ML20196E927
| ML20196E927 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 11/30/1988 |
| From: | SYSTEM ENERGY RESOURCES, INC. |
| To: | |
| Shared Package | |
| ML19295G794 | List: |
| References | |
| NUDOCS 8812120138 | |
| Download: ML20196E927 (31) | |
Text
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GRAND GULF NUCLEAR STATION UNIT 1 CYCLE 4 RELOAD S'JMMARY REPORT November 1988 i
SS12120138 881206 PDR ADOCK 05000416 p FDC SLtMARY - 1
CONTENTS Page 4
1.0 INTRODUCTION
5 2.0 CYCLE 4 RELOAD SC0PE.............................................
6 3.0 CfCLE 3 OPERATING HIST 0RY........................................
' 7 4.0 CYCLE 4 CORE DESCRIPTION.........................................
8 5.0 FUEL MECHANICAL 0ESIGN...........................................
i, 10 6.0 THERMAL HYORAULIC 0ESIGN..........;..............................
4 6.1 Safety Linit MCPR........................................... 10 6.2 Core Bypass F1cw............................................ 10 11 l 6.3 Core Stability..............................................
' 11 7.0 NUCLEAR DESIGN...................................................
i 11
- 7.1 Fuel Bundle Nuclear 0esign..................................
12 7.2 Core Reactivity.............................................
7.3 S pent Fu el Pool Cri ti c al ity. . . . . . . . . . . . . . . . . . . . . . .'. . . . . . . . . 13 l
14 8.0 CORE MONITORING SYSTEM...........................................
l -
l l
SUMMARY
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t L
CONTENTS (cont'd)
P,gg 14 9.0 ANTICIPATED OPERATIONAL OCCURRENCES..............................
15 9.1 Core-Wide Transients........................................
16 9.2 L oc al T r a ns i e n t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
16 9.3 Reduced Flow and Power Operation. . . . . . . . . . . . . . . . . . . . . . . . . . . .
18 9.4 ASME Overpressurization Analysis............................
19 10.0 POSTULATED ACC10ENTS.............................................
19 10.1 L o s s -of-Cool ant Acc ide nt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
20 10.2 Rod Drop Accident...........................................
20 11.0. REFUELING OPERATIONS............,.................................
21
12.0 REFERENCES
APPEN0!CES...........................................................
25 APPENDIX A: 9 x9-5 LE AD TEST ASSEMBLIES. . . . . . . . . . . . . . . . . . . . . . . . .
APPENDIX B: ENVIRONMENTAL EFFECTS OF TRANSPORT OF 31 NUCLEAR FUELS......................................
SUMMARY
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1.0 INTRODUCTION
This repet is a supplementary document which provides a general scope and sum nizes the results of the analyses performed in support of GGNS Unit 1 !
Cycle 4 operation. This report includes a description of the new ANF 8x8 fuel type and the ANF 9x9-5 Lead Test Assemblies (LTAs) which will be I t
inserted into the core for Cycle 4. The fresh fuel to be inserted in this !
i cycle is similar to that used in previous reloads except for changes in enrichment and gadolinia loadings. Four 9x9-5 LTAs will be loaded in order to monitor performance in anticipation of 9x9-5 fuel reloads for subsequent cycles. A description of the LTAs and their licensing basis is I presented in Appendix A. The fresh fuel to be loaded for Cycle 4 is !
designed for higher discharge exposures than that of previous reloads.
Environmental effects considerations for transportation of Cycle 4 reload i
fuel and compliance with applicable regulations are addressed in Appendix !
B. ,
t
[
t The ANF /C cle 4 Reload Analysis Report (Reference 1) and the Cycle 4 Plant Transient Analysis Report (Reference 2) serve as the basic framework for the reload analysis. Where appropriate, reference is made to these and other supporting documents for more detailed information and/or l specifics of the applicable analyses. The ANF Reload Analysis Report is intended to be used in conjunction with ANF topical report XN-NF-80-19(A),
Vol. 4, Revision 1. ' Exxon Nuclear Methodology for Boiling Water Reactors: !
Application of the ENC Methodology to BWR Reloads" (Reference 4), which l I
L describes the analyses performed in support of the reload and identifies the methodology used for those analyses. A list of references is provided ,
SUWARY - 4
- containing the GGNS specific documents and the applicable generic documents prepared by ANF which are being used in support of the Cycle 4 reload submittal.
- 2. 0 CYCLE 4 RELOAD SCOPE During the third refueling outage at GGNS Unit 1, System Energy Resources, Inc. (SERI) will be replacing the remaining GE initial core fuel assemblies and some of the'first ANF reload fuel assemblies with ANF 8x8 fuel assemblies. The Cycle 4 core will contain ANF supplied fuel exclusively. Limiting cases of fuel related analyses previously performed for Cycle 3 were repeated for Cycle 4. This includes analyzing Cycle 4 for anticipated operational occurrences to confirm operating limits, performing LOCA confirmatory analyses for compliance with 10CFR50.46, and analyzing for the rapid drop of a high worth control rod to assure that excessive energy will not be deposited in the fuel. Additional analyses were performed to support revised thermal limits for the full ANF core based on ANF's methodology. Analyses for normal operation of the reactor consisted of fuel evaluations in the , areas of mechanical, thermal-hydraulic, and nuclear design.
Based on ANF's design and safety analyses of the Cycle 4 reload core, the GGNS Unit 1 Technical Specifications are changed as follows:
- a. MAPLHGR curve for tne new 8x8 fuel type is added.
SUMMARY
- 5
- b. MAPLHGR curve to be used in determining limits during single loop operation is revised.
- c. Flow-dependent thermal limits for off-rated conditions (MAPFAC and MCPR f
) are revised, f
3.0 CYCLE 3 OPERATING HISTORY Cycle 3 core-follow operating data available at the time of the reload .
design analysis, together with projected plant operation through the end of Cycle 3 was used as a basis for the Cycle 4 core design and as input to the plant safety analyses. Cycle 3 has continued as expected and no operating anomalies have occurred which would affect the licensing basis for Cycle 4.
The end-of-cycle 3 (EOC3) licensing exposure window ranges from 1340 GWd to 1570 GWd with a nominal exposure of 1455 GWd. This window provides an allowable EOC3 core average exposure range for which the Cycle 4 plant safety analyses are valid.
SUMMARY
- 6
4.0 CYCLE 4 CORE DESCRIPTION The Cycle 4 core will consist of i)0 fuel assemblies which include 272 fresh ANF 8x8 assemblies (third re oad), 288 once burned ANF 8X8 assemblies (second reload), 236 'tice burned ANF 8x8 assemblies (first l reload) and 4 ANF 9x9-5 LTAs. A breakdown by bundle type / bundle average enrichment is provided in the following table:
Number of Bundles Bundle Type 272 ANF 8x8/3.37 w/o U235 l w/8 rods 4.0/5.0 w/o Gd203 4 ANF 9x9/3.25 w/o U235 l
w/8 rods 5.0/6.0 w/o Gd203 204 , ANF 8x8/3.01 w/o'u235 w/6 rods 4.0 w/o Gd203 84 ANF 8x8/3.01 w/o U235 ,
! i w/8 rods 4.0 w/o Gd203 236 ANF 8x8/2.81 w/o U235 ;
w/5 rods 3.0 w/o Gd203 ,
I Of the 276 assemblies being discharged at EOC3, 248 are the remaining GE i 8x8 bundles and 28 are ANF 8x8 first reload bundles.
The anticipated Cycle 4 core configuration along with additional core design details is provided in section 4.0 of the ANF Cycle 4 Reload Analysis Report (Reference 1). The Cycle 4 core is a conventional scatter 1
SUMMARY
- 7 , {
load with the lowest reactivity bundles placed in the periphery region of the core. The loadinc pattern was designed to maximize the cycle energy and minimize power peaking factcrs. Cycle 4 is estimated to provide 1,698 GWd of energy based on a Cycle 3 energy output of 1455 GWd.
i 5.0 FUEL MECHANICAL DE51GN j
The mechanical design analyses for the ANF 8x8 fuel (first, second and third reloads) are described in XN NF-85-67(P)(A), Revision 1 (Reference 5). The8x8ANFfuelassembl/designcontains62fuelrodsandtwowater rods, one of which functions as a spacer capture rod. Seven spacers maintain fuel rod spacing. The fuel rods are prepressurized, contain U02
{ pellets and use a diametral pellet-to clad gap which is smaller on the
'1 l interior high enrichment rods than on the remaining rods in the bundle to -
improve ECCS margin.
I,
) Mechanical design analyses were p*rformed to evaluate cladding steady-i state strain, transient stresses, fatigue damage, creep collapse, l corrosion buildup, hydrogon absorption, fuel rod maximum internal l
I pressure, differential fue.' rod growth, creep bow, and grid spacer spring design. These analyses were performed to support a batch average discharge burnup of 34,000 MWo/MTV. All parameters meet their respective design limits as shown in Reference 5. This reference presents the fuel thermal analysis that shows no fuel centerline melting at 120% overpower conditions for all exposures within the design end-of-life exposure.
SUMMARY
- 8
l The Reload 3 fuel design increases the maximum batch average exposure from 4
30,000 mwd /MTU to 34,000 mwd /MTV and the maximum assembly exposure from 33,000 mwd /MTU to 39,000 MVd/MTU (Reference 10). For Reload 3 fuel the design is such that margin to fuel mechanical design limits (e.g.,
centerline melt, transient stress, etc.) is assured for overpower conditions throughout the life of the fuel as demonstrated by the fuel -
design analyses.
As described in Reference 1, the MAPLHGR operating limit has been defined I
for each fuel type to ensure conformance with the LHGR mechanical design
- limit. For all expected Cycle 4 operations, conformance to the MCPR, MAPLHGR and LHGR operating limits ensures that tne power distribution for I ANF fuel remains within the assumptions of the fuel design analyses.
i The mechanical response of the ANF assembly design during seismic-LOCA l events is essentially the same as the response of a GE assembly because the physical properties and bundle natural frequencies are similar.
Reference 6 presents the seismic LOCA analysis for the GE fuel which shows that resultant loadings do not exceed the fuel design limits.
! Reference 7 presents the seismic-LOCA analysis for ANF fuel which showed i
large design margins for all assembly components. Therefore, based on the l
similarity between the fuel types and the large margin calculated for ANF i
fuel in a similar application, the loadings for GGNS Unit I do not exceed 1 design limits for ANF fuel assembly components.
SUMMARY
+ 9
_ __ . _ _- -_ . _ _ . - _ , _ _ _ _ _ _ _ _ - _ - - _ - . __ . _ _ ~
6.0 THERMAL HYORAULIC DESIGN XN-NF-80-19(A), Volume 4, Revision 1 (Reference 4) contains the thermal hydraulic design criteria which are used in the determination of the fuel cladding integrity safety limit and bypass flow characteristics. ANF analyses were performed in accordance with XN-NF-80-19(A), Volume 3, Revision 2 (Reference 19) to demonstrate compliance with these design criteria. The analyses performed to determine each of these parameters are discussed below.
6.1 Safety Limit MCPR i
I The MCPR fuel cladding integrity safety limit remains 1.06. The methodology and generic uncertainties used in the Cycle 4 MCPR safety limit calculation are provided in Reference 8. The MCPR safety limit calculation and GGNS Unit 1 specific inputs are provided in Reference 2.
l I 6.2 Core Bypass Flow Core bypass flow is calculated using the the methodology described in XN NF-80-19(A). Vale-e 3. Revision 2 (Reference 19). The core bypass l
flow traction, excluding water rods, for Cycle 4 is 10.0% of total core .
flow (Reference 1).
l l
SUMMARY
- 10 L
6.3 Core Stability ;
A GGNS Unit 1 Technical Specifications proposed change is being submitted under a separate cover. This has been discussed with and agreed to by the staff in a meeting on November 9, 1983.
7.0 NUCLEAR DESIGN The neutronic methods used for the design and analysis of ANF follow-on reloads are described in the ANF topical report XN-NF-80-19(A), Volume 1 and Supplements 1 and 2 (Reference 9). These methods have been reviewed ,
and approved by the Nuclear Regulatory Commission for generic application to BWR reloads, i
7.1 Fuel Bundle Nuclear Desian The Cycle 4 reload fuel bundle design utilizes ANF 8x8 fuel assemblies.
The 8x8 lattice consists of two inert (water) rods and 62 fuel rods containing 150 inches of active fuel each. The top and bottom six inches of each fuel rod contain natural urani,pm and the central 138 inches (enriched zone) of each rod contains enriched uranium at one of five different enrichments. The fuel bundle burnable poison design utilizes eight gadolinia bearing rods. These rods contain 5.0 w/o Gd230 in the lower 114 inches of the enriched fuel zone and 4.0 w/o Gd 023 in the l remaining top 24 inches. They are utilized to control the reactivity and power peaking of the core.
SUMMARY
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l The average enrichment of the bundle enticned zone is 3.6% w/o U235 and t the bundle average enrichment (including the top and bot'.om natural uranium blankets) is 3.37 w/o U235. The number of fuel sds at each enrichment is given below:
f i
Number of Rods Enrichment (w/o U235) of Enriched Zone 1 1.50 5 2.00 ;
8 2.90 i
28 3.48 (8 containing 4.0/5.0 w/o 2 '
Gd230) 20 4.57 The neutronic design parameters and rod enrichment distribution are .
]
described in section 4.0 of the Cycle 4 Reload Analysis Report (Reference l
1). !
l 7.2 Core Reactivity .
1 The beginning of Cycle 4 (80C4) cold core K,ff value with all-rods-out was calculated to be 1.13315. Based on a minimum Cycle 3 length of 1,340 GWd, l ]
a minimum Shutdown Margin of 1.11% delta k/k (with the strongest worth i
control rod fully withdrawn at cold, 68 degrees F reactor conditions) was i
determined to occur at the beginning of Cycle 4. Therefore, the ,
l difference between the minimum Shutdown Margin in the cycle and the BOC ,
- Shutdown Margin, R, is 0.00% delta k/k. The calculated Shutdown Margin is
, i SUtHARY - 12 i
) ,
-__,. ..~-._-_.- _.,__ _ .._ _,____,_, , ,. ,_ - ___._._-.. __ _ _ _ _ . . . - , - - _ _ _ _-
well in excess of the 0.38% delta k/k Technical Specification requirement, and will be verified by testing at 80C4 to be greater than or equal to R +
0.38% delta k/k.
The Standby Liquid Control (SLC) system, which is designed to inject a quantity of boron that produces a concentration of no less than 660 ppe ,
l in the reactor core within approximately 90 to 120 minutes after !
initiation, was calculated to provide a minimum shutdown margin of at f
least 3.93% delta k/k with the reactnr in a cold, xenon free state, all control rods in their critical full power positions, and the reactor at the most limiting cycle exposure. These are the most conservative analytical assumptions and are therefore applicable to all plant operating i
conditions, including the ME00. This assures that the reactor can be brought from full power to a cold, xenon free shutdown, assuming that none of the withdrawn control rods can be inserted and thus for the Cycle 4 reload core, confirms the basis of the Technical Specification (
requirement, f t
7.3 Spent Fuel Pool Criticality :
A GGNS-1 specific High Density Spent Fuel Storage Rack (HDSFSR) criticality safety analysis was performed and submitted previously (Reference 3). This analysis shows that with the introduction of the higher enriched Cycle 4 fuel into the HDSFSR, the infinite multiplication factor of the HDSFSR remains at or below 0.936. This is below the NRC acceptance criteria of Kgf=0.95.
SUMMARY
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8.0 CORE MONITORING SYSTEM The POWERPLEX core monitoring system is and will continue to be utilized to monitor reactor parameters at GGNS. The POWERPLEX core monitoring system incorporates ANF's core simulation methodology and is used for both online core monitoring as well as an offline predictive and backup tool.
I I The system is fully consistent with ANF's nuclear analysis methodology as I described in XN-NF-80-19(A) Volume 1 Supplements 1 & 2 (Reference 9). In addition, the measured power distribution uncertainties are incorporated into the calculation of the MCPR Safety Limit as described in ANF's i
! Nuclear Critical Power Methodology Report XN-NF-524(A) (Reference 8).
h .
i 9.0 ANTICIPATED OPERATIONAL OCCURRENCES 1
1 In order to support the Cycle 4 operating limits, eight categories of 1
! system transients are considered as described in ANF's Plant Transient 1
' Methodology Report XN-NF-79-71(P) (Reference 11). ANF has provided plant
) '
specific analysis results for the following three system transients to f determine the thermal margin requirements for operation during Cycle 4
- (Reference 2)
- 1) Generator Load Rejection without Bypass (LRNB)
- 2) Feed.ater Controller Failure (FWCF)
- 3) Loss of Feed ater Heating (LFWH) As shown in XN-NF-79-71(P) l l
SUMMARY
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l l
t
l' As shown in XN-NF 79-71(P) (Reference 11), the other system transients are ,
non-limiting and, therefore, bounded by one of the above. In addition, i the Fuel Loading Error was analyzed in accordance with the methodology
)
described in XN-NF-80-19(A) Volume 1 (Reference 9). The Control Rod j
[
, Withdrawal Error (CRWE) transient has been analyzed generically in Reference 18 and is applicable to Cycle 4 (Reference 1). These analyses b
t
- - support a MCPR operating limit of 1.18 at rated operating condition.
9.1 Core Wide Transients i
l The plant transient codes ussd to evaluate the Load Rejection Without i
j Bypass (LRNB) and Feedwater Controller Failure (FWCF) events are the ANF's i COTRANSA cnde (Reference 11) and XCOBRA-T code (Reference 20), which incorporate a one dimensional neutronics model to account for siifts in j
! axial power shape and control rod effectiveness. Technical Specification scram times were used in the bounding analysis. The results of the LRNB
\
! and FWCF analyses are provided in the Cycle 4 Plant Transient Analysis i 3
Report (Reference 2) and a result summary is provided in the Cycle 4 I
i Reload Analysis Report (Reference 1).
l
! t l -
The Loss of Feedwater Heating (LFWH) event was analyzed within the ME00 l power / flow operating map for various cycle exposures anticipated during i
! Cycle 4. A summary of this analysis is provided in Reference 2.
4 :
i t l
f I l
SUMMARY
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_- _ _ - . _ . _ . . _ _ _ _ _ _ - - ~ _ _ _ _ . _ _ _ . - _ _ - _ _ _ _ _ . _ _ _ ,
9.2 Local Transients The Control Rod Withdrawal Error (CRWE) transient has been analyzed generically in Reference 18. The Reference 18 analysis provides a statistical evaluation of the consequences of 1.he CRWE transient for BWR/6 plant configurations under conditions s'aich cove' the normal operating power flow map, the extended load line region and the increased i core flow region. This analysis is applicable to GGNS-1 Cycle 4 and the P
generic conclusions support a power-dependent MCPR limit as sho~n in l Figure 3.6 of Reference 2. This limit was considered in evaluating the
! power-dependent Cycle 4 MCPR limits documented in Reference 2.
l l 9.3 Reduced Flow and Power Operation The current of f-rated thermal limits (MCPR , MCPR g
,pMAPFACf and MAPFACp )
were first established by GE in support of Cycle 1 ME00 operation (Reference 6). Because the remaining GE bundles will be discharged from the core during the third refueling outage the thermal limits are revised ;
based on ANF methodology for the full ANF core for Cycle 4 operations, f The flow dependent limits were constructed based on Cycle 2 and 3 ANF analyses for GGNS (Reference 12). Additional conservatism was added to bound possible variations in future cycles. ANF validated the applicability of these limits for Cycle 4 (Reference 2).
SUMMARY
- 16
The flow dependent limits are applied for the Loop Manual and Non Loop Manual modes of operation. A conservative maximum core flow of 110% of rated is used, which bounds the achievable maximum core flow for the limiting flow runout event. Conservative flow rate increases are assumed t for the Loop Manual mode which are based on the one loop flow runout event. This is the only credible flow runout event in the Loop Manual mode of operation because each flow control valve is operated independently. A complete description of the flow dependent thermal limits and the runout flow rates is provided in Reference 12.
l The MCPR f limits and MAPFAC f values were determined for Cycle 4. The Cycle 4 MCPR f limit and ANF validation results are shown in Figure 2.3 of i
Reference 2. The Cycle 4 MAPFAC f values and ANF validation results are shown in Figure 2.4 of Reference 2.
i l Single Loop Operation (SLO) is addressed in Sections 1.0, 5.2, 7.1.1, i
and 7.2.1 of the Cycle 4 Reload Report (Reference 1). The revised single j loop MAPLHGR curve (Figure 1.2, Reference 1) was conservatively I l constructed to bound all fuel types for Cycle 4 operation. The Cycle 3 SLO MAPFAC f curve is applicable to Cycle 4.
4 I
The power-dependent MCPR operating limit for Cycle 4 was determined based l l
! on the generic CRWE analysis and the plant-specific analyses of LRNB and I
, FWCF transients at representative conditions blanketing the operating !
, i power flow map. The MCPR p
limit for Cycle 4 consists of the ANF generic j CRWE curve above 40% of rated power. The current Technical Specification l
l i
SUMMARY
- 17 [
I
values at and below 40% of rated power are supported by ANF CycIn 4 analysis. The power-dependent MCPR limit is given in Figure 3.6 of Reference 2.
The MAPFAC p operating limit factor was validated for Cycle 4 through the analysis of the FWCF, LRNB, and CRWE transients. The transient analyses which were used to evaluate the MCPR p limit were also used to evaluate the MAPFAC p limit. The FWCF, LRNB and CRWE are compared with the MAPFACp limit in Figure 2.2 of Reference 2. The MAPFACp Cyc1c 3 limit was found applicable for Cycle 4.
9.4 ASME Overpressurization Analysis In order to demonstrate como11ance with the ASME Code over pressurization criterion of 110% of vessel design pressure, the MSIV closure event with failure of the MSIV position switch scram was analyzed with ANF's COTRANSA code (Reference 11). The Cycle 4 analysis assumes seven safety / relief valves are out of service. Furthermore, for analysis purposts only, the setpoint tolerances for the safety valves were increased to 6% (an increase of approximately 4% over the Cycle 3 analysis). The results show that the safety valves have sufficient capacity with the increased setpoint tolerances to protect the sessel pressure safety limit of 1375 psig during Cycle 4 (Table 2.1, Reference 2). An ANF evaluation shows that the GE analysic of ATW3 ov e pressuetzation is applicable to ANF fuel and therefore rec,ains valia for Cycle 4.
SUMMARY
- 18 l
t The increase in the analysis satpoint tolerance to 6% is in anticipation j of a Technical Specification change in the safety valve setpoint tolerance which may be requested subsequent to the third refueling outage. At tnis time, no Technical Specification change is being made to the setpoint tolerance.
1 I 10.0 POSTULATED ACCIDENTS In support of Grand Gulf operation, ANF has analy;:ed the Loss-of-Coolant s Accident (LOCA) to demonstrate that MAPLHGR limits for Reload 3 fuel I
i comply with 10CFR50.46 criteria. Methodology for the LOCA analysis is !
provided in References 13 through 15. The Rod Drop Accident (RDA) was i analyzed for ANF Reload 3 'uel to demonstrate compliance with the 280 1
l cal /p Design Limit. Methodology for the RDA analysis is described in I
XN-NF-80-39(A) Volume 1 (Reference.9).
i i i
f 10.1 Loss-of-Coobnt Accident (LOCA) f i
)
l The generit. BWR/6 LOCA break spectrum analysis as described in i
Reference 16 and perfore 1 in support of the Cy',le 2 submittal remains l
) applicable for Cycle 4. A confirmatory heatup analysis for operation
! within the Cycle 4 MAPLHGR limits was performed. The enalysis confirms -
{
i that the Peak Cladding Temperaturc (PCT) remains well below the 10CFR50.46 l PCT limit of 2200 degrees F for all fuel types expected to be in the core i
- during Cycle 4. Similarly, confirmatory analyses were performed and show i
l i
SUMMARY
- 19 .
l
- that local Zr-H O reaction remains below 17% and corewide hydrogen 2
- p*oduction reasins below 1% for the limiting LOCA event. The results of j these analyses are presented in Section 6.1 of Reference 1.
i' 10.2 Rod Drop Accident i
1 ANF's methodology for analyzing the Rod Drop Accident (RDA) utilizes a generic parametric snelysis which calculates the fuel enthalpy ri u during the postulated RDA over a wide range of reactor operating
! conditions. For Cycle 4. Section 6.2 of Reference 1 shows a value of 3 172 cal /gm for the maximum deposited fuel rod enthalpy during the worst i case postulated RDA. This value is well below the design limit of 280 cal /gm. To ensure complience with the RDA analysis assumption, control l rod sequencing below 20% core thermal power complies with GE's Banked Position Withdrawal Sequencing constraints (Reference 17).
I 11.0 Refuelino Operations I
i During refueling for the GGNS Cycle 4 reload plant operations will be l
! testricted to modes 4, 5 and *. Mode
- is defined in Section 3/4.6 of the GGNS Technical Specifications as, "when irradiated fuel is being handled f in the primary or secondary containment and during CORE ALTERATIONS and 1
j operations with a potential for draining the reactor vessel."
i l
l i
SUP9tARY - 20 4
[ . ._
s The :*1. applicable FSAR accidents for refueling operations tre listed below:
l a) Section 15.4.1, Rod Withdrawal Error - low power ,
L) Section 15.4.3, Control Rod Maloperation (system malfunction or operator error) c) Section 15.4.7, Misplaced Bundle Accident
, d) Section 15.7.4, fuel Handling Accident - Auxiliary building e) Section 15.7.6, Fuel Handling Accident - In containment.
Other FSAR accidents are not considered since the reactor must be maintained in a subcritical condition during modes 4, 5 and *.
Consequently, the accidents are not possible and/or refueling activities have no effect on the cause of the accidents.
Inadvertent criticality either from a Rod Withdrawal Error oi Control 4
Rod Maloperation is precludad by plant design and administrative controls when operations are restricted to modes 4, 5 and *. The Hisplaced Bundle and Fuel Handling Accidents were evaluated by ANF.
12.0 REFERENCES
j 1) ANF-88-149, "Grand Gulf Unit 1 Cycle 4 Reload Analysis", Advanced a Nuclear c ueis Corpori ...,, October 1938.
i
SUMMARY
- 21
- 2) ANF-88-150, "Grand Gulf linit 1 Cycle 4 Plant Transient Analysis",
Advanced Nuclear Fuels Corporation, October 1988.
- 3) Letter from J. G. Cesare, System Energy Resources, Inc., to USNRC, AECM-88/0206, October 1988.
- 4) XN-NF-80-19(A), Vol. 4, Revision 1, "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads", Exxon Nuclear Co., June 1965.
- 5) XN-NF-85-67(A), Rev. 1, "Generic Mechanical Design for Exxon Nuclear e Jet Pump BWR Reload Fuel", Exxon Nuclear Co., September 1986.
- 6) "GGNS Maximum Extended Operating Domain Analysis", General Electric
. Company, San Jose, California March 1986.
- 7) XN-NF-81-51(A), "LOCA-Seismic Structural Response of an ENC BWR Jet Pump Fuel Assembly", Exxon Nuclear Co., May 1986.
- 8) XN-NF-524(A), Rev. 1, "Exxon Nuclear Critical Power Methodology for Boiling Water Reactors", Exxon Nuclear Co., November 1983.
- 9) XN-NF-80-19(A), Vol.1 Supplements 1 & 2, "Exxon Nuclear Met hodology for Boilirg Water Reactors: Neutronic Methods for Desi0n and Analysif , Exxon Nuclear Co., March 1983.
SUMMARY
- 22
- 10) ANF-88-183(P), "GGNS Unit 1 Reload XN-1,3 Cycle 4 Mechanical Design" ANF Corporation, November 1988.
'l
- 11) X'i-NF-79-71(P), Rev. 2, "Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors", Exxcn Nuclear Co., November 1981.
- 12) NESDQ-88-003, Revision 0, "Revised Flow Dependent Thermal Limits",
MSU System Services Inc., November 1988.
- 13) XN-NF-80-19(A), Vols. 2, 2A, 2B, & 20, "Exxon Nuclear Methodology for Boiling Water Reactors: EXEM BWR ECCS Evaluation Model", Exxon Nuclear Co., September 1982.
- 14) XN-NF-CC-33(A), Rev. 1, "HUXY: A Generalized Multirod Heatup Code with 10CFR50 Appendix K Heatup Opt , Exxon Nuclear Co., November 1975.
- 15) XN-NF-82-07(A), Rev. 1, "Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model", Exxon Nuclear Co., November 1982.
- 16) XN-NF-86-37(P), "Generic LOCA Break Spectrum Analysis for BWR/6 i
Plants" Exxon Nuclear Co., May 1986, i
- 17) NED0-21231, "Banked Position Withdrawal Sequence", General Electric Co. , January 1977.
SUMMARY
- 23
- 18) XN-NF-825(A), Supplement 2, "BWR/6 Generic Rod Withdrawal Error Analysis, MCPR p for All Plant Operations Within the Extended Operating Domain", Exxon Nuclear, January 1986.
- 19) XN-NF-80-19(A), Volume 3, Revision 2, "Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Thermal Limits Methodology",
Exxon Wuclear Co., January 1987.
- 20) XN-NF-84-105(P)(A), Volume 1, "XCOBRA-T: A Computer Coda for BWR Transient Thermal Hydraulic Core Analysis", Exxon Nuclear Company, Inc., February 1987.
J i
l l
l l
l
SUMMARY
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g APPENDIX A 9X9-5 LEAD TEST ASSEMBLIES A.1 INTRODUCTION Evaluations have been performed to establish a licensing basis for the four (4) ANF 9x9-5 Lead Test Assemblies (LTAs) in the Grand Gulf Cycle 4 core. Justification is provided which demonstrates the applicability of Grand Gulf Cycle 4 operating limits to the LTAs.
The insertion of four ANF 9x9-5 LTAs will have negligible effects upon core-wide transient performance. However, specific analyses were performed for the 9x9-5 LTAs ts assure that the Cycle 4 operating limits also apply to the LTAs. The analysis methodology was consistent with that used in analyzing the 8x8 assemblies and is described in ANF Topical Report XN-NF-80-19(A). Fuel type specific limits (LHGR and related MAPLHGR limit) were developed for the LTAs. The LTAs will be placed in non-limiting locations in the core. This placement, in conjunction with their improved thermal performance, will increase the margin to operating limits l
for the LTAs.
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A.2 FUEL MECHANICAL DESIGN A mechanical design analysis was performed for the 9x9-5 fuel type o
consistent with ANF's approved methodologies (Reference A.1).
Fuel design issues related to Anticipated Operational Occurrences (A00's) and accident analysis have been evaluated. These evaluatior confirm that the LTAs meet NRC critera of no centerline melting and less than 1% clad strain. .
A.3 THERMAL HYDRAULIC DESIGN Component hydraulic resistances have been determined and it has been found that the 9x9-5 LTAs are hydraulically compatible with the co-resident ANF 8xS fuel assemblies. Unique design features of the 9x9-5 (two rod diamete,rs, injection water rod) have been ,
modeled to demonstrate compatibility over the full range of expected operating conditions. Steady state thermal hydraulic analyses have shown that even though the 9x9-5 design has a somewhat smaller flow area than the 8x8 design no reduction in thermal margin is experienced in the mixed core configuration because of the increased thermal performance of the 9x9-5 in conjunction with the low power region in which they will be located.
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A.4 NUCLEAP DlSIGN i
lhe core wide neutronic impact of replacing four of the 800 fuel assemblies in the Grand Gulf Cycle 4 core is negligible. The LTAs are designed to be neutronically "transparent" relative to the 8x8 fuel; that is, reactivity characteristics are similar.
The ANF 9x9-5 LTA contains 76 fueled rods and five non-fueled rods, one of which is a spacer capture rod. The fuel assembly consists of a 138 inch central zone enriched to 3.47 w/o U-235 and six inch top and bottom natural (0.711 w/o U-235) uranium blanket.
The fuel assembly average enrichment is 3.25 w/o U-235. Five different fuel rod enrichments are utilized to yield a flat local power distribution. Each of the assemblies utilizes axial gadolinia in eight rods containing both 5.0 and 6.0 w/o Gd 023 to reduce the initial reactivity of the assembly.
The four (4) 9x9-5 LTAs were modeled explicitly in the Less of Feedwater Heating, Control Rod Drop Accident, MAPFACf , shutdown margin and Shutdown Liquid Boron Control analyses. The LTA Hisload 3 has been evaluated separately using the XN-3 correlation, which has been demonstrated to conservatively predict critical power in the 9x9-5 design (Reference A.2).
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A. 5 ANTICIPATED OPERATIONAL OCCURRENCES Analyses of limiting transients have shown that the bundle power needed to produce boiling t ansitions during transients in the 9x9-5 fuel design is higher than that for the 8x8 fuel design. It has been shown that ANF's approved BWR CHF correlation, XN-3, is conservative when applied to the 9x9-5 CHF data. Therefore, applying 8x8 MCPR operating limits based on XN-3 to the current 9x9-5 LTAs assbres lower bundle powers than would be necessary to reach the 9x9-5 boiling transition.
Because of the neutronic similarity of the 9x9-5 LTAs to the 8x8 assemblies, the consequences of occurrences such as control rod withdrawal error and fuel rotation errce are essentially the same as in the case of 8x8 fuel.
A.6 DOSTULATED ACCIDENTS i
I Because h$atup is primarily a planar and not an axial phenomenon, the appropriate bundle power limit derived from LOCA analyses is the peak bundle planar power. It has been demonstrated that the 9x9-5 LTAs provide better LOCA performance relative to an 8x8 fuel l
assembly due to the greater surface area provided by the larger 1
number of fur.1 rods, more inert surface from the water rods and l
less stored energy in the rods. The 9x9-5 LTA MAPLHGR limit is based on the LHGR limit provided in "Generic Mechanical Design for
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ANF 9x9-5 BWR Reload Fuci" (Reference A.1) divided by the maximum local peaking as a function of exposure. Analyses performed by ANF demonstrate that this limit meets 10CFR50.46 criteria.
The consequences of a control rod drop accident are governed primarily by the dropped rod worth. Since the reactivity of the LTAs is comparable to the co-resident 8x8 fuel'and the LTAs are loaded into low reactivity regions, no appreciable difference will be experienced due to the LTAs.
Comparison of the radiological consequences of fuel handling accidents with 8x8 and 9x9 fuel showed less radioactivity released ,
for the 9x9 fuel.
A.7 T.ECHNICAL SPECIFICATIONS All operational limits used for 8x8 fuel are applicable to the 9x9-5 LTAs except for fuel type specific MAPLHGR limits and a 9x9-5 LHGR limit. The LHGR limit is shown in Figure A-1 of Reference A.2. The MAPLHGs is shown in Figure *A.2 of ANF-88-149 and is the smaller j
of the equivalent 8x8 MAPLHGR limit and the MAPLHGR limit consistent with the 9x9 LHGR limit. The off-rated MAPLHGR during single loop operation is the product of the LTA MAPLHGR and the smallest of the corresponding MAPFAC f , MAPFACp or 0.86.
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A.8 CRITICALITY OF 9x9-5 FUEL A GGNS-1 specific High Density Spent Fuel Storage Rack (HDSFSR) criticality safety analysis was performed and is being submitted under separate cover. This analysis shows that with the introduction of the ANF 9x9 LTAs into the HDSFSR, the infinite multiplication factor of the HDSFSR remains at or below 0.920, This is below the NRC acceptance criterit of K,ff=0.95.
A.9 REFERENCES A.1) ANF-88-152, "Grand Gulf Unit 1 XN-1.3 Cycle 4 Design Report Mechanical, Thermal-Hydraulic and Neutro.11cs Design for Advanced Nuclear Tests 9x9-S Leads", September 1988.
A.2) ANF-88-149, "Grand Gulf Unit 1 Cycle 4 Reload Analysis",
Advanced Nuclear Fuels Corporation October 1988.
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- APPENDIX B S
ENVIRONMENTAL EFFECTS OF TRANSPORT OF NUCLEAR FUELS The Grand Gulf Nuclear Station Unit 1 Cycle 4 reload fuel was designed to achieve a batch average discharge exposure of r approximately 34 GWd/MTU. This is slightly above the assumption (33 GWd/MTU) used in assessing the environmental effects of transportation of fuel and wastc as set forth in Table S-4 of 10 CFR 51.52(c). The environmental impacts of transportation resulting from the use of extended irradiation and higher y
enrichment fuel are discussed in the NRC staff assessment
[ entitled, "NRC Assessment of the Environmental Effects of s
Transportation Resulting from Extended Fuel Enrichment and
$ Irradiation," dated July 7,1988.
The assumptions related to transportation used in that assessment and the supporting analyses were reviewed ar.d found applicable and/or bounding for the Cycle 4 reload. Therefore the conclusion of that assessment, that the environmental impacts of transportation resulting froin the use of extended ,
irradiation and enrichment remain unchanged or may in fact be reduced, is valid for the GGNS-1 Cycle 4 reload.
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