AECM-90-0092, Cycle 5 Reload Summary Rept
| ML20043F040 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 05/31/1990 |
| From: | ENTERGY OPERATIONS, INC. |
| To: | |
| Shared Package | |
| ML20043F017 | List: |
| References | |
| AECM-90-0092, AECM-90-92, NUDOCS 9006140114 | |
| Download: ML20043F040 (23) | |
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AECM 90/0092 GRAND GULF NUCLEAR STATION UNIT 1 i
CYCLE 5 RELOAD
SUMMARY
REPORT 9
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May 1990 9006140114 900608 ADOCK 0500(g6 DR
CONTENTS EAM
1.0 INTRODUCTION
1 2.0 CYCLE 5 RELOAD SC0PE........................................
2 3.0 CYCLE 4 OPERATING HIST 0RY...................................
3 4.0 CYCLE 5 CORE DESCRIPTION....................................
3 5.0 FVEL MECHANICAL DESIGN......................................
4 6.0 THERMAL HYDRAVLIC DESIGN...................................
5 6.1 Safety limit MCPR...............
6
.6.2 Exposure-dependent MCPR...............................
6 6.3 Core Stability........................................
7 7.0 NUCLEAR DESIGN..............................................
8 7.1 Fuel Bundl e Nucl ear De s ign.............................
8 7.2 Core Reactivity.......................................
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7.3 Spent Fuel Pool Critical i ty...........................
9 L
8.0 CORE MONITORING SYSTEM......................................
10 9.0 ANTICIPATED OPERATIONAL OCCURRENCES.........................
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9.1 Core-Wide Transients..................................
11 9.2 Local Transients......................................
12 9.3 Reduced Flow and Power 0perat'on......................
12 9.4 ASME Overpressurization Analysi s......................
13 10.0 POSTULATED ACCIDENTS........................................
13 i
10.1 Lo s s-o f Cool ant Accident..............................
14 10.2 Rod Drop Accident.....................................
15 11.0 REFUELING 0PERATIONS........................................
15
12.0 REFERENCES
16
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1.0 INTRODUCTION
This report is a supplementary document that summarizes the results of the analyses performed in support of GGNS Unit 1 Cycle 5 operation. The fresh fuel to be inserted in this cycle is an ANF 9x9 5 fuel type.
It is similar to the four 9x9 5 Lead Test Assemblies (LTAs) inserted for Cycle 4 except for differences in enrichment, gadolinia loadings, and a simplified water rod design.
This fuel has been shown to be compatible with the 8x8 fuel types that will be resident in the core during Cycle 5 (Reference 1).
- -, Jew.metnodologies are introduced to support the reload analyses.
.These include the MCpR Safe',y Limit (Reference 8), the ANTB critical porter correlation (Reference 27), the CASM0 3G/ HICROBL!PN B neutronics code (Reference 24), end the revised thermal-hydraulic code COTRANSA2 (Reference 26).
The ANF Cycle 5 Reload Analysis Report (Reference 1) and the Cycle 5 Plant Transient Analysis Report (Reference 2) serve as the basic framework for the reload analyses. Where appropriate, reference is made to these and other supporting documents for more detailed information and/or specifics of the applicable analyses. A list of references comprising both the generic and the GGNS specific documuits used in support of the Cycle 5 reload submittal is provideo in Section 12.0 of this report.
i.
2.0 CYCLE 5 RELOAD SCOPE During the fourth refueling outage at GGNS Unit 1, depleted ANF 8x8 fuel assemblies will be replaced by ANF 9x9 5 fuel assemblies.
Fuel related analyses of the limiting events were performed in support of Cycle 5.
This included analyzing Cycle 5 for anticipated transients, the Fuel Misload Error Event, confirmatory analyses for LOCA, and the 1
Control Rod Drop Accident.
These analyses were performed to support the safety and operating limits based on ANF methodology for two loop and single loop operations. Analyses for normal operation of the reactor consisted of fuel evaluations in the areas of mechanical,
-,--thermal hydraulic and nuclear design.
r.
Based on AiF's design and safety analyses of the Cycle 5 reload core,-
the proposed changes to the GGNS Unit 1 Technical Specifications are as follows:
l a.
MCPR Safety limits are revised.
b.
MAPLHGR curves for the.8x8 and 9x9-5 fuel types are revised.
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c.
Thermal limits for off-rated conditions are determined based on LHGR multipliers instead of MAPLHGR multipliers.
d.
The LHGR limit curve for 8x8 fuel type is extended to higher exposure.
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1 e.
Flow-and power-dependent MCPR limits are revised.
Exposure-dependent MCPR limits are introduced.
f.
The Rod Action Control System (RACS) minimum bypass power is revised.
g.
The Power / Flow Map is revised.
i 3.0 CYCLE 4 OPERATING HISTORY Cycle 4 core-follow operating data available at the time of the
... -reload.. design analysis, together with projected plant operation through the end of Cycle 4, was used as a basis.for the Cycle 5 core design and as input to the plant safety analyses.
Cycle 4 has continued to operat,e as expected.
No operating anomalies have occurred that would affect the licensing basis for Cycle 5.
Cycle 5 analyses were performed assuming a Cycle 4 energy range of 1698 GWd to 1740 GWd.
'4.0 CYCLE 5 CORE DESCRIPTION The Cycle 5 core will consist of 800 fuel assemblies comprising 284 fresh ANF 9x9-5 assemblies (fourth reload), 272 once burned ANF 8X8 assemblies and 4 9x9-5 LTAs (third reload), and 240 twice burned ANF 8x8 assemblies (second reload). A breakdown by bundle type / bundle average enrichment is provided in the following table: -
6 Number of Bundles Bundle Type 284 ANF 9x9/3.42 w/o U235 4
ANF 9x9/3.25 w/o U235 272 ANF 8x8/3.37 w/o U235 t
240 ANF 8x8/3.01 w/o U235 i
The anticipated Cycle 5 core configuration, together with additional bundle and core design details, is provided in Section 4.0 of the ANF Cycle 5 Reload Analysis Report (Reference 1). The Cycle 5 core is a conventional scatter load with the lowest reactivity bundles placed in the peripheral region of the core.
The loading pattern was designed to maximize cycle energy and minimize power peaking factors.
Cycle 5 is estimated tc provide 1698 GWd nf energy based on a Cycle 4 energy nutput of 1740 GWd.
5.0 FUEL MECHANICAL DESIGN i
The mechanical design analyses for the ANF 8x8 and 9x9-5 fuel types are described in References 4, 5, and 10.
The 8x8 fuel assembly design contains 62 fuel rods and two water rods, one of which functions as a spacer capture rod.
Seven spacers maintain fuel rod i-spacing. The 9x9-5 fuel assembly design contains 76 fuel rods and five water rods, one of which serves as a spacer capture rod.
Seven spacer > maintain fuel rod spacing.
The fuel rods are prepressurized, and use a diametral pellet-to-clad gap that is smaller on the interior high enrichment rods in order to improve ECCS performance.
l Mechanical design analyses were performed to evaluate cladding steady-state strain, transient stresses, fatigue damage, creep collapse, corrosion buildup, hydrogen absorption, fuel rod maximum 4
1 internal pressure, differential fuel rod growth, creep bow, and grid spacer spring design.
These analyses were performed to support peak assembly discharge burnups of 39 GWd/MTU and 40 GWd/MTU for the 8x8 1
and 9x9-5 fuel types, respectively. As shown in References 4 and 5, all parameters meet their respective design limits; no fuel centerline melting will occur at 120% and 135% overpower conditions for 8x8 fuel and 9x9 5 fuel types, respectively. The Cycle 5 core design is bounded by the assumptions used in these analyses.
Fuel channels manufactured by Carpenter Technology Corporation
+
~-(CarTech).will be introduced in Cycle 5 for the reload 9x9 5 fuel.
y, These channels are= equivalent to the GE channels used in previous cycles.
The mechanical responses of the 8x8 and 9x9 5 ANF assembly designs during seismic-LOCA events are essentially the same because the physical properties and bundle natural frequencies are similar.
Reference 7 presents the seismic-LOCA analysis for the 8x8 fuel and shows that the resultant loadings do not exceed the fuel design limits.
Reference 23 presents the corresponding seismic-LOCA analysis for 9x9 5 fuel. The applicability of these analyses to the 8x8 and 9x9 5 fuel assemblies in the Grand Gulf Unit I core has been confirmed by ANF (Reference 1).
l 6.0 THERMAL HYDRAVLIC DESIGN XN-NF-80-19(A), Volume 4, Revision 1 (Reference 3) discusses the L i
thermal-hydraulic design criteria that are used in the determination of the fuel cladding integrity safety limit and the bypass flow characteristics. ANF analyses were performed in accordance with XN NF-80 19(A), Volume 3, Revision 2 (Reference 19) to determine the parameters that demonstrate compliance with these design criteria. -
6.1 Safety Limit MCPR The MCPR fuel cladding integrity safety limit is 1.08 for both Two Loop Operation and Single Loop Operation (SLO).
The methodology and generic uncertainties used in the Cycle 5 MCPR safety. limit calculation, including the effects of channel bow, are.provided in Reference 8.
6.2 Exoosure-dependent MCPR Exposure Dependent MCPR limits (MCPR,) are introduced to define Operating Limit MCPRs as a function of core exposure for
. Cycle 5.
Revised MCPR operating limits are defined,for Cycle 5 compared to Cycle 4.
Progressively higher MCPR limits are defined for the last 2 GWd/MTV of the cycle and for the exposure window up to the Maximum Licensing Exposure.
l Analyses of the most limiting core wide transients and local events were performed to confirm the acceptability of the MCPR, limits for use in Cycle 5.
These limits were established consistent with the Cycle 5 operating strategy.
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6.3 Core Stabilitv The GGNS Unit 1 Technical Specifications implement the BWROG/GE Inter'.m Recommendations for Stability Actions (IRSA). The IRSA boundaries, which were developed based on GE fuel experience, have been approved for application at GGNS Unit I containing ANF 8x8 fuel (Amendment No. 62 to Facility Operating License-No. NPF 29 Reference 25).
The Cycle 5 core will contain the first ANF 9x9-5 reload batch.
The 9x9-5 fuel has been shown to be thermal hydraulically and neutronically compatible with the ANF 8x8 fuel (Reference 1).
ANF performed confirmatory analyses for Cycle 5 for core stability calculations.
These analyses consisted of a comparative evaluation of the Cycle 4 and Cycle 5 stability characteristics at the same statepoints using nominal power
. distributions.
The analyses results showed that the core decay ratios for the two cycles are equivalent, with the decay ratios for Cycle 4 being slightly higher than those for Cycle 5; the difference in stability performance was shown to be comparable to the variations observed for previous cycles.
1 In summary, the GE/BWROG recommendations (IRSA) on operating domain boundaries and operator actions have been shown to be applicable for a wide range of 8x8 fuel and core design configurations.
The ANF 8x8 and 9x9 5 fuel types have been 7
shown to be compatible in the Grand Gulf core. The results of ANF's analyses show that the differences in decay ratio between Cycles 4 and 5 are comparable to the variations observed relative to previous GGNS-1 cycles. Therefore, the current GGNS-1 stability-related technical specifications are applicable for Cycle 5 operation.
l 7.0. NUCLEAR DESIGN The neutronic methods used for the design and analysis of ANF reloads are described in ANF topical reports (References 9 and 24).
7.1 Fuel Bundle Nuclear Desion The Cycle 5 reload fuel utilizes ANF 9x9 5 fuel assemblies.
Two basic bundle designs are used with different axially distributed burnable poison concentrations.
For both designs, the top 12 inches and bottom 6-inches of each fuel rod 'contain natural uranium and the central 132 inch zone of each rod contains enriched uranium at one of six different enrichments.
The average enrichment of the bundle enriched zone is 3.80 weight percent (w/o) U235 and the bundle average enrichment is 3.42 w/o U235.
The neutronic design parameters and rod enrichment distribution are described in Section 4.0 of the Cycle 5 Reload Analysis Report (Reference 1).
7.2 Core Reactivity with the The beginning of Cycle 5 (B005) cold core Keff
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strongest worth control rod fully withdrawn at cold, 68 degrees F reactor conditions was calculated to be 0.98956. This corresponds to a shutdown margin of 1.06% delta k/k.
B005 was j
determined to be the most limiting condition.
Therefore, the
)
difference between the minimum shutdown margin in the cycle and j
the B0C shutdown margin, R, is 0.00% delta k/k. The calculated shutdown margin is well in excess of the 0.38% delta k/k Technical Specification requirement (Section 3/4.1.1), and will be verified by testing at BOC5 to be greater than or equal to R
+ 0.38% delta k/k.
The Standby Liquid Control (SLC) system is designed to inject a -
- quantity of boron that produces a concentration of no less than 660 ppm in the reactor core. Analyses were performed to show that the minimum shutdown margin is at least 3.0% delta k/k with the reactor in a cold, xenon free state, at the most limiting cycle exposure, and with all control rods in their critical full power positions.
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 confirms the basis of the Technical Specification requirement for the Cycle 5 reload core.
7.3 Soent Fuel Pool Criticality A GGNS-1 specific High Density Spent Fuel Storage Rack (HDSFSR) criticality safety analysis was performed and submitted
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previously (Reference 21). This analysis shows that with the introduction of the higher enriched Cycle 5 fuel into the HDSFSR, the infinite multiplication factor of the HDSFSR remains at or below 0.946. This is below the NRC acceptance criteria of K,77=0.95.
i 8.0 CORE MONITORING SYSTEM The POWERFLEX-core monitoring system is and will continue to be utilized to monitor reactor parameters at GGNS. The core monitoring system is fully consistent with ANF's nuclear analysis methodology as j
.descr.ibed,in= References 9 and 24.
In addition, the measured power
~ distribution uncertainties. are. incorporated into the calculation of...
1 the MCPR Safety Limit-as described in ANF's Nuclear Critical Power Methodology Report (Reference 8).
9.0 ANTICIPATED OPERATIONAL OCCURRENCES In order to support the Cycle 5 operating ilmits, eight categories of system transients are considered as described in ANF's Plant j
1 Transient Methodology Report (Reference 11). ANF has provided plant
)
specific analysis results for the following system transients to determine the thermal margin requirements for operation during Cycle 5 (Reference 2):
- 1) Generator load Rejection without Bypass (LRNB)
- 3) Loss of Feedwater Heating (LFWH)
- 4) Flow Excursion I
Analyses performed for previous cycles have shown that the other system transients are non limiting and, therefore, are bounded by one of the above.
In addition, the fuel Loading Error was analyzed in accordance with the methodology described in Reference 9.
The Control Rod Withdrawal Error (CRWE) transient has been analyzed generically in Reference 18.
In addition, CRWE analyses specific to~
Cycle 5 have been performed (Reference 1).
Single Loop Operation is addressed in Appendix A of the Cycle 5 Transient Analysis Report (Reference 2).
9.1 Core-Wide Transients The plant transient codes that were used to evaluate the LRNB and FWCF events are ANF's COTRANSA2.(Reference 26) and XCOBRA T i
(Reference 20), which incorporate a one dimensional neutronics model to account for shifts in axial power shape and control rod effectiveness. Technical Specification scram times (Section 3/4.1.3) were used in the bounding analysis.
The results of the LRNB and FWCF analyses'are provided in the Cycle 5 Plant Transient Analysis Report (Reference 2) and a summary of results is provided in the Cycle 5 Reload Analysis Report (Reference 1). The LFWH event was analyzed consistent with the ME00 power / flow operating map for various cycle exposures anticipated during Cycle 5.
A summary of this analysis is provided in Reference 2.
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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 4
of the CRWE transient for BWR/6 plant configurations under conditions which cover the normal operating power / flow map, the extended load line region, and the increased core flow region.
This analysis was reevaluated using the ANFB Critical Power j
Correlation (Reference 27) and the MICR0 BURN-B neutronics code (Reference 24). Additionally, GGNS-1 Cycle 5 statepoints were l
- also analyzed. The results of these analyses were used to confirm.the power-dependent Cycle 5 MCPR limits documented in Reference 2.
9.3 Reduced Flow and Power Ooeration o
and The off-rated thermal limits (MCPR, MCPR, MAPFACf 7
p MAPFAC ) were first established by GE in support of Cycle 1 p
MEOD operation (Reference 6).
These limits were confirmed or revised, as appropriate, for subsequent cycles.
For Cycle 5, and MAPFAC ) are replaced the MAPLHGR multipliers (MAPFACf p
and LHGRFAC ).
The use of' by LHGR multipliers (LHGRFAC7 p
the LHGR limits and multipliers at off-rated conditions is equivalent to the use of the MAPLHGR limits and multipliers in-previous cycles and ensures that the fuel mechanical design criteria are satisfied; the MAPLHGR limits ensure that the 2200 degrees F PCT limit is not challenged.
4
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Revised flow dependent limits were established for Cycle 4 to provide for Non Loop Manual and Loop Manual modes of operation (Reference 12). The flow dependent MCPR limits were revised for Cycle 5 consistent with new ANF methodology.
l The revised power dependent MCPR operating limits for Cycle 5 were determined based on the system transient analyses at representative conditions within the operating domain.
.i
- 9.4, ASME Overoressurization Analvd1 in order _to demonstrate compliance with the.ASME Code overpressurization criterion of 110% of vessel design pressure, the MSly closure event with failure-of the MSIV position switch i
scram was analyzed using ANF's COTRANSA2 code (Reference 26).
The Cycle 5 analysis assumes seven safety / relief valves are out
'of service. As was done for the Cycle 4 analyses, the setpoint-1 l
tolerances for the safety valves were assumed conservatively to j
-be 6%.
The results show that the safety valves have sufficient capacity to protect the vessel pressure safety limit of 1375 psig during Cycle 5 (Reference 2).
10.0 POSTULATED ACCIDENTS In support of Grand Gulf operation, ANF has analyzed the Loss of-1 Coolant Accident (LOCA) to demonstrate th t MAPLHGR limits for j
S Cycle 5 reload fuel comply with 10CFR50.46 criteria.
Methodology for..
)
i the LOCA analysis is provided in References 13 through 15.
The Rod Drop Accident (RDA) was analyzed for the Cycle 5 core to demonstrate compliance with the 280 cal /gm Design Limit. Methodology for the RDA analysis is described in XN NF-8019(A), Volume 1 (Reference 9). An i
ANF evaluation shows that the GE analysis of ATWS overpressurization is applicable to ANF fuel and therefore-remains valid for Cycle 5.
10.1 Loss obCoolant Accident (LOCA) t The generic BWR/6 1.0CA break spectrum analysis as described in Reference 16 and performed in support of the Cycle 2 submittal remains applicable for Cycle 5.
A heatilp analysis was
. performed for the. reload 9x9 5 fuel.
The analysis. confirms that the Peak Cladding Temperature (PCT) remains well below the 1
10CFR50.46 PCT limit of 2200 degrees F.
Revised MAPLHGR curves t
for 8x8-and 9x9 5 fuel types, with an appropriately revised j
single loop operation (SLO) multiplier (Reference 1), were l
conservatively constructed to bound both Two Loop Operation and-l-
SLO for Cycle 5.
L Confirmatory analyses were performed to show that the local Zr-H O reaction remains below 17% and that the core-wide 2
metal-water reaction (CMWR) remains below 1% for the limiting LOCA event as required by 10CFR50.46.
The results of these analyses are presented in Section 6.1 of Reference 1.
As stated in the GGNS-1 UFSAR, the hydrogen recombiners have been sized to process the hydrogen released from 0.8% CMWR.
1 l
Consistent with the Regulatory Guide 1.7 requirements for post LOCA combustible gas control, the hydrogen released from CHWR for Cycle 5 has been calculated and shown to be within the i
hydrogen recombiner design basis.
P 10.2 Rod Droo Accident ANF's methodology for analyzing the Rod Drop Accident (RDA) utilizes a generic parametric analysis that calculates the feel enthalpy rise during the postulated RDA over a wide range of i
reactor operating conditions.
For Cycle 5. Section C.2 of Reference I shows a value of 192 cal /gm for the maximum deposited fuel rod enthalpy during the worst case postulated RDA.
This value is well below the desigr. limit of 280 cal /gm, The RDA analysis assumption complies with GE's Banked Position Withdrawal Sequencing constraints (Reference 17).
Based on BWROG methodology, the CRDA has been shown to be inherently
.self limiting for core powers above 10% due to the presence of voids in the core (Reference 22).
l' 11.0 REFUELING OPERATIONS As was done for Cycle 4, refueling operations will be addressed by a 1
10CFR50.59 Safety Evaluation. l l
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12.0 REFERENCES
1)
ANF-90 022, Revision 1, ' Grand Gulf Unit 1 Cycle 5 Reload Analysis,"
Advanced Nuclear Fuels Corporation, May 1990.
2)
ANF-90-021, Revision 1, " Grand Gulf Unit 1 Cycle 5 Plant Transient Analysis," Advanced Nuclear Fuels Corporation, May 1990.
3)
XN NF 8019(P)(A), Volume 4, Revision 1, " Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads,"
Exxon Nuclear Co., June 1986.
4)
XN NF 85-67(P)(A), Revision 1, " Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel," Exxon Nuclear Co., September 1986.
5)
ANF 88-152(P), Amendment 1, " Generic Mechanical Design for Advanced Nuclear Fuels 9x9 5 BWR Reload Fuel," Advanced Nuclear Fuels Corporation, September 1969.
6)
"GGNS Maximum Extended Operating Domain Analysis," General Electric Company, March 1986.
7)
_XN NF 81-$1(A), "LCCA Seismic Structural Responsa of an ENC BWR Jet Pump -
Fuel Assembly," Exxon Nuclear Co., May 1986.
8)
XN NF-524(P), Revision 2, ' Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors," including Supplements, Advanced Nuclear Fuels Corporation, April 1989, 9)
XN NF-80-19(A), Volume 1, Supplements 1 & 2. " Exxon Nuclear Methodology for Boiling Water Reactors: Neutronics Methods for Design and Analysis," Exxon Nuclear Co., March 1983,
- 10) ANF-88-183(P), " Grand Gulf Unit 1 Reload XN-1.3 Cycle 4 Mechanical-Design,"
ANF Corporation, November 1988.
- 11) 'XN NF-79-71(P), Revision 2, " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors," Exxon Nuclear Co., November 1981.
- 12) NESDQ-88-003, Revision 0, " Grand Gulf Nuclear Station Unit 1 Revised Flow Dependent Thermal Limits," MSU System Services Inc., November 1988.
- 13) XN NF-80-19(A), Volumes 2, 2A, 2B, & 20, ' Exxon Nuclear Methodology for j-Boiling Water Reactors:
EXEM BWR ECCS Evaluation Model," Exxon Nuclear Co.,
j September 1982, 14)
XN NF-CC-33(A), Revision 1, "HUXY: A Generalized Multirod Heatup Code with 10CFR50 Appendix K Heatup Option," Exxon Nuclear Co., November 1975.
- 15) XN NF 82-07(A), Revision 1, " Exxon Nuclear Company ECCS Cladding Swelling l
and Rupture Model," Exxon Nuclear Co., November 1982.
l,
^
- 16) XN NF-86-37(P), " Generic LOCA Break Spectrum Analysis for BWR/6 Plants,"
Exxon Nuclear Co., May 1986.
- 17) NED0 21231, " Banked Position Withdrawal Sequence," General Electric Co.,
January 1977.
1
- 18) XN NF-825(P)(A), Supplement 2, "BWR/6 Generic Rod Withdrawal Error Analysis, MCPR for All Plant Operations Within the Extended Operating Domain,"
ExxoRNuclearCompany, October 1986.
i
- 19) XN NF-80-19(P)(A), Volume 3. Revision 2. " Exxon Nuclear Methodology for Boiling Water Reactors THERMEX:
Thermal Limits Methodology Summary Description," Exxon Nuclear Co., January 1987.
20)
XN NF 84-105(P)(A), Volume 1, "XCOBRA T: A Computer Code for BWR Transient Thermal Hydraulic Core Analysis," Exxon Nuclear Company, Inc., February 1987.
- 21) AFCM-90/0068, " Criticality Analysis for Cycle 5," Letter to NRC from J. G.
Cesare, SERI, April 26, 1990.
22)
Safety Evaluation By the Office of Nuclear Reactor Regulation Relating to Amendment 17 of GE Topical Report NEDE 240ll-P, " General Electric Standard.
l
-Application for Reactor Fuel," dated-12/27/87.
- 23) XN NF 84-97(P)(A), "LOCA Seismic Structural Response of an ENC 9x9 BWR. Jet Pump Fuel Assembly," Exxon Nuclear Company Inc., August 1984.
- 24) XN NF-8019(P), Volume 1, Supplement 3, "ANF Methodology for BWRs:
Benchmark Results for the CASMO 3G/MICR0 BURN B Calculation Methodology,"
Advanced Nuclear Fuels Corporation, February 1989.
25)
" Issuance of Amendment No. 62 to Facility Operating License No. NPF Grand Gulf Nuclear Station, Unit 1, Regarding Technical Specifications Revisions - Thermal-Hydraulic Stability (TAC No. 71808)," Letter from L
L. L. Kintner,.NRC, to W. T. Cottle, SERI, dated August 31, 1989.
- 26) ANF-913 Volume 1, Supplements 1, 2, and 3, *COTRANSA2: A Computer Program l
for Boiling Water Reactor Transient Analysis," Advanced Nuclear Fuels Corporation, June 1989.
l
- 27) ANF-ll25(P), Supplement 1, "ANFB Critical Power Correlation," Advanced Nuclear Fuels Corporation, April 1989.
l 1
i to AECM-90/0092 GGNS UNIT 1 CYCLE 5 PROPOSED STARTUP PHYSICS TESTS MAY 1990 l
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Proposed Startup Physics Tests for Cycle 5 1.
Core Loadino Verification The core will be visually checked to verify conformance to the vendor supplied core loading pattern.
Fuel assembly serial numbers, bundle orientations, and core locations will be recorded. A height check will be performed to assure that all assemblies are properly seated in their respective locations.
2.
Control Rod Functional Testing Prior to criticality following the refueling outage, functional testing of the control rods will be performed to assure proper operability.
This testing will include coupling. verification, withdrawal and insertion timing, and friction testing where required.
3.
Shutdown Marain Determination Control rods will be withdrawn in their standard sequence until criticality is achieved. The shutdown margin of the core will be determined from calculations based upon the critical rod pattern, the l
reactor period, and the modere..or temperature.
To assure there is no i
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reactivity anomaly, the actual critical control rod position will be l
l verified to be within 1% dk/k of the predicted critical control rod position.
4.
TIP Asymmetry A gross asymmetry check will be performed as part of a detailed statistical uncertainty evaluation of the TIP system.
A complete set of TIP data will be obtained at a steady state, equilibrium xenon condition 1
t greater than 85% rated power. A total average deviation or uncertainty will be determined for all symmetric TIP pairs as well as the maximum
~
t absolute deviation. The results will be evaluated to assure proper operation of the TIP system and symmetry of the core loading.
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