ML20195H010
| ML20195H010 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 03/29/1988 |
| From: | Oleary P, Tandy J, Jason White SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML19292H967 | List: |
| References | |
| ANF-87-95, ANF-87-95-R03, ANF-87-95-R3, TAC-66524, TAC-66525, TAC-68688, NUDOCS 8806280200 | |
| Download: ML20195H010 (25) | |
Text
a s a a ns t_ s u._.k- A ab,.r '
.gaa- a-44 _.b,LA. ao a m.__.ua s. ,,_.,
, 7.;
- Georgia Power A 1
ENCLOSURE 4 HATCH 9X9 LEAD FUEL ASSEMBLIES SAFETY ANALYSIS' REPORT.
ADVANCED NUCLEAR FUELS CORPORATION REPORT ANF - 87.- 95 b~
s P
2055C
.SL-4676 E4-1 06/20/88 a
' 700775
ENCLOSURE 4 ANF-87-95 REVISION 3
+ ,
11 ADVANCED NUCLEAR FUELS CORPORATION HATCH 9x9 LEAD FUEL ASSEMBLIES SAFETY ANALYSIS REPORT MARCH 1988
&KWU
. l l
ADVANCED NUCLEAR FUELS CORPORATION 1
ANF-87-95 Revision 3 ;
Issue Date: 3/29/88 HATCH 9X9 LEAD FUEL ASSEMBLIES SAFETY ANALYSIS REPORT Prepared By:
- 0. A/
/
J. A. White, Erig~ilieer BWR Safety Analysis Licensing and Safety Engineering Fuel Engineering and Technical Services A$
P. M. O' Leary g ngineer BWR Neutronics Neutronics and Fuel Management Fuel Engineering and Technical Services 3 I /l !I Engineer J. /R. Tandy,,ign BWR Des Fuel Design
,,,,avareer..aurau a v c ,
&MWU
(
1 CUSTOWER DISCLAIMER RAPORTANT NOTICE REGAROpeQ CONTENTS ANO USE OF THIS OOCUMENT PLEASE READ CAREPULLY henced Nucteer Fuste Corporenon's warrettues and represemanons con-comang me euctoct menor of tras occument are moes set form in me Agreement between Mvenced Nucseer Fuses Corporenon and me Customer pursuant to wfucri tra document as amoued. hWi, except as omervnee expresery on>
mood n ucn Agreement nesmer Mvenced Nucsear Fuete Corporaron nor any person actW on to benest mawas any warranty or representanon, expressed or irreeled, with 'espect to the accuracy, completences, or usefumees of me infor.
menon corm 1 lrt mee occumert or met the use of affy informabon, accaratus, momed or proct 1 N in thee document vnel not irtmnge ortvetery owned .
ilgnto; or answnsh trvy am witti respect to me use of arty eformanon, ao. )
paratus, momed or vocess W in true occumerft. ,
The eformenon centes.M herem se for me so6e use of Customer I
in order to swomiimperrn
- of ngrtto of Advanced Nucsear Fuele Corporenon in potente or rtwencons wrucft i'ey be inciuced in me informenon contamed in this documers me reciosent. Dy a acceptance of mes document, agrees not to puttien or messe puosic use (m tA w use of me terrn) of sucn informacon untd so authermed m vmtmg Oy Adver*1 Nucseer Fuese Coracionen or untd after six (6) months following termineon or eAveDon of the aforseesd Agreernent and arty emeneen mereof, unises amerwee woewy primoed m me Agrooment. No rignto or liconese in or to arty potente ars'irrWiled by me fumietting of this occu-ment.
ANF-3145 4M.A (12/87)
ANF-87-95 Revision 3 TABLE OF CONTENTS Section Elag
1.0 INTRODUCTION
....................................................... I 2.0 FUEL MECHANICAL DESIGN ANALYSIS.................................... 3 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS.................................. 4 3.1 Hydraulic Compatibility............................................ 4 3.2 Th e rmal M a rg i n Pe r fo rman c e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 3.3 Single Loop 0peration.............................................. 4 4.0 NUCLEAR DESIGN ANALYSIS............................................ 6 4.1 Standby Liquid Control System...................................... 6 4.2 Col d S h u tdown Ma rg i n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 4.3 Fuel Pool Criticality.............................................. 6 5.0 ANTICIPATED OPERATIONAL OCCURRENCES................................ 7 5.1 Overpower Events................................................... 7 5.2 Control Rod Wi thdrawal Error. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 5.3 Fuel Mislocation Error............................................. 8 5.4 Fu el Ro t a t i o n E rro r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 6.0 POSTULATED ACCIDENTS............................................... 9 l 6.1 Lo s s -O f-Cool ant Acci den t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 1
- 6.2 Control Rod Drop Accident.......................................... 11
(
1 6.3 Fuel Handling Accident............................................. 11 7.0 TECHNICAL SPECIFICATIONS........................................... 12 7.1 Limiting Safety System Settings.................................... 12 7.2 Limiting Conditions For 0peration.................................. 12 7.3 Surveillance Requirements.......................................... 12
8.0 REFERENCES
......................................................... 14 l
l
S
- ii - ANF-87-95 Revision 3 LIST OF TABLES Table P.Agg 1 Compari son Of Bundl e Powers At MCPR Limi t. . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 2 Comp ari s on Of APLHGR L imi t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 LIST OF FIGURES Fiaure Engg 1 - Hatch APLHGR Limit Comparison Based On Equivalent Planar Power. .. .... 17 2 Hatch 9x9 LFA's LHGR Compared To ANF 9x9 LHGR Limit.................. 18
. l 1 ANF-87-95 Revision 3
1.0 INTRODUCTION
Evaluations have been performed to define the impact upon the core performance as a result of inserting four (4) ANF 9x9 lead fuel assemblies (LFA's) manufactured by Advanced Nuclear Fuels Corporation (ANF) into the Hatch Nuclear Power Stations. In addition, justifications are provided which demonstrate that application of GE P80RB284H 8x8 operating limits, as defined in the Technical Specifications, to the ANF 9x9 LFA's is acceptable and will not result in decreasing the reactor's margin to safety during operation.
Revision 1 of this document addressed application of ANF 9x9 fuel in Hatch Unit 2. This revision expands the application of the ANF 9x9 fuel to both Hatch Units I and 2. The expanded application covers operation with the recently NRC approved (l) GE SAFER /GESTR analysis regarding LOCA-ECCS analysis.
The insertion of only four ANF 9x9 assemblies will have negligible effects upon the core-wide transient performance relative to the core fully loaded without the four 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 GE P8DRB284H 8x8 fuel assemblies. This includes the analyses of anticipated plant transients, LOCA, and stability which are used to support ARTS, extended load 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 P80RB284H 8x8 fuel assembly. Therefore, existing fuel storage limits for GE fuel bound those necessary for the ANF 9x9 LFA's.
Analyses performed for the 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 virtuo of
2 ANF-87-95 Revision 3 the assemblies meeting compatibility requirements of reactivity and hydraulic demand.
The evaluations provided herein thus provide assessments of the ANF 9x9 assemblies relative to the GE P80RB284H 8x8 fuel assemblies and justify application of the current Hatch Technical Specifications for that fuel to the ANF 9x9 fuel assemblies.
, , + . , - - - - - .. - - - , ,, - - . . . - . , , , , , , . , , , , , , , , ,, .-,-,.4
l 1
3 ANF-87-95 !
Revision 3 1
2.0 FUEL MECHANICAL DESIGN ANALYSIS The expected operating requirements of Hatch Units 1 and 2 are bounded by the assumed power history in ANF's fuel mechanical design analyses (2). 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(3). These evaluations also assure that the four ANF 9x9 LFA's will meet operating and safety design requirements of the Hatch Nuclear Power Stations.
4 ANF-87-95 l Revision 3 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS 3.1 Hydraulic Comoatibility Component hydraulic resistances for the ANF and GE fuel assemblies have been determined in single phase flow tests of full scale assemblies. There is no significant difference between the GE fuel assembly which was tested and Hatch reload fuel assemblies since both are the standard GE 8x8R product offering.
The small mechanical design differences between the ANF 9x9 test assembly and the Hatch LFA's are not significant from a thermal hydraulic standpoint.
Hydraulic compatibility of the ANF 9x9 and GE coresident fuel types has been demonstrated (4).
3.2 Thermal Marcin Performance -
Analyses of the limiting BWR/4 trisients have shown that the bundle power needed to produce transition boiling in the ANF 9x9 fuel is higher than that for the GE 8x8 fuel. Table 1 shows that the ANF 9x9 fuel must be operated 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 reach their actual MCPR limit. ANF analyses in support of extended operating domains for a typical BWR/4 show the equivalence of 8x8 and 9x9 MCPR limits throughout extended operating domains. It follows that monitoring the ANF 9x9 j LFA's based on GE 8x8 MCPR limits adequately protects the ANF 9x9 LFA's from boiling transition.
3.3 Sinale looo Ooeration l ANF analysis of a typical BWR/4 with a full ANF 9x9 reload in Single Loop I
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 l without bypass valve operation. This analysis showed that single' loop limits
- - - - - - - - + - - - --, , y- 4 .mm -9
. 1 1
5 ANF-87-95 Revision 3 ,
are unaffected by the introduction of the ANF- 9x9 LFA's. In addition, j monitoring the ANF 9x9 assemblies with GE P8DRB284H 8x8 fuel limits for SLO results in a conservative estimate of the margin to critical power for the 9x9 fuel in single loop operation.
.r - -. , . , --. - --.--,,.-------,--n.,-- - , - - - - - - - - -
6 ANF-87-95 Revision 3 4.0 NUCLEAR DESIGN ANALYSIS 4.1 Standby Liouid Control System The neutronic impact of replacing four of the 560 fuel assemblies with ANF 9x9 LFA's which demonstrate similar reactivity characteristics will be negligible on the standby liquid control system reactivity worth.
4.2 Cold Shutdown Marain Infinite assembly calculations at 0 GWd/t show the ANF 9x9 LFA'.s to have approximately 0.6 mk higher cold uncontrolled reactivity relative to the GE P8DRB284H 8x8 fuel. This results in a control cell reactivity less than 0.2 mk higher than an all GE P80RB284H 8x8 fuel loaded control cell, which is a negligible contribution to cold shutdown margin. For exposures greater than 1.8 GWd/t the ANF 9x9 LFA design has slightly lower :old uncontrolled reactivity than for the GE P8DRB284H 8x8 fuel. This results in a slight increase in cold shutdown margin for a control cell with an ANF 9x9 fuel assembly in place of a GE P8DRB284H 8x8 fuel assembly at exposures graiter l than 1.8 GWd/t. 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.
4.3 Fuel Pool Criticality The maximum k. of an ANF 9x9 LFA is approximately 2 mk less than a GE P80RB284H 8x8 fuel assembly. Therefore, spent fuel storage critical limits existing for GE 8x8 fuel bound those required for the ANF 9x9 LFA's.
l l
i 1
7 ANF-87-95 Revision 3 5.0 ANTICIPATED OPERATIONAL OCCURRENCES l
Operation of the four ANF 9x9 LFA's using GE P80RB284H 8x8 fuel MCPR operating limits is conservative. Analyses of the limiting BWR/4 transients have shown that the ANF 9x9 bundle power at the Technical Specification MCPR operating limit is higher than for a GE 8x8 fuel bundle.
5.1 Overoower Events The limits in effect for the GE P8DRB284H 8x8 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 limits (2). This compares to approximately 20 percent margin for GE 8x8 fuel regarding overpower transient LHGR limits. -
ANF has reviewed the MAPFACf curve in the Hatch Technical Specifications, and concluded that monitoring the ANF 9x9 LFA's to the GE P8DRB284H limits will protect the fuel from failures. The MAPFACf curve protects the fuel from exceeding LHGR limits in the case of a flow transient which starts from less than rated flow. The estimated margin of safety using the ANF protection against fuel failure (PAFF) limit is 44 percent at 20 GWd/t, while monitoring to the GE limit results in an estimated margin of safety of 63 percent at 20 GWd/t. At 35 GWd/t, the margin of safety is equal between ANF and GE limits at approximately 30 percent.
ANF has also reviewed the MAPFACp curve in the Hatch Technical Specifications, and concluded that the GE MAPFACp curve will protect the ANF 9x9 LFA's from fuel failures. The MAPFAC p curve protects the fuel from exceeding LHGR limits in the event of a power transient which starts from less than rated power.
l 8 ANF-87-95 Revision 3 5.2 Control Red Withdrawal Error Infinite assembly calculations of the control rod worth for the ANF 9x9 '_FA's and the GE P80RB284H 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 P80RB284H 8x8 fuel. Thus, the ACPR values for the ANF 9x9 fuel design will not be substantially different than those obtained for the GE P80RB284H 8x8 fuel and are within the variation that is seen between specific reactor cycles for a reactor which utilizes GE 8x8 fuel.
5.3 Fuel Mislocation Error The consequences of the mislocation of an ANF 9x9 LFA are no more limiting than that associated with the GE P80RB284H 8x8 fuel. This is substantiated by a comparison of the reactivity valtes between the two fuel types. The 9x9 values are comparable and in most cases less than that associated with GE fuel, thus the change in local power due to the mislocation of a 9x9 fuel assembly is no greater than that obtained by the mislocation of a GE assembly.
Thus, the mislocation ACPR for the ANF 9x9 fuel design is not significantly different from those for the GE P80RB284H 8x8 fuel.
5.4 Fuel Rotation Error The consequences of the fuel rotation error have been evaluated comparing the ANF 9x9 LFA design to the GE P80RB284H 8x8 fuel design. The results indicate an increhse in ACPR for the rotated ANF 9x9 LFA of up to 0.06 relative to a rotated GE P80RB284H 8x8 fuel assembly. Typically, the rotated 8x8 fuel has not been the limiting event for Hatch, and more than 0.06 ACPR margin has existed to the MCPR operating limit. If necessary, selection of 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.
9- ANF-87-95 Revision 3 6.0- POSTULATED ACCIDENTS 6.1 loss-Of-Coolant Accident The appropriate bundle power limit derived from a LOCA analysis is the peak bundle-planar power because heatup is primarily a planar phenomena, not an
. axial phenomena. The bundle is contained in a channel and the peak clad temperature (PCT) is primarily governed by rod-to-rod and rod-to-channel radiation, and local convection. Presently for Hatch, the peak bundle-planar power determined from the LOCA analysis is converted to a Maximum Average Planar LHGR limit (MAPLHGR) by dividing the peak planar power by the number of heated rods in a bundle; this MAPLHGR limit is used as the LOCA monitoring limit. Alternatively, this peak bundle-planar power could be directly used as the LOCA monitoring limit; in this report this alternate limit is termed equivalent planar power.
ANF 9x9 fuel has equivalent or improved LOCA-ECCS performance when compared to both ANF 8x8 and GE 8x8 fuel for two fundamental reasons. First, because of its lower LHGR's for the same planar power, ANF 9x9 fuel has less stored anergy than 8x8 fuel. Seco'ndly, ANF 9x9 fuel has better heat transfer characteristics because of the greater surface area per unit volume. Of further benefit is that ANF fuel has a larger upper tie plate flow area than GE 8x8 fuel, resulting in less restrictive countercurrent flow limiting (CCFL) characteristics.
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 the ANF fuel is less restrictive than GE fuel. This remains the case for most bundle . exposures .. Figure I shows a comparison of- APLHGR limits converted to equivalent planar power for the ANF 9x9 and the GE P80RB284H 8x8 fuel. In order to provide a comparative bases between 8x8 and 9x9 arrays, the equivalent planar power is shown as the APLHGR limit times the number of fuel rods per assembly. The GE APLHGR limits are based on GE's SAFER /GESTR LOCA
10 ANF-87-95 Revision 3 methodology for their P80RB284H fuel, recently approved by the NRC for application to Hatch Units 1 and 2(1).
The ANF 9x9 maximum planar power curve crosses the GE P80RB284H 8x8 fuel curve at approximately 32 GWd/t. The fact that the curves cross does not present an operational problem. ANF 9x9 fuel is not peak clad temperature (PCT) limited past 20 GWd/t, but rather LHGR limited. Figure 2 shows a plot comparison of the ANF 9x9 LHGR limit (2) and tlie maximum 9x9 LHGR values in the Hatch plants monitored to GE limits. The conclusion drawn from Figure 2 is that the ANF 9x9 LFA's in Hatch will not be LHGR limited. Monitoring the ANF 9x9 LFA's to GE limits will assure that the criteria specified in 10 CFR 50.46 will be satisfied for the four ANF 9x9 LFA's.
Generic analysis has shown that for plant LOCA-ECCS performance consideration, BWR/4's can be grouped into two major subgroups--those with loop selection logic (i.e., plants that have not incorporated low pressure coolant injection (LPCI] sy1tep modification) and those which have LPCI modification (5). Since Hatch falls into the latter subgroup and ANF has performed a LOCA analysis for a BWR/4 with LPCI modifications (3), ECCS performance differences can be considered insignificant.
The recently implemented Technical Specification modifications (l) resulted in the following:
o MAPLHGR values for the P8DRB284H 8x8 fuel were modified, o Core spray response time changed from 27 seconds to 34 seconds, o LPCI response time changed from 40 seconds to 64 seconds, o The required core spray flow rate decreased from 4625 gpm to 4250 gpm.
11 ANF-87-95 Revision 3 It was established above that the ANF LFA's can be monitored to the GE reference fuel limits and not violate the PCT limit of 10 CFR 50.46 or the ANF LHGR limits. The change in the maximum response time for core spray and LPCI will not affect the ANF typical BWR/4 analysis. The response time used in the ANF analysis is 60 seconds for the core spray system and 90 seconds for the LPCI system. Calculations were made using an equivalent lower core spray flow rate for a typical BWR/4 analysis. The results confirued that ANF 9x9 LFA's will meet 10 CFR 50.46 when monitored to GE limits.
6.2 Control Rod Oroo Accident The consequences of a control rod drop accident (6) have been determined by ANF to be a function of dropped rod worth, Doppler reactivity, delayed neutron fraction, and fucl rod local peaking. A comparison of these parameters between the ANF 9x9 and the GE 8x8 fuel indicates that the deposited enthalpy for the ANF 9x9 fuel will have a value comparable to that calculated for the GE 8x8 fuel and maintain sufficient margin to the limit of 280 cal /gm.
6.3 Fuel Handlino Accident A comparison of the radiological consequences of fuel handling accidents with 8x8 and 9x9 fuel for a typical BWR/4(3) showed less radioactivity released for the 9x9 fuel.
- f. . . .
12 ANF-87-95 Revision 3 7.0 TECHNICAL SPECIFICATIONS 7.1 Limitina Safety System Settinas The four ANF 9x9 LFA's will not materially affect the safety limits of Hatch operation.
7.2 Limitino Conditions For Ooeration ANF analysis of a typical BWR/4(3) has shown that the ANF 9x9 bundle power at the MCPR operating limit is higher than for a GE 8x8 fuel. It follows that application of the GE P80RB284H 8x8 fuel MCPR limits to the ANF 9x9 LFA's adequately protects the LFA's from boiling transition.
Restricting the ANF 9x9 LFA's to the planar power consistent with GE APLHGR limits protects ANF 9x9 APLHGR and LHGR limits. As discussed in the previous
,. section, GE APLHGR limits for the GE P80RB284H 8x8 fuel type in Hatch Units 1 and 2 will protect the ANF 9x9 LFA's from violating limits. ANF 9x9 APLHGR limits are more restrictive or equivalent to (depending on exposure) ANF 9x9 LHGR limits.
7.3 Surveillance Reouirements Stability tests have been performed on the Commonwealth Edison Company's Dresden Unit 2 reactor with ANF 9x9 LFA's in core. The results of these tests indicate that the ANF LFA's have no measurable impact on local stability.
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(7) Detect and Suppress region. ;
13 ANF-87-95 Revision 3 l
l The Hatch Units 1 and 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 the Hatch plants meets the requirements of GDC 10 and 12(8),
. . - - , _ - - _ . . _ - - , . _ _ _ _ _ _ _ - . - _ _ - _ . . . . ~ . _ . . . _ _ _ _ , - - - - _ _ . , - _ , _ . _ , . . _ , _ _ . , . , - _ - - - - - - _ _ _ _ _ , - -
14 ANF-87-95 Revision 3
8.0 REFERENCES
- 1. Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendments 150 and 87 to Facility Operating Licenses DPR-57 and NPF-5, Georgia Power Company, Edwin I. Hatch Nuclear Plant, Units 1 and 2, Docket Nos. 50-321 and 50-366.
- 2. "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel,"
XN-NF-85-67(A), Rev. 1, September 1986.
- 3. Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment 31 to Facility Operating License No. NPF-22, Pennsylvania Power and Light Company, Susquehanna Steam Station Unit 2, Docket No. 50-388.
- 4. "Hatch Lead Assembly Compatibility Report - Mechanical, Thermal and Neutronic Design for ANF 9x9 Fuel Assemblies," XN-NF-87-77(P.)., Rev. 1, September 1987.
- 5. General Electric Licensing Topical Report, Generic Reload Fuel Application, NED0-240ll-2.
- 6. "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Ar.alysis," XN-NF-80-19( A), Volume 1 and Supplements 1 and 2, March 1983.
- 7. General Electric Information Letter, "BWR Core Thermal Hydraulic Stability," SIL No. 380, Revision 1, February 10, 1984.
- 8. Code of Federal Regulations, 10 CFR 50, Appendix A.
15 ANF-87-95 Revision 3 TABLE 1 COMPARISON OF BUNDLE POWERS AT MCPR LIMIT (BASED ON TYPICAL BWR/4)
BUNDLE POWER FUEL TYPE AT MCPR LIMIT (MW)
ANF 9x9 6.7 GE 8x8 6.4
l 9
16 ANF-87-95 Revision 3 TABLE 2 COMPARISON OF APLHGR LIMITS (BASED ON TYPICAL BWR/4)
EQUIVALENT PLANAR POWER PEAK APLHGR (APLHGR LIMIT
- NO. OF FUEL LIMIT (KW/FT) RODS) (KW/FT)
GE 8x8 i e 12.4 769 (BWR/4',
ANF 9x9 Fuel 10.2 806 (BWR/4) 6
~
l 900 ANF 9x9 Limit
- -------- -- GE 8x8 Limit (P8DRB284H) 800-m.>
N ,. - .,
y ..- .
x ..- ..
w .- ..
U 700-y .-
o ~
L u ' ... .
e a .
. ~ .
c . ~
4 m
g 600-
- s B .'..
H '.
6
- s -'..
500-
- n >
m2
< n I I I I 5 5 5 3
0 5 10 15 20 25 30 35 40 45 ose7' Planar Exposure (GWd/t) w Figure 1 Hatch APLHGR Limit Comparison Based on Equivalent. Planar Power
I e
w i
ANF 9X9 LHGR Limit Note:
Maximum 9x9 LHGR Max. 9x9 LHGR = P8DRB284H MAPLHGR Limit
! x Normalized Peak Rod Power x (62/79) 1 l
+ -
w n
N k
a v ....,...
o ....
o ...... .......................... - .. -
p: en u
a.
e-
- n >
mz
<n e . .
0 5 1O l5 NO ' N5 $0 ' $5 4O 45 aw o
Planar Exposure (GWd/t) w
~
Figure 2 Hatch 9x9 LFAs LHGR Compered to ANF 9x9 LHGR LIMIT .
e h
ANF-87-95 Revision 3 Issue Date: 3/29/88 HATCH 9X9 LEAD FUEL ASSEMBLIES SAFETY ANALYSIS REPORT Distribution D. A. Adkisson O. C. Brown R. E. Co11ingham L. J. Federico R. G. Grummer J. G. Ingham T. H. Keheley T. L. Krysinski T. E. Millsaps (5)
P. M. O' Leary J. R. Tandy C. J. Volmer G. N. Ward J. A. White H. E. Williamson Doc =:r.t Control (5)