ML20196E956
| ML20196E956 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 11/11/1988 |
| From: | Reynolds R SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML19295G794 | List: |
| References | |
| ANF-88-150, NUDOCS 8812120147 | |
| Download: ML20196E956 (40) | |
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,7 Y:& ADVANCED NUCLEAR FUELS CORPORATION E f4a .
GRAND GULF UNIT 1 CYCLE 4 PLANT TRANSIEN T ANALYSIS l l i NOVEMBER 1988 GG1212o147 881206 fDR ADOCK 0500 6 i l , . _ , - -
ADVANCED NUCLEAll FUELS CORPORATION ANF-88-150 Issue Date. 11/11/88 i GRAND GULF VNIT 1 CYCLE 4 PLANT TRANSIPP ANALYSIS Prepared by DJ //klW R~. Ji' Reynolds BWR Sfgfety Analysis Licensing and Safety Engineering Fuel Engineering and Technical Services tb su, l l- f~ff
/ 'R. G. Grumer BWR Neutronics Neutronics and Fuel Management Fuel Engineering and Technical Services November 8, 1988
l l CUSTOMER DISCLAIMER I IMPORTANT NOTICE REGARCING CONTENTJ AND USE OF THIS DOCUMENT PLEASE RE/ O CAREFULLY Advanced Nuclear Fuoes Corporacon's warrances and reoresentatens con-cormng me sue;ect maaer of the document are thoos set form in the Agreement ) between Advanced Nucseer Fuee Corporacon and the Customer pursuant to I which thee document e eeued A:cordegty, except as otherwee expreesty pro-wood in such Agreement, nenner Advanced Nue: ear Fueos Corporator. ..or any person acnng on its betwf makes any warranty or representaten, exortese1 or imphed, wien respect to the accuracy, comc'eteness, or userumeos of tr,e efor. t mason coritamed in the document or that me use of any mformauon. accaratus. (
- m. mod w proc.es decio.ed m ini. accum.ni wm not infnnge onva:e .n.d ngfus; or efeumes any secesitee vnth respect to the de of any eformaten, ao-paratus. method or procese disclosed a this document.
The ebemenon contamed heren is for the so6e use of Customer. ) in order to evoed enoeirment of nghts of Advanced Nuc! oar Fuess Corporaten o patents or wtvetAons wasch may be ec 9ded in the informt. ton contameo in this document, me rocceent Oy its accessence of this document, agrees not to puchen or make Duchc use (in the patent use of the term) cf sucn mformaton untd , so authonzod in wnteg Dy Advanced Nuc:es* Fuets Corporaten or untd after six ' (6) rnonths followeg termnaten or expiratce of the aforesa4 Agreement and any extensson thereof. urwest otherwise encrossty proviced a the Agreement. No nghts or licensee in or to any patents us inShed Dy me furmsnirg of this cocu-meet. l l ANF-3145 472A (12.37)
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ANF 88-150 Page i I6BLI 0F CONTENTS Section Eigt 1.0 INTR 0000'.JN ........................ ... 1 2.0
SUMMARY
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 3.0 THERMAL LIMITS ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . 12 3.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . 12 3.2 System Transients ....................... 12 3.2.1 Design Basis ...................... 13 3.2.2 Anticipated Transients ................. 13 3.2.2.1 Loss Of Feedwater Heating ........... 13 3.2.2.2 Load Rejection No Bypass . . . . . . . . . . . . 14 3.2.2.3 Feedwater Controller Failure . . . . . . . . . . 15 3.2.2.4 Control Rod Withdrawal Error . . . . . . . . . . 16 3.3 Flow Excursion Analysis .................... 16 3.4 Safety Limit . . . . . . . . . . . . . . . . . . . . . . . . . . 17 3.5 Summary of Results . . . . . . . . . . . . . . . . . . . . . . . 17 3.5.1 Power Dependent Thermal Limits and Values . . . . . . . . 18 3.5.2 Flow Dependent Thermal Limits and Values ........ 18 4.0 MAXIMUM OVERPRESSURIZATION ..................... 29 4.1 Design Basis . . . . . . . . . . . . . . . . . . . . . . . . . . 29 4.2 Maximum Pressurization Transients ............... 29 4.3 Results ............................ 30
5.0 REFERENCES
............................. 33 1
ANF-88-150 Page ii LIST OF TABLES Table East 2.1 Results of Analyses ...................... 6 2.2 Operating Limit Coordinates .................. 7 3.1 Grand Gulf Unit 1 Cycle 4 LFWH Data Summary .........19 (11.' 0F FIGURES Ficure Eigg 1.1 Power / Flow Map Used For Grand Gulf Unit 1 ME00 Analysis .... 3 2.1 Power Dependent MCPR Limits For Grtnd Gulf Unit 1 Cycle 4 ... 8 2.2 Power Dependent MAPFAC Factor For Grand Gulf Unit 1 Cycle 4 .. 9 2.3 Flow Dependent MCPR Limits For Grand Gulf Unit 1 Cycle 4 . . . 10 2.4 Flow Dependent MAPFAC Factor for Grand Gulf Unit 1 Cycle 4 . . 11 3.1 Analysis of LFWH Initial MCPR Versus Final MCPR .......20 3.2 Load Rejection Without Bypass (Power and Flows) . . . . . . . 21 3.3 Load Rejection Without Bypass (Vessel Pressure and Level) . . 22 3.4 Feedwater Controller Failure (Power and Flows) . . . . . . . . 23 3.5 Feedwater Controller Failure (Vessel Pressure and Level) . . . 24 3.6 Design Basis Radial Power Distribution . . . . . . . . . . . . 25 3.7 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis Local Power Distribution (XN 12.99-5G3 Fuel) . . . . . . . . 26 3.8 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis Local Power Distribution (XN-2 3.21-6G4 Fuel) . . . . . . . . 27 3.9 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis Local Power Distribution (ANF-1.3 3.61 CG4 Fuel) ......28
- 4.1 MSIV Closure W
- thout Direct Scram (Power and Flows) . . . . . 31
( 4.2 MSIVClosureWithoutDirectScram(VesselPrassure i and Level) .........................32 I 1
4 ANF-88-150 Page iii ACKNOWLEDGEMENT The authors would like to acknowledge the following individuals for their contribut-lons to the results reported in this document: ,
.; D. J. Braun M. E. Byram S. J Haynes D. E. Hershberger M. J. Hibbard D. F. Richey S. 6. State ,
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ANF-88-150 Page 1
1.0 INTRODUCTION
This report presents the results of analyses performed by Advanced Nuclear Fuels Corporation (ANF) for reload fuel in Grand Gulf Unit 1 Cycle 4 for operation within the Maximum Extended Operating Domain (ME00). The NSSS vendor performed extensive transient analyses for Grand Gulf Unit 1 in conjunction with the extension of the power / flow operating map to the ME00 in Cycle 1 (Reference 1). These analyses established conservative operating limits for ME00 operation. The initial reload of ANF fuel in Grand Gulf Unit i occurred in Cycle 2. In support of the initial reload of ANF fuel, extensive additional transient analyses were performed by ANF to either justify the NSSS vendor operating limits or, where necessary, to provide appropriate limits for ANF fuel using ANF methodologies (Reference 2). The objective of these analyses was to confirm the applicability of the Grand Gulf Unit 1 Cycle 3 Technical Specification MCPR at rated conditions, establish MAPLHGR limits for Cycle 4 operation, and establish revised thermal limits for off-rated conditions for the all-ANF core. An additional objective l was to demonstrate that vessel integrity is protected during the rmt limiting Cycle 4 pressurization event. Changes from Cycle 3 to Cycle 4 for Grand Gulf Unit 1 include the discharge of remaining GE fuel, an additional reload of ANF fuel and vi ! increase in cycle energy from 1420 GWd to 1698 GWd while maintaining the cycle length at 18 months. The reload fuel for Cycle 4 is the same as that for Cycles 2 and 3 except for changes in enrichment, the number of rods per bundle containing gadolinia, and the gadolinia concentration (Reference 3). The Cycle 4 transient analysis consists of recalculation of the limiting transients at state points having the least margin to operating limits to confirm that the effects of the Cycle 4 changes on transient results are small and establish appropriate limits. Reanalysis of the limiting transients for Cycle 4 assures that the less limiting transients which were previously addressed will continue to be protected by the established operating limits for Cycle 4. The power / flow conditions analyzed in Cycle 3 and Cycle 4 are presented in Figure 1.1.
1 l ANF-88-150 l Page 2 j l l The MCPR p , MCPRr, and MAPFACr limits have been revised to reflect ANF calculated limits using ANF methodology. The Grand Gulf Unit 1 power and flow dependent MCPR analyses for Cycle 4 were performed at limiting power / flow conditions. Flow dependent MAPFAC analyses were performed on the 100% rod line with the initial core flow varying from 40% to 80% of rated flow, l l l l l l l l
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ANF-88-150 Page 4 2.0 SUMARY - The results of the Grand Gulf Unit 1 Cycle 4 transient analyses support appropriate thermal limits for the first Grand Gulf all-ANF core. ANF thermal limits have been provided for MCPRp above 40% power that are based on generic ANF Control Rod Withdrawal Error (CRWE) analyses (Reference 4). Additionally, MCPRf limits and MAPFACf values (Reference 12) have been confirmed for both "loop manual" and "non-loop manual" operation. Minor differences in the maximum local peaking as a function of exposure for the different ANF fuel types require that different MAPLHGR limits be monitored. These MAPLHGR limits are consistent (differ by maximum local peaking factor) with the LHGR limits so that at reduced power and/or reduced flow the LHGR limit will be protected by the MAPFACf and MAPFACp multipliers on MAPLHGR. Cycle 4 reload fuel MAPLHGR limits are included because of the slight changes in the local peaking factor. Table 2.1 summarizes the transient analyses results applicable to Grand Gulf Unit 1 Cycle 4. These results, together with the Grand Gulf Unit 1 Cycle 4 calculated safety limit MCPR of 1.06, support continued use of the existing 1.18 MCPR operating limit (at rated conditions) for Cycle 4 operation. The plant transient and safety limit analyses results reported herein support revising the Cycle 3 power dependent Minimum Critical Power Ratio (MCPRp ) so that it is based on the generic CRWE results of Reference 4 above 40% power and supports the continued use of Cycle 3 limits below 40% power. The power dependent Maximum Average Planar Lincar Heat Generation Factor (MAPFAC p ) for Cycle 3 is confirmed for Cycle 4 operation. The revised MCPf.p limits, the MAPFAC p confirmation, and the results of ANF's analyses are presented in Figures 2.1 and 2.2, respectively. The flow dependent Minimum Critical Power Ratio (MCPR f ) and the results of ANF's analysis are presented in Figure 2.3. The flow dependent Maximum Average Planar Linear Heat Generation Rate Factor (MAPFAC f ) is presented in (
. _ _ _ _ _ - - _ - _ _ _ _ _ _ _ . _ . . . _ _ _ _ _ )
ANF-88-150 Page 5 Figure 2.4. These flow dependent MAPFACf values and MCPRf limits have been revised from Cycle 3 to support Cycle 4 in both the "loop manus 1" and the "non-loop manual" mode of operation. These curve, are based on conservative maximum core flow rates. Table 2.2 shows the coordinates used to construct Figures 2.1 through 2.4. The results of the maximum system pressurization transient analysis are presented in Table 2.1. The safety valve pressure setpoint tolerances in this analysis have been increased to 6% for Cycle 4; the results show that the Grand Gulf Unit I safety valves have sufficient capacity and performance wMh the increased setpoint tolerances to protect the vessel pressure safety limit of 1375 psig during Cycle 4. The fuel related Technical Specification limits for Cycle 4 operation are included in the reload analysis report (Reference 3). l
ANF-88-150-Page 6 Table 2.1 Resu!ts of Analyses THERML LIMITS Transient Delta CPR Loss of Feedwater Heating (all conditions) 0.11 Control Rod Withdrawal Error (100% power, Ref. 4) 0.10 Feedwater Controller Failure (104.2/108)* 0.04 Load Rejection Without Bypass 5_fower/% Core Flow 104.2/108* 0.12
, 104.2/73.g, 0.02 40/108*** 0.15 40/108 0.32 25/73.8 0.93 t 25/40 0.69 104.2/100 0.09 92.5/67 0.02 70/40 0.04 55/40 0.03 40/40** 0.03 MAXIMUM SYSTEM PRESSURIZATION J
Transiertt % Power /% Core Flow yessel Lower Plenum Steam Dome MSIV Closure 104.2/108* 1298 psig 1271 psig 104.2/73.8 1297 psig 1280 psig
*104.2% power /108% core flow is used for the Reload Licensing Analysis (RLA) y conditions to conservatively bound 100% power /105% core flow. ** Direct scram on turbine trip enabled *** Direct scram on turbine trip disabled i
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ANF-88-150 Page 7 Table 2.2 Operating Limit Coordinates GRAND GULF UNIT 1 CYCLE 4 MCPR(o) Limits MAPFACfo) Limits (Figure 2.1) (Figure 2.2) Percent of Rated Percent of Rated Core Power MCPR(o) Core Power MAPFAC(o) 100 1.18 100 1.0 70 1.24 40 0.60* I 70 1.40 40 0.69** 40 1.48 24.4 0.57* ' 40 1.85* 24.4 0.61** 40 2.10** 25 2.05***
- Core tiow > 50%
25 2.15 ** Core Flow < 50%
- Core Flow < 50% -
** Core Flow 5 50% ,
l l MCPR(f) Limits MAPFAC(f) Limits (Figure 2.3) (Figure 2.4) , Percent of Loop Non-loop Percent of Loop Non-loop ! Rated Core Manual Manual Rated Core Manual Manual i Flow MCPR(f) MCPR(f) Flow MAPFAC(f) MAPFAC(f) ' i 30 1.55 1.73 110 1.00 1.00 40 1.41 1.57 91.0 1.00 1.00 50 1.31 1.44 90.0 1.00 .992 1 60 1.24 1.35 84.3 1.00 - 70 1.19 1.27 80.0 .977 .904 73.4 1.18 - 70.0 .928 .827 80 1.18 1.21 60.0 .880 .757 86.3 1.18 1.18 50.0 .837 .695 ! 105 1.18 1.18 40.0 .794 .638 l 30.0 .752 .586 - l t l
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ANF-88-150 Page 12 3.0 THER' MAL LIMITS ANALYSIS . 3.1 Introduction The scope of the thermal limits analysis incitides system transients, localized core events, and safety limit analysis. Results of these analyses e are used to confirm / establish power and flow dependent MCPR and MAPFAC values, COTRANSA (Reference 5), XCOBRA-T (Reference 6), and XTGBWR (Reference 7) Y. are the major codes used in the thermal limits analyses as described in ANF's THERMEX Methodology Report (Reference 8) and Neutronics Methodology Report (Reference 7). COTRANSA is e transient system simulation code which includes an axial one-dimensional neutronics model. XCOBRA-T is a transient thermal-hydraulic code used in the analysis of thermal margins of the limiting fuel assembly. XTGBWR is a three-dimensional steady state core simulation code which is used for Control Rod Withdrawal Error (CRWE), Loss of Feedwater Heating (LFWH), and flow excursion events. 3.2 System Transients Revised thermal limits were established for the all-ANF Cycle 4 core, f Figure 1.1 shows the ten power / flow conditions that were analyzed in support of the Cycle 4 reload. These state point: were analyzed for Grand Gulf Unit 1 Cycle 4 using COTRAllSA. The load Reject No Bypass (LRNB) pressurization transient analysis was performed at each of the ten state points. The Feedwater Controller Failure (FWCF) analysis was performed at 104.2% power and 108% flow. ASME pressurization analyses were performed at state points 104.2%/108% and 104.2%/73.8%. LFWH analyses were performed with XTGBWR at five different state points for eight exposures. The generic analysis for control rod withdrawal error is applicable to Grand Gulf Unit 1 Cycle 4. These analyses show less restrictive results or little change from the Cycle 3 analyses due to Cycle 4 changes, thus justifying that the less limiting transients not analyzed for Cycle 4 will continue to be protected.
ANF-88-150 Page 13 S.2.1 Desian Basis The LRNB and FWCF transients have been determined to be most limiting at end of full power capability when control rods are fully withdrawn from the core. The delta CPR calculated for end of full power conditions is conserva-tive for cases where control rods are partially inserted. The analysis for Grand Gulf Unit I with ME00 was performed using conservative analytital limits for trips and setpoints. Events initiated at core powers below 40% rated were analyzed with the direct scram due to turbine control and stop valve fast closure disallowed, and with the recirculation pump high to low speed transfer disabled- The Loss of Feedwater Heating (LFWH) transient has been analyzed throughout the cycle at state points which bound the ME0D operating map. 3.2.2 Anticioated Transients ANF's transient methodology report for jet pump BWRs (Reference 5) considered eight categories of anticipated transients. The most limiting transients were evaluated at various power / flow points within ME00 to verify the power dependent thermal margin for Grand Gulf Unit 1 Cycle 4. The limiting transients analyzcd for Grand Gulf Unit 1 Cycle 4 were: 1 o loss of Feedwater Heating o load Rejection No Bypass o Feedwater Controller Failure 1 a l Other transients are inherently non limiting or bounded by one of the above as shown in the NSSS vendor HE00 analyses for Cycle 1 and the ANF Grand Gulf Unit 1 Cycle 2 analyses. Control Rod Withdrawal Error is an exception in that it has been analyzed generically. - l 3.2.2.1 Loss Of Feedwater Heatina 1 l Analysis of the loss of feedwater heating event was performed to reflect I reactor operation over the ME00 operating power versus flow map and conditions - anticipated during actual Grand Gulf reactor operation.
ANF-88-150 Page 14 Calculations performed for Cycle 4 assumed a conservative reduction of 100*F in the feedwater temperature. Table 3.1 provides the conditions of each case analyzed in terms of cycle exposure, core power, and core flow. The initial and final MCPR values are presented for each case. Analysis of the data revealed a strong correlation between the initial and final MCPR, A least squares fit of these data resulted in a linear relationship such that: MCPR(initial) ~ 0.0514 + 1.1130
- MCPR(final).
In order to conservativ21y bound all of the calculated data, the largest deviation between the calculated and fitted results were applied to the least squares fit such that the LFWH MCPR operating limit is defined by OLHCPR(LFWH) = -0.0112 + 1.1130
- SLHCPR R .
This bounding relationship is presented in Figure 3.1. Substituting the SLHCPR of 1.06, the MCPR operating limit for the LFWH event for all operating conditions is 1.17. 3.2.2.2 Load Re.iection No BYDasS The Load Rejection No Bypass (LRNB) event is the most limiting of the class of transientt characterized by rapid vessel pressurization for Grand Gulf Unit 1. The lead rejection causes a fast closure of the turbine control valves. The resulting compression wave travels through the steam lines into the vessel and creates the rapid pressurization condition. A reactor scram is initiated by the fast closure of the contrcl valves as well as the recircula-tion pump high to low speed transfer. Condenser bypass flow, which can mitighte the pressurization effect, is not allowed. The excursion of the core power due to void collapse is primarily terminated by reactor scram and void growth due to the recirculation pump high to low speed transfer.
ANF-88-150 Page 15 Figures 3.2 and 3.3 present the response of various reactor and plant l parameters to the LRNB event initiated at the Reload Licensed Analysis condi- ! tion (104.2% power /108% core flow). Table 2.1 lists the delta CPRs for this transient at the other power / flow conditions analyzed for Grand Gulf Unit 1. 1 i 3.2.2.3 Feedwater Controller, Failure l The failure of the feedwater controller to maximum demand (FWCF) is the I most limiting of the vessel inventory increase transients. Failure of the feedwater control system to maximum demand would result in an increase in the l coolant level in the reactor vessel. Increased feedwater flow results in lower temperatures at the core inlet, which in turn cause an increase in core power level. If the feedwater flow stabilizes at the increased value, the core power will stabilize at a new, higher value. If the flow increase continues, the water level in the downcomer will eventually reach the high level setpoint, at which time the turbine stop valve is closed to avoid damage to the turbine from excessive liquid inventory in the steamline. The high water level trip also initiates reactor scram, and recirculation pump trip. Turbine bypass is assumed to function for this analysis, mitigating the consequences tG some extent. The core power excursien is terminated by the same mechanisms that end the LRNB transient. Figures 3.4 and 3.5 present the response of various reactor and plant parameters to the FWCF event initiated at the Reload Licensed Analysis condi-tion (104.2% pcwer/108% core flow). The delta-CPR for this event was calculated to be 0.04, indicating a MCPR operating limit requirement of 1.10 for the event, in support of the Cycle 2 reload, FWCF transients were also analyzt.d without condenser bypass and with a 100*F reduction in feedwater l temperature. It was shown that tnese conditions had a minor impact on the delta-CPR and that sfgnificant margin exists to limits (Reference 2). Since the FWCF transient analyzed for Cycle 4 results in a delta-CPR similar to that obtained for Cycle 3, it is not necessary to repeat the other FWCF transients fer the Cycle 4 reload. In Reference 2. the LRNB transient was shown to bound all FWCF transients at rated and off rated conditions.
I ANF-88-150 Page 16 3.2.2.4 Control Rod Withdrawal Error Reference 4 documents ANF's generic CRWE analysis for Grand Gulf Unit 1 operation within the HE00. This generic analysis is applicable to Cycle 4. The results from Reference 4 and the results of the Cycle 4 system transient confirmatory analyses show that the CRWE is limiting above 40% of rated power. The Grand Gulf Cycle 4 CRWE based limits, and analysis results are presented in Figure 2.1. These data demonstrate that the CRWE 1"Jts may be used as a basis for the Grand Gulf Unit 1 MCPR p Technical Specification limits in Cycle 4 above 40% power. The rated condition MCPR operating limit remains unchanged at 1.18. 3.3 Flow Excursion Analysis The flow excursion transient is analyzed to determine the flow dependent thermal limits and values (MCPRf and MAPFACf ). This transient is analyzed by assuming a failure of the recirculation flow control system such that the recirculation flow increases slowly to the physical maximum attainable by the equipment. Two modes of operation are analyzed for Grand Gulf Unit 1 Cycle 4, "loop manual" and "non-loop manual." These two modes of operation correspond to a single recirculation loop flow excursion event and a dual recirculation loop flow excursion event, respectively. For both flow excursion events, the Cycle 4 MCPRf confirmation analysis of the power ascension associated with a flow increase was determined to be conservative when compared to the MCPRf operating limit (Reference 12). In the confirmation calculation the change in critical power along the ascension
;.ath was calculated with XCOBRA (Reference 8). Peaking factors were selected such that the bundle with the least margin would reach the safety limit MCPR of 1.06 at the maximum flow. Figure 2.3 presents the MCPRf limits for maximum achievable core flows for both events, assuming that the recirculation system equipment is capable of 110% of rated.
The Cycle 4 MAPFACf confirmation analysis of the power ascension associated with a flow increase was determined to be conservative when compared to MAPFACf limits. Confirmation calculations were performed for
l ANF-88-150 l Page 17 j 1 both "loop manual" and "non-loop manual" modes of operation with XTGBWR. The Cycle 2 analysis of reduced flow LHGR limits (MAPFACf values) was performed statistically based upon a wide variety of initial conditions. For "non-loop , manual" operation, confirmatory calculations were performed for Cycle 4 at O.5, 2.0, 3.5, 6.0, 8.0, 9.5, and 11.0 GWd/MTV with XTGBWR to simulate the flow runup event from 40% of rated flow. For "loop manual" operation, these analyses simulated a flow runup where the initial flow was varied from 40% to 80% of rated. Final flow was established based on the mode of operation (Reference 12). Figure 2.4 presents the MAPFACr limits for maximum achievable ( core flows for both events as well as the resuits of the Cycle 4 analysis. 3.4 Safety Limit i The safety limit MCPR is defined as the minimum value of the critical l power ratio at which the fuel could be operated, with the expected number of rods in boiling transition not eruding 0.1% of the fuel rods in the core. The safety limit is the minimum critical power ratio which would be permitted l to occur during the limiting anticipated operational occurrence. The safety limit MCPR for all fuel types in Grand Gulf Unit 1 Cycle 4 operation was confirmed io remain at 1.06 using the methodology presented in References 9 and 11. l The input parameter values for uncertainties used in the safety limit MCPR analysis are unchanged from the Cycle 2 analysis presented in Reference 2. Cycle 4 specific design basis radial and local power distribu-tions are shown in Figures 3.6 to 3.9. 3.5 Summary of Results The results of the Grand Gulf Unit 1 Cycle 4 thermal limits analysis confirm the Cycle 3 safety limit MCPR of 1.06 and a MCPR operating limit cf 1.18 at rated conditions. l l l l 1 l l
ANF-88-150 Page 18 3.5.1 Power Deoendent Thermal Limits and Values The power dependent MCPR limit (MCPR p ) protects against exceeding the safety limit MCPR during anticipated operational occurrences from off-rated power conditions. The MCPR p limit is determined by adding the delta CPR for the limiting event to the calculated safety limit MCPR. Tl y power dependent MAPFAC (MAPFAC p ) is used to protect against both fuel meltiric and 1% clad strain during anticipated system transients from off-rated power conditions. The conservative LPGR values for protection against fuel failure during anticipated operational occurrences are given in Reference 10. The results are then presented in a fractional form for application to the MAPLHGR value. The MAPLHGR is developed to be consistent with the LHGR limit through con:ideration of the maximum local peaking factor. The MCPR p limits and MAPFAC p values for Cycle 4 are shown to bound the results of ANF's analysis in Figures 2.1 and 2.2, respectively. Above 40% power the MCPRp limit is based on the ANF CRWE limit of Reference 4. Below 40% power, the Cycle 3 MCPRp limit remains applicable to Cycle 4. The Cycle 4 MAPFAC p value remains unchanged from Cycle 3. 3.5.2 Flow Decendent Thermal limits and Values The flow dependent MCPR limit (MCPRr) protects against exceeding the safety limit MCPR for flow excursion events. The results of the MCPRr analysis for Grand Gulf Unit 1 Cycle 4 ard presented in Figure 2.3. The flow dependent MAPFAC (MAPFACr) protects against both fuel molting and 1% clad strain. The MAoFACr values to be used in Cycle 4 are presented in Figure 2.4. The flow dependent thermal limits were confirmed to be conservativo for Cycle 4 operation.
ANF-88-150 Page 19 Table 3.1 Grand Gulf Unit 1 Cycle 4 LFWH Data Summary Initial State Final State Cycle Total Core Total Core Core Total Core Total Core Core Exposure Power Flow Minimum Power Flow Minimum (GWd/MT) MWt !M1b/hr) _ _C PR MWt (M1b/hr) CPR 0.500 3833.0 118.13 1.35 4369.2 118.13 1.27 0.500 3066.4 118.13 1.68 3506.1 118.13 1.57 0.500 3833.0 84.38 1.20 4356.2 84.38 1.14 0.500 2376.5 34.88 1.41 2736.8 34.88 1.31 0.500 1533.2 112.50 3.17 1735.8 112.50 2.91 2.000 3833.0 118.13 1.35 4387.9 118.13 1.25 2.000 3066.4 118.13 1.68 3514.5 118.13 1.54 2.000 3833.0 84.38 1.24 4379.7 84.38 1.16 2.000 2376.5 34.88 1.41 2753.2 34.88 1.29 2.000 1533.2 112.50 3.25 1743.8 112.50 2.93 3.500 3833.0 118.13 1.36 4364.9 118.13 1.28 3.500 3066.4 118.13 1.69 3500.2 118.13 1.57 3.500 3833.0 84.38 1.19 4360.2 84.38 1.13 3.500 2376.5 34.88 1.37 2733.9 34.88 1.2) 3.500 1533.2 112.50 3.27 1728.0 112.50 3.00 5.000 3833.0 118.13 1.31 4359.8 118.13 1.23 5.000 3066.4 118.13 1.63 3498.4 118.13 1.51 5.000 3833.0 84.38 1.18 4358.6 84.38 1.12 5.000 2376.5 34.88 1.36 2735.0 34.88 1.26 5.000 1533.2 112.50 3.17 1725.1 112.50 2.89 6.500 3833.0 118.13 1.29 4357.0 118.13 1.21 6.500 3066.4 118.13 1.60 3490.5 118.13 1.49 6.500 3833.0 84.38 1.15 4149.7 84.38 1.09 6.500 2376.5 34.81 1.36 771'.3 34.88 1.24 6.500 1533.2 112.50 3.02 1728.3 112.50 2.76 8.000 3833.0 118.13 1,29 4348.8 118.13 1.20 8.000 3066.4 118.13 1.59 3486 '6.13 1.47 8.000 3833.0 84.38 1.13 434t.5 4.38 1.07 8.000 2376.5 34.88 1.35 2729.4 .4.88 1.24 8.000 1533.2 112.50 2.95 1723.9 112.50 2.70 9.500 3833.0 118.13 1.32 4337.3 118.13 1.23 9.500 3066.4 118.13 1.63 3471.8 110 13 1.51 9.500 3833.0 84.38 1.16 4336.3 84.J8 1.09 9.500 2376.5 34.88 1,39 2718.6 34.88 1.30 9.500 1533.2 112.50 3.02 1716.7 112.50 2.78 11.000 3833.0 118.13 1.30 4329.4 118.13 1.21 11.000 3066.4 118.13 1.60 3467.1 116.13 1,48 11.000 3833.0 84.38 1.15 4337.1 84.38 1.08 11.000 2376.5 34.88 1.40 2713.7 34.88 1.30 11.000 1533.2 112.50 2.96 1718.7 112.50 2.71
= - - _ _ . -_
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Figure 3.7 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis Local Power Distcibution (XN.1 2.99 5G3 Fuel)
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f P ANF 88 150 ! Page 27
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Figure 3.8 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis Local Power Distribution (XN.2 3.21 6G4 Fuel) l I Gadolinia location ' 4 I i i i l l
I t ANF-88 150 Page 28
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ANF-88 150 Page 29 4.0 MAXIMUM OVERPRESSURIZATION Maximum system pressure has been calculated for the containment isolation event (rapid closure of all main steam isolation valves) with an adverse scenario as specified in the ASME Pressure Vessel Code. This analysis showed that the Grand Gulf Unit I safety valves have sufficient capacity and performance to prevent pressure from reaching the established transient pressure safety limit of 110% of design pressure (1.1 x 1250 1375 psig). The maximum vessel pressures at the most limiting power / flow point (104.2% power /108% flow) are shown in Table 2.1. t 4.1 Desion Basis j During the transient, the most critical active component (direct scram on MSIV closure) was assumed to fail. The event was terminated by the high flux scram. Credit was taken for actuation of only 13 of the 20 safety / relief valves: 6 in the relief mode and 7 in the safety mode. The calculation was performed with ANF's plant simulation code COTRANSA, which includes an axial one dimensional neutronics model. The safety valve analysis setpoints for this calculation included a 6% tolerance. Relief valve setpoints for this analysis remain unchanged from Cycle 3. 4.2 Maximum pressurization Transients Scoping analyses described in Reference 5 found the closure of all main steam isolation valves (MSIVs) without direct scram to be limiting. The MSIV [ closure was found to be limiting whan all transients are evaluated on the same I basis (without direct scram) because of the smaller steam line volume associated with MSIV closure. Though the closure rate of the MSIVs is i substantially slower than turbine stop or control valves, the compressibility I i of the additional fluid in the steam lines associated with a turbine isolation causes these faster closures to be less severe. Once the containment is isolated, the subsequent core power production must be absorbed in a smaller volume compared to that of a turbine isolation resulting in higher vessel pressures. i
i AkF-88 150 Page 30 t 4.3 Results The results of the maximum system pressurization analysis are presented in Tatle 2.1. Figures 4.1 and 4.2 present the respc::se of various reactor and plant parameters during the MSIV closure event from 104.2% power /108% flow. These results show that the Grand Gulf Unit 1 safety valves have sufficient capacity and performance to protect the previously established
- maximum vessel pressure safety limit of 1375 psig for Cycle 4. Two state points were analyzed in order to cover the ME00 range for full power operation.
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ANF 88 150 Page 33
5.0 REFERENCES
- 1. Lester L. Kintner, USNRC, Letter to 0. D. Kingsley, Jr., MP&L, "Technical Specification Changes to Allow Operation with One Recirculation Loop and ,
Extended Operating Domain," August 15, 1986.
- 2. "Grand Gulf Unit 1 Cycle 2 Plant Transient Analysis," XN-NF-86-36, ;
Revision 3, Exxon Nuclear Company, Inc., Richland, WA, August 1986.
- 3. "Grand Gulf Unit 1 Cycle 4 Reload Analysis," ANF 88149, Advanceu Nuclear
,, Fuels Corporation Richland, WA, November 1988.
- 4. for Plant Operations "BWR/6 Generic Rod Withdrawal Error Analysis; within th', Extended Operation Domain," XN NF-82_5 MCPRp(P)(A),
Supplement 2, , Exxon Nuclear Company, Inc., Richland, WA, October 1986. t
- 5. "Exxon Nuclear Plant Transient Methodology for Boilirig Water Reactor,"
XN NF-79-71(P), Revision 2, including Supplements 1, 2 & 3(A), Exxon Nuclear Company, Inc., Richland, WA, November 1981.
- 6. "XCOBRA-T: A Computer Code for > BWR Transient Thermal Hydraulic Core i Analysis," XN NF-84-105(P)(A), Volume 1, Exxon Nuclear Company, Inc.,
Richland, WA, February 1987.
- 7. "Exxon Nuclear Methodology for Boiling Water Reactors. Neutronics Methods for Design and Analysis," XN NF 80-19(A), Volume 1. Exxon Nuclear Compan/, Inc., Richland, WA, March 1983; as supplemented by letter, I R. A. Copeland, Advanced Nuclear Fuels, to M. W. Hodges, USNRC, "Void :
History Correlation," RAC:058:88, September 13, 1988. l
! 8. "Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Thermal ,
Limits Methodology Summary Description," XN NF-80 19(?)(A), Volume 3, t j Revision 2. Exxon Nuclear Company, Inc., Richland, WA, January 1987, f i i 9. "Exxon Nuclear Critical Power Methodology for Boiling Water Reactor," I i XN NF-524(P)(A), Revision 1, Exxon Nuclear Company, Inc., Richland, WA, ; November 1983.
- 10. "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel,"
XN NF-85 67(P)(A), Revision 1 Exxon Nuclear Company, Inc., Richland, WA, September 1986.
- 11. "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the !
ENC Methodology to BWR Reloads," XN NF-80-19fP)(A), Volume 4, i Revision 1. Exxon Nuclear Company, Inc., Richland, WA, June 1986,
- 12. "Grand Gulf Nuclear Station Unit 1 Revised Flow Dependent Thermal l Limits," NESDO 88-003, MSU System Services Inc., November 1988. [
t i i
?
6 ANF-88-150 Issue Date: 11/11/88 ( 4 i I 1 i GRAND GULF UNIT 1 CYCLE 4 PLANT TRANSIENT ANALYSIS l i : r Qistribution l D. A. Adkisson ! ! D. J. Braun i
- 0. C. Brown M. E. Byram R. E. Collingham ;
d R. A. Copeland f , W. S. Dunnivant L. J. Federico 7 N. L. Garner ( R. G. Grummer - D. E. Hershberger l
; M. J. Hibbard 4
T. L. Krysinski : ! A. Reparaz !
- R. S. Reynolds ;
1 S. E. State e R. B. Stout
, C. J. Volmer ! G. N. Ward t H. E. Williamson SERI/N. L. Garner (40) 4 Document Control (5) l f b l i i
i t i i i t l I I f !}}