ID ENT S...............................................8 6.1 Loss-Of-Coolant Accident...........................................
8 6.2 Control Rod Drop Accident.......................
9 7.0 TECHNICAL SPECIFICATIONS...........................................-
10 7.1 Limiting Safety System Settings....................................
10 7.2 Limiting Conditions for Operation..................................
10 7.3 S uTve i l l ance Requi reme n t s..........................................
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8.0 REFERENCES
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'ANF 87-95 Revision 1 1
i LIST OF TABLES O
?aSt 1
1 Comparison'Of MCPR Limits............................................
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_2 Comparison Of APLHGR Limits..........................................-
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LIST OF FIGURES-1 Fiaure PAgg 1
APLHGR Limi t Comparison On _ Total Planar Power Basi s.................. -15 2
ANF 9x9 LHGR Compared To Lim1t.......................................
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1.0 INTRODUCTION
Evaluations have been performed to define the impact upon the core performance =
as. a result of inserting four (4): 9x9 lead fuel assemblies -(LFA's) manufactured by Advanced Nuclear Fuels Corporation-(ANF) into the' Hatch Unit 2 Nuclear Station.
In addition, justifications 'are provided which -demonstrate that application of GE P80RB284H 8x8 operating - limits, as defined in the Technical Specifications to these lead fuel assemblies, is acceptable and will not result in decreasing the reactor's margin.to ' safety during operation w,th the ANF 9x9 assemblies.
The insertion of only four ANF 9x9 assemblies will have negligible, effects upon the core-wide transient performance relative to'. the core fully cloaded '
without the four ANF lead fuel assemblies.
As such, the analyses of the core transient performance used to establish ~ the current Hatch Technical Specification limits for a core loaded without the four ANF 9x9 LFA's applies-directly to the core loaded with the four ANF 9x9 assemblies replacing four.
P80RB284H 8x8 assemblies.
This includes the analyses of anticipated plant transients, LOCA, and stability which are used to support ARTS, extended load l
line, single loop operation, increased core flow, and feedwater temperature reduction.
The maximum k. of an ANF 9x9 LFA is slightly less than a GE P80R8284H.8x8 assembly.
Therefore, existing fuel storage limits for GE fuel bound those necessary for the ANF 9x9 LFA's.
Analyses performed for GE P80RB284H 8x8 fuel to determine the effects of core related events, such as control rod withdrawal, control rod drop, and fuel assembly misloading, also apply to the ANF 9x9 assemblies by-virtue of the 9x9 assemblies meeting compatibility requirements of reactivity and hydraulic demand.
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4 2-ANF-87-95
' Revision.1 lj The' evaluations provided herein thus provide assessments of the 9x9 assemblies relative to the GE P80RB284H 8x8' assemblies and justify application of the current Hatch Technical ~ Specifications ; for that fuel to ' the ANF 9x9 fuel assemblies.
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ANF-87-95
. Revision 1 2.0 FUEL MECHANICAL DESIGN ANALYSIS The expected operating requirements of Hatch Unit 2 are bounded by the assumed power history in ANF's fuel mechanical design analyses (l).
Fuel design issues related. to operational occurrences and accident analysis (fuel centerline melting, clad rupture, LOCA-seismic response) have been evaluated for full' reloads in Susquehanna and found acceptable by'the NRC(2).. These evaluati.ons-also assure that the four ANF.9x9 LFA's will meet operating and safety design requirements of the Hatch 2 nuclear plant.
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3.0 THERMAL HYORAULIC DESIGN ANALYSIS'
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3.1 Hydraulic Compatibility
' Component hydraulic resistances-for the ANF 9x9 - and GE 8x8 fuel assemblies
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L have been determined'in single phase flow tests of full scale assemblies which:
j were identical in : mechanical design. to the. Hatch LFA's ' and GE 8x8 fuels.
l Hydraulic compatibility of the ANF 9x9 and GE 8x8 coresident. fuel. types (3): bas i
been demonstrated.
N 3.2 Thermal Narain performance-Analyses of the limiting BWR/4 transients have shown that' the. bundle power needed to produce transition boiling in the 9x9 fuel is higher than that for l
the GE 8x8 bundle.
Table 1 shows that the 9x9 fuel must be operatedt at.a higher bundle power than the GE 8x8 fuel in order to reach the MCPR operating limit.
Therefore, applying GE 8x8 MCPR operating limits to ANF. 9x9 fuel will keep the 9x9 bundle powers to levels lower than would be needed to rr.ach their 4
actual MCPR limit.
ANF analyses in support of_ extended operating domains for a BWR/4 show the equivalence of 8x8 and 9x9 MCPR limits throughout extended operating domains.
It follows that. monitoring the ANF 9x9-LFA's based on GE 8x8 MCPR limits adequately protects the ANF 9x9 LFAs from boiling transition.
3.3 Sinale Looo Goeration ANF anafysis of a typical BWR/4 with a full ANF 9x9 ' reload in Single ' Loop Operation (SLO) has shown that the most. limiting transient with regard to thermal margin is bounded by the 104%' power /100% flow generator ' load rejection without bypass valve operation.
This -analysis showed that single loop operation is unaffected by the introduction of the ' ANF 9x9 LFA's.
In addition, monitoring these ANF 9x9 assemblies with GE P80RB284H limits for SLO results in a c
- arvative estimate of the margin to critical power for the'9x9 l
fuel in single loop operation.
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l 4.0 NUCLEAR'OESIGN ANALYSIS' 4.1 Standby Liouid' Control System
.The neutronic impact of replacing four of t'he 560 fuel assemblies with' ANF 9x9-LFA's which demonstrate similar reactivity characteristics 'will.be negligible on the standby liquid control s'ystem reactivity: worth.
4.2 Cold Shutdown Marain=
Infinite assembl,y calculations at 0 mwd /MTU show the ANF.9x9 LFA's to have approximately 0.6.mk higher. cold uncontrolled reactivity? relative-- to GE l
P80RB284H 8x8 fuel.
This results in a contro11 cell reactivity less than~ 0.2 mk higher than an all P80RB284H 8x8 loaded control ' cell, which is a negligible.
contribution to cold shutdown margin..
For exposures greater ' than. 2,000 l
mwd /MTU the 9x9 LFA. design has slightly lower. cold ' uncontrolled reactivity.
than for the GE 8x8 reference fuel. This results in a slight increasm iti cold shutdown margin for a control cell with an ANF. 9x9 fuel' assembly.f:( place of.
I an 8x8 assembly at exposures greater than 2,000 mwd /MTV.
Thus, the cold shutdown margin evaluations performed for control cells containing all 8x8 fuel apply to control cells containing the ANF 9x9 LFA without~ significantly reducing the calculated cold shutdown margin.
i 4.3 Egg' %ol Criticality The maximum k. of an ANF 9x9 LFA is approximately 2 mk less than a GE
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P80RB284H 8x8 assembly.
Therefore, spent fuel storage critical limits existing for GE 8x8 fuel bound those required for ANF 9x9 LFA's..
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6-ANF-87-95 Revision 1 5.0 ANTICIPATED OPERATIONAL OCCURRENCES Operation of the four ANF 9x9 LFA's using GE P80R8284H MCPR operating limits is conservative.
Analyses of the limiting 8WR/4 transients have shown that i
the ANF 9x9 bundle power at the Technical Specification MCPR operating limit is higher than for a GE 8x8 bundle.
5.1 Overoower Events The limits in effect for GE P80R8284H fuel will conservatively protect the ANF 9x9 LFA's for overpower events.
In the event of an overpower transient, more than 30 percent' margin exists to ANF 9x9 transient LHGR ' l imits(l).. This compares to GE 8x8 fuel, where approximately 20 percent margin exists for overpower translent LHGR limits.
5.2 Control Rod Withdrawal Error Infinite assembly calculations of the control rod worth for the ANF 9x9 LFA's and GE P80R8284H 8x8 fuel indicate that the worth of the withdrawn rod for the module containing the ANF 9x9 fuel will not exceed the value obtained for a similar module containing all GE 8x8 fuel.
Thus, the A CPR values for the ANF 9x9 fuel design will not be substantially different than those obtained for GE 8x8 fuel and are within the variation that is seen between specific reactor cycles for a reactor which utilizes GE 8x8 fuel.
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5.3 Fuel Dislocation Error l
The consequences of the dislocation of an ANF 9x9 LFA are no more limiting than that associated with the GE 8x8 fuel.
This is substantiated by a comparison of the reactivity values between the two fuel types.
The 9x9' values are comparable and in most cases less than that associated with GE 8x8 fuel, thus the change in local power due to the dislocation of a 9x9 fuel
7' ANF-87 -Revision l'-
assembly is no' greater than that obtained tfy the. dislocation.of a GE 8x8 assembly.
Thus, the dislocation A MCPR for.the ANF ; 9x9 fuel " design..iis not:
significantly different-from those for the GE 8x8 fuel.
5.4 Fuel Rotation Error i
The consequences of the fuel rotation error have been evaluated comparing the ANF 9x9 LFA design to GE P80R8284H ' 8x8 - design.
.The results c indicate Han increase in A CPR for the rotated ~ ANF ~ 9x9 LFA of up to 0.06-relative to.
j rotated 8x8 GE P80RB284H fuel assembly.
Typically the. rotated ;8x8 - fuel
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l assembly has,not been the limiting event for Hatch,- and more:than 0.06. ACPR R
margin has 'xisted to the MCPR operating limit.
If necessary, selection of e
1 non-limiting core locations for the four ANF 9x9 LFA's can be used to preclude any concern relative to thermal limits for the fuel rotation error.
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'6.1 Loss-Of-Coolant Accident I
f' The' ~ appropriate ' bundle power limit derived from. a LOCA! analysis is'.the peak'
'h bundle-p1anar power because heatup is. primarily -a.planaq phenomena, not an axial phenomena.
The bundle is contaihed in a'. channdi and ' the peak. clad
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temperature (PCT)' is primarily governed ' by:- rod-to-rod and " rod-to-channel R
radiation, and-local convection to droplets.
Presently..- the. peak bundle-planar power determined-' from the LOCA' analysis is convertedito a: MaxirnuS
,y Average Planar LWGR limit (MAPLHGR) by.' dividing by the number of hered' rods
- [ h in a bundle; this MAPLHGR limit is used as the LOCAL monitoring. limit.
Alternatively, this peak bundle-planar power' could be directly used.as 'the-LOCA-monitoring limit; in this report.this. alternate' limiti:is.' termed-equivalent planar power.
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(c ANF 9x9 fuel has equivalent or improved f.0C/sECCS performance when compared to-both ANF 8x8 and GE 8x8 fuel for two fundamental reasons.
First,s because of its lower LHGR's for the same planar. power, ANF 9x9 fuel has 'less.ptored energy than 8x8 fuel.
Secondly,, ANF. 9x9 ' fuel has better heat tdrisfer '
!.j ij characteristics because of the g'reater surface area pginnit s
volene.
Of further benefit is that ANF, fuel has a larger,uptrer tie. plate' flow ai es than.
- fuel, resulting in l'ess restrictive countercurrent flow limiting characteristics.
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.d Table 2 provides a comparison on an equivalent. basis of average planar power limits for ANF 9x9 and GE 8x8 fuel for a typical BWR/4.
The table shows that-I the ANF fuel is less ' restrictive than GE fuel.
This rem'ains : the case regardless of. bundle exposure.
As a' result of this comparison,- it is;
.I concluded that the APLHGR limits for the. GE - 8x8 fuel
('P8h82'84H) ' ' will conservatively bound the use of ANF ~ 9x9 - fuel. in Hatch 2 for all bundle -
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These3 11.aits will assure that the criteria specified, in 10 CFR 50.46 will be satisfied for the four A?N 9x9 LFA's.
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BWR/4's can be groupedT into two ma,jor subgroups--those with loop selection j
icgic (i.e., plants that have not. incorporated low pressure coolant injection'
[LPC1] system modification) and those which have' LPCI modification (4).
Since Watch falls.into the latter subgroup and ANF' ha's performed a LOCA analysis for 1
ki WR/4 with LPCI modifications (2), ECCS performance differences can 'be censidued insignificant.
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o 6.2 control Rod Droo Accident 1
The consequences of a control rod'urop accidant have.been determined by ANF to be a function of dropped rod worth,' Doppler reactivity, delayed neutron fraction, and fuel rod local peaking.
A comparison of these parameters between the ANF 9x9 and GE 8x8 fuel-indicates that the deposited enthalpy' for the ANF 9x9 fuel will have a value comparable to that calculated for ths GE 8x8 fuel and maintain sufficient margin to the limit of 280 cal /gm.
I 6.3 Fuel Landlina Accident A comparisui of the radiological consequences of fuel handling accidents with 8x8 and 9x9 fuel for a typical BWR/4 showed less radioactivity relea' sed for the'9x9 fuel.
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j' The four. ANF 9x9 LFA's;will not materially effect th'e safety' limits /of Hatch 2' operation.
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.,ANFanalysisofjatypied1.BWR/4hasshownthattheANF9x9bundlepoweratthe H
t MCPR operating 1. bit ;Us higher than for f t GE 8x8 bundle.
It' follows that application of GD "P80R8284H MCPR liMts to the ANF-9x9 LFA's adequately -
prots?.ts the LFA's from boiling transition.
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Restricting the ANF 9x9,LFA's to the planar; power consistantiwith GE APLHGR-l limits protects ANF 9x9'APLHGR and LHGR limih.
As'disedssed'in the previous section, GE APLHGR limits for P80R8284H fuell type in' Hatch ' Unit-2 'are 'more-
" restrictive than ANF 9x9 APLHGR ' limits.
AN '9x9 APLHGR ' limits Lare-more restrictive or equivalent to (depending (on exposure) ANF 9x9 LHGR limits.
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g Figure 1 provides a comparisor. of ' APLHG,R. limits for ANF 9x9 and GE P80RB284H 8x8 fuel.
In order to provide comparative bases between 8x8 and 9x9 arrays, R
the equivalent planar power is shown as the APLHGR limit times the. number of.
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-fuel rods per assembly.
Figure 4providesacomparisonofANF9x9LHGRlimits and the maximutx LHGR allowed by monitoring to GE APLHGR limits for P80R8284H fuel.
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Stability tests have been performed - on the~ Cor.4onwealth Edison' Company's
'f<0resdenUnit2reactorwithANF9x9LFA'sincore.
The'resultsiof these; tests
.dndicate that the ANF LFA's have no measurable, impact ~on' local' stability, o
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11 ANF 87-95 Revision 1 Additionally, the ' Pennsylvania Power and. Light Company's Susquehanna Unit: 2-reactor was analyzed and tests performed for stability -with a core containing a full 'ANF 9x9 ' reload (approximately. 42 percent of the total core loading)'..
Results of.these analyses and tests indicate.the core-is very stable; a decay ratio of 0.33 was measured at the right' hand boundary of the SIL 380 Detect and Suppress region.
The Hatch Unit' 2 mechanical, core design 'and analyzed power / flow map are.the same as those for Susquehanna.
The. nuclear design of the Hatch LFA's is such' that the thermal hydraulic stability is no worse than the. fuel tested. in Susquehanna;~ - Therefore, the local and core-wide ' stability of the LFA's in l
Hatch 2 meets the requirements of GDC 12.
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l 12 ANF-87-95 Revision 1
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8.0 REFERENCES
1 (1)
" Gen'eric Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel,"
XN-NF-85-67(P)(A), Rev. 1, September 1986.
l (2) Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting-
-)
Amendment 31 to Facility Operating License No. NPF-22, Pennsylvania Power l
and Light Company, Susquehanna Steam Station Unit 2, Docket No. 50-388.
1 i
(3)
" Hatch Lead Assembly Compatibility. Report Mechanical, Thermal and Neutronic Design for ANF 9x9 Fuel Assemblies," XN-NF-87-77(P), Rev. O.
(4) General Electric Licensing Topical
- Report, Generic Reload Fuel Application, NEDO-24011-2.
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TABLE.1 COMPARISON OF MCPR' LIMITS (BASED ON TYPICAL BWR/4).
1 BUNDLE POWER.
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ANF 9x9 6.7 di8x8 6.5
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<4 14 ANF-87-95
. Revision it TABLE:2 COMPARISON ~OF APLHGR-LIMITS (8ASED.ON TYPICAL-BWR/4) l
.i EQUIVALENT PLANAR POWER ~
l PEAK APLHGR
-(APLHGR LIMIT *-N04 0F FUEL-J LIMIT (KW/FT)-
R005) (KW/FT)
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- l GE 8x8 12*2 756 (BWR/4)
ANF 9x9 10 2 806 (BWR/4).
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Issue Date: 7/21/87-1 i
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.j HATCH 9X9 LEAD. FUEL ASSEMBLIES
)
SAFETY ANALYSIS REPORT
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l Distributhn.
D. A. Adkisson.
R. E. Collingham l
J. B. Edgar (5) i L. J. Federico R. G. Grummer M. J. Hibbard T. H. Keheley T. L. Krysinski P. M. O' Leary J. R. Tandy C. J. Volmer G. N. Ward H. E. Williamson Document Control (5) 1 e
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