ML20235U819
| ML20235U819 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 08/31/1987 |
| From: | Collingham R, Grummer R, Ingham J SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML20235U774 | List: |
| References | |
| ANF-87-66, ANF-87-66-R01, ANF-87-66-R1, NUDOCS 8710140275 | |
| Download: ML20235U819 (41) | |
Text
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1 ANF-87-66 REVISION 1 t
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fIDVANCEDNUCLEARFUELSCORPORMIO GRAND GULF UNIT 1 CYCLE 3 PLANT TRANSlENT ANALYSIS I
AUGUST 1987-G i
AN AfflLIAll OF AAAFTWERK voyrON 0
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? l ADVANCEDNUCLEARFUELS CORPORATION l 1
ANF-87-66 {
Revision 1 1
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. Issue Date: 8/10/87 GRAND GULF UNIT 1 CYCLE 3 PLANT TRANSIENT ANALYSIS Prepared By: 0 // , j , ,, I, ,,, f_s -r7 K G. Ingham
(/ ' BWR,' Safety Analysis Licensing and Safety Engineering -
Fuel Engineering and Technical Services Prepared By: It,2/
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/ R. G. Grummer BWR Neutronics i Neutronics and Fuel Management !
Fuel Engineering and Technical Services i I
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AN arvtuart or nRAFTWERK UNCN
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L ANF-87-66 .l Revision 1 l p- ,
' Issue Date: 8/10/87 L . . , .
' GRAND GULF UNIT 1 CYCLE 3 PLANT TRANSIENT-ANALYSIS Prepared By: J. -G. Ingham, BWR Safety Analysis and -
R. G. Grummer, BWR-Neu ronics p Concur: [ uje ,
R. E..Collin m, Manager
//I@9 Dat'e BWR~ Safety alysis Concur:
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L G< N.' Ward, Managdr ' Date L Reload Licensing 6.pv %
~ Con' cur:. / I e
' ' J jrf Morgan, Manager f7//7
/ Date i Customer Services Engineering
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Approve: M L. J/ Federico, Manager / 7/77 Dafe Neutronics and Fuel Management Approve:
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/I, __ m E.' Willi'ams6n, Manager f/4/f7 Date' Licensing and Safety Engineering L
Approve: .
lo G. L. Ritter, Manager Date Fuel Engineering and Technical Services llh 1
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l CUSTOMER DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY Advanced Nuclear Fuels Corporation's warranties and representations con-coming the subject matter of this document are those set forth in the Agreement .
between Advanced Nuclear Fuele Corporation and the Customer pursuant to 1 which this document is issued. Accordingly, except as otherwise expreesty pro.
vided in such Agreement, neither Advanced Nucteer Fuels Corporation nor any person acting on its behalf makes any warranty or representation, expressed or impiled, with roepect to the accuracy, completeness, or usefulness of the infor- {
mation contained in this document, or that the use of any information, apparatus, l method of process disclosed in this document will not infringe privately owned rights; or assumes any liabilities with respect to the use of any information, ap-paratus, method or process dieciceed in this document. ' ,
The information contained herein is for the soie use of Customer.
in order to avoid impairment of rights of Advanced Nuclear Fusie Corporation in potents or inventions which may be included in the information contained in this .
documem, the recipient, by its acceptance of this document, agrees not to ,
publish or tre he public use (in the patent use of the term) of such information until ( '
so authonted in writing by Advanced Nuclear F:Jois Corporat6on or until after six (6) months following termination or expiration of the aforesaid Agreement and any
! extensson thereof, unless otherwise expressly provided in the Agreement. No nghts or liconess in or to any patents are implied by the fumishing of this docu-ment.
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Resis'ionelYtoH ANF-87-66 .is fissued . to incorporated editorial changes? These
- changesi include adding a; reference. to the Grand Gulf Unit' ) Cycle 3 . Reload
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Anal'y si,s : and5 replacing: the fuel type identifications ?XN-1, XN-2 3.21-6G4' and
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- 3.21'.-8G4 with MF299E5G3S8, ANF321E6G4S8. and, > ANF321E8G45S; o 3
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The new fuel type identification _is ' explained as- follows, F
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Gadolinia:2 3 .
's Five rods _per bundle.containing Gadolinia The specified U235 enrichment- is for the Enriched zone (B is used to designate' Bundle average enrichment) 0; 2.99 w/o1 enrichment of U235- l Advanced Nuclear Fuel i
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i TABLE OF CONTENTS Section Paae 1.0 INTR 000CTION.................................................... I 2.0
SUMMARY
......................................................... 2 I
3.0 THERMAL LIMITS ANALYSIS......................................... 11 3.1 Introduction.................................................... 11 3.2 System Transients............................................... 11 3.2.1 Design Basis.................................................... 12 3.2.2 Anticipated Transi6nts........................................... 12 3.2.2.1 Loss of Feedwater Heating....................................... 13 3;2.2.2 Load Rejection Without Bypass................................... 13 3.2.2.3 Feedwater Controller Failure.................................... 14 3.2.2.4 Control Rod Wi thdrawal Error. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 3.3 Fl ow Excurs i on An alys i s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 3.4 Safety Limit.................................................... 16 1
.3.5 Results......................................................... 16 j i
3.5.1 Power Dependent - Thermal Limits And Val ues . . . . . . . . . . . . . . . . . . . . . . . 16 3.5.2 Flow Dependent Thermal Limits And Values.'....................... 17 i 4.0 MAXIMUM 0VERPRESSURIZATION...................................... 28 1 i
) 4.1 D e s i g n B a s i s . . . . . . . . . . . .' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 f-4 4.2 Maximum Pressurization Transients............................... 28 .
4.3' Results......................................................... 29 5.0
%EFERENCES................................ ..................... 32 I
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l gj LIST OF TABLES l- 4 i-L Table *' Eagg 2.1- Re s u l' t s ' o f An al y s e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 30, V : LF9H Transient Analysis ~ Data Summary - Grand Gulf Reactor-XTGBWRpesults' Assuming 100*FDropInFeedwaterTemprature..... 18 ff !
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LIST OF FIGURES 2
Fiouro Pace
~2.1 Power / Flow' Map Used: For Grand Gulf Unit 1 ME00 Analysis. . . . . . . . . 5-
.2.2 Power Dependent MCPR Limits For Grand Gulf Unit 1 Cycle 3. . . . . . . 6 2.3 Power Dependent MAPFAC' Factor For Grand Gulf Unit 1 Cycle 3.. . .. 7 2.4 ' Grand Gulf Unit 1 Cycle 3 MCPP Limits.vs. CRWE Based MCPR p Limit...................j .................................. 8 2.5 ~F low Dependent MCPR Limits For Grand Gulf Unit 1 Cycle 3. . . . . . . . 9 2.6 Flow Dependent MAPFAC Factor For Grand Gulf Unit 1 Cycle 3. . ... . . 10 3.1 Load Rejection Without Bypass (Power And Flows)................. 19 3.2 - Load Rejection Without Bypass (Vessel Pressure And Level) . . . . . . . 20 3.3 Feedwater Controller Failure (Power And Flows) . . . . . . . . . . . . . . . . . 21 3.4 Feedwater Controller Failure (Vessel Pressure And Level)_ . . . . .. . 22 3.5 Design Basis Radial Power Distribution. . . . . . . . . . . . . . . . . . . . . . . . . 23 3.6 Design Basis Local Power Distribution - G.E. Fue1. . . . . . . . . . . . . . . 24 3.7 Design Basis Local Power Distribution - ANF299E5G3S8 Fuel . . . . .. . 25 38 Design Basis local Power Distribution - ANF321E6G4S8 Fuel . . . . . . . 26 3.9 Design Basis Local Power Distribution - ANF321E8G4S8 Fuel. . . . . . 27 4.1 MSIV Closure Without Direct Scram (Power Md Flows) . . . . . . . . . . . . . 30 4.2 MSIV Closure Without Direct Scram (Vessel Pressure And Level)... 31
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Revision:1 ACKNOWLEDGEMENT.
The - authors would.~ like - to ' acknowledge : the' .following -individuals for their l contributions to.the results reported in this document:
'M. J. Ades
.D. J..Braun-S...E. Jensen J. A'.~White-W.C.-Arcieri-(ENSA)
B. J. Gitnick (ENSA)-
D. A. Prelewicz (ENSA)
J. C. Rawlings (ENSA)
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1 ANF-87-66 Revision 1 9
1.0 INTRODUCTION
This report presents the results of analyses performed by Advanced Nuclear Fuels Corporation (ANF), formerly Exxon Nuclear Company, Inc., for Grand Gulf Unit 1 Cycle 3 operation within the Maximum Extended Operating Domain (ME00).
The NSSS vendor performed extensive transient analyses for Grand Gulf Unit' )
in conjunction with the extension of the power / flow operating map to the ME00 in Cycle 1 (Reference 1). These analyses established appropriate operating limits for ME0D operation. The initial reload of ANF fuel in Grand Gulf Unit 1 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).
Changes from Cycle 2 to Cycle 3 for Grand Gulf Unit 1 include an additional reload of ANF fuel and extending the cycle from twelve to eighteen months.
The reload fuel for Cycle 3 is the same as Cycle 2 except for minor changes in enrichment and increasing the number of rods per bundle containing Gadolinia from five to either six or eight (Reference 3). These changes are of a minor nature, and are not expected to significantly change transient behavior from that calculated in Cycle 2 or to require a change in operating limits. The Cycle 3 transient analysis consists of recalculation of the limiting transients at the ME00 statepoints having the least margin to operating limits to confirm that the effects of the Cycle 3 changes on transient results are small and bounded by current limits. Confirmation of the limiting transients I
for Cycle 3 assures that the less limiting transients which were addressed for Cycle 2 will continue to be protected by the established operating limits.
l The objective of these analyses was to confirm the applicability of the Grand Gulf Unit 1 Cycle 2 Technical Specifications MCPR and MAPLHGR limits for Cycle 3 operation. Also, the Cycle 3 results should demonstrate that vessel integrity is protected during the most limiting pressurization event.
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~2.0
SUMMARY
The 'results 'of- the1 Grand Gulf, Unit 1 Cycle 3 transient analyses show that' all-
- currentz Cycle . 2 -Technical Specification thermal limits are,. applicable for Cycle 3 operation. 7 For.'~ the. two new reload batch designs - two new - MAPLHGR 111mits are included because of slight changes in local peaking. The MAPLHGR llimits need
- t'o be consistent with the LHGR limit'so that at reduced . power
- and/or reduced flow the 'LHGR limit - will be protected by the MAPFACf and MAPFAC p . multipliers on :MAPLHGR. 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 for .the -different ANF_ fuel 1 types.
, The Technical Specification limits for Cycle 3 operation are included in the
. reload analysis. report (Reference 3).
The Grand Gulf-Unitl1 analyses for Cycle 3 were performed at the power / flow
~ conditions found to be closest to thermal limits as shown in the Grand Gulf Unit 1 Cycle 2 transient analysis report (Reference 2). These power / flow I conditions are presented in Figure 2.1.
- The change-in the Critical Power Ratio (delta CPR) calculated for the limiting pressurization transient for Cycle 3 (Load Rejection Without Bypass) is presented in Table 2.1. Also presented in Table 2.1 are the results of ANF's
. generic Control Rod Withdrawal Error (CRWE) (Reference 4). The Grand' Gulf Unit 1 plant specific Loss of Feedwater Heating (LFWH) analyses results are also presented in Table 2.1. These results together with the Grand Gulf Unit 1 -Cycle 3 calculated safety limit MCPR-of 1.06 support continued c > of the Lexisting 1.18 MCPR operating limit (at rated conditions) for Cycle 3 operation.
-The plant transient and safety limit analyses results reported herein, together. with the results from Reference 4, support continued use of the Cycle 2 power dependent Minimum Critical Power Ratio (MCPRp ) and the power dependent Maximum Average Planar Linear Heat Generation Factor (MAPFAC p ) for Cycle 3
i 3 ANF-87-66 Revision 1 operation. The current MAPFAC p values and MCPR p limits and the results of ANF's analysis are presented in Figures 2.2 and 2.3 respectively. The results of ANF's_ generic CRWE analysis (Reference 4) and the existing MCPR p limits are presented in Figure 2.4. These results also support the continued use of the l Cycle 2 power dependent limits for Cycle 3 operation. l The flow dependent Minimum Critical Power Ratio (MCPR f ) and the results of ]
ANF's analysis are presented in Figure 2.5. The flow dependent Maximum i l
Average Planar Linear Heat Generation Rate Factor (MAPFAC f ) is presented in Figure 2.6. These flow dependent MAPFACf values and MCPRf limits are the same as those in the Grand Gulf Unit 1 Cycle 2 Technical Specifications and are applicable to Cycle 3 operation.
The results of the maximum system pressurization transient analysis are presented in Table 2.1. These results show that the Grand Gulf Unit I safety valves have sufficient capacity and performance to protect the established vessel pressure safety limit of 1375 psig during Cycle 3. .
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Thermal Limits g 1 Transient Delta CPR Losslof Feedwater Heating 0.11
> Control Rod Withdrawal Error 0.10 Feedwater Controller Failure (104.2/108)* 0.03' ~!
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. Load. Rejection Without Bypass (ME00 Conditions).
t .L Point % Power /% Core F1qw 1 104.2/108* 0.08 6 40/108 0.34
'10 25/73.8 1.04 11 25/40 0.81 l
Maximum System Pressurization 1
Transient % Power /% Core Flow Vessel Lower Plenum Steam D_gm.g i MSIV Closure 104.2/108* 1286 psia 1261 psia W 104.2/73.8 1285 psia 1269 psia I
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- 104.2% power /108% core flow is used for the Reload Licensing Analysis (RLA) conditions to conservatively bound 100% power /105% core flow.
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11 ANF-87-66 Revision 1 i
3.0 THERMAL LIMITS ANALYSIS l
3.1 Introduction The scope of the thermal limits analysis includes system transients, localized core events, and safety limit analysis in the determination of the MAPFACp values and the MCPR p limits. Results of flow excursion analyses were used to determine the MAPFACf values and the MCPRf limits.
COTRANSA (Reference 5), XCOBRA-T (Reference 6), and XTGBWR (Reference 7) are the major codes used in the thermal limits analyses as described in ANF's THERMEX Methodology Report (Reference 8). COTRANSA is a 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 CRWE, LFWH and flow excursion events.
3.2 System Transients The application of the Cycle 2 power dependent thermal limits to Grand Gulf Unit 1 Cycle 3 operation was confirmed. Figure 2.1 shows the power / flow conditions that were analyzed in support of the Cycle 2 reload. Five of these power / flow conditions were analyzed for Grand Gulf Unit 1 Cycle 3 using l
COTRANSA since they were found to be closest to thermal limits based on the l
Grand Gulf Unit 1 Cycle 2 results (Reference 2). The pressurization transient analysis was performed at Figure 2.1 power / flow statepoints 1, 6,10, and 11.
The feedwater controller failure analysis was performed at statepoint 1. ASME l pressurization analyses were performed at statepoints 1 and 3. Confirmatory loss of feedwater heater analyses were performed with XTGBWR at rated power and flow for four different exposures. The generic analysis for control rod withdrawal error is applicable to Grand Gulf Unit 1 Cycle 3. These analyses results show little change from the Cycle 2 ME00 analyses results due to Cycle
~
12 ANF-87-66 Revision 1
- 3. changes, thus justifying that the less limiting transients not analyzed for Cycle 3 will continue to be protected by the current Technical Specification limits.
3.2.1 Desian Basis The most limiting exposure for system transients in a cycle has been determined to be 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 conservative for cases where control rods are partially inserted. The analysis for the Grand Gulf Unit I with ME0D was performed
.using conservative analytical limits for trips and setpoints. Events initiated at core powers below 40 percent 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.
3.2.2 Anticipated 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 margins for Grand Gulf Unit 1 Cycle 3. Based on the results from the Grand Gulf Unit 1 Cycle 2 analysis (Reference 2), the limiting transients analyzed for Grand Gulf Unit 1 Cycle 3 were:
o Loss of Feedwater Heating (LFWH) o Load Rejection Without Bypass (LRNB) o Feedwater Controller Failure (FWCF)
Other transients are inherently non-limiting or bounded by one of the above as shown in the NSSS vendor ME00 analyses for Cycle 1 and the ANF Grand Gulf Unit 1 Cycle 2 analyses. Control Rod Withdrawal Error (CRWE) is an exception in that it has been analyzed generically.
13 ANF-87-66 Revisien 1 l
l .3.2.2.1 Loss Of Feedwater Heatina l
Analysis of the loss of feedwater heating event was performed to reflect reactor operation over the ME00 operating power versus flow map and conoitions anticipated during actual Grand Gulf reactor operation. This analysis is documented in Reference 9.
Additional calculations performed for Cycle 3 are shown in Table 3.1. This table provides the conditions of each case analyzed in terms of cycle exposure, core power, and core flow. The initial and final MCPR values and the associated' delta MCPR value are presented for each case. This
! demonstrates that the analysis presented for Cycle 2 is applicable to Cycle 3 operation as well.
3.2.2.2 Load Re.iection Without Bvoass The Load Rejection Without Bypass (LRNB) event is one of the most limiting of l' the class of transients characterized by rapid vessel pressurization for Grand Gulf Unit 1. The load 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 control valves as well as the recirculation pump high to low speed transfer. Condenser bypass flow, which can mitigate 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.
Figures 3.1 and 3.2 present the response of critical reactor and plant parameters to the LRWB event initiated at the Reload Licensed Analysis condition (RLA,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. These delta CPRs were calculated using XCOBRA-T, whereas previous i
4
14 ANF-87-66 Revision 1 cycle delta CPRs were calculated with the COTRANSA hot channel model and confirmed with XCOBRA-T.
3.2.2.3 Feedwater Controller Failure The failure of the feedwater controller to maximum demand (FWCF) is the most limiting of the vessel inventory increase transients. Failure of the feedwater control system to maximum demand would result in an increase in the s 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 to some extent. The core power excursion is terminated by the same mechanisms that end the LRNB transient.
Figures 3.3 and 3.4 present the response of critical reactor and plant parameters to the FWCF event initiated at the Reload Licensed Analysis condition (RLA, 104.2% power /108% core flow). The delta-CPR for this event was calculated using XCOBRA-T to be 0.03, indicating a MCPR operating limit requirement of 1.09 for the event. In support of the Cycle 2 reload, FWCF transients were also analyzed without condenser bypass and with a 100*F reduction in feedwater temperature. It was shown that these conditions had a minor impact on the delta-CPR and that significant margin exists to limits (Reference 2). Since the FWCF transient that was analyzed for Cycle 3 confirms the delta-CPR that was calculated for this transient for Cycle 2, it is not necessary to repeat the other FWCF transients for the Cycle 3 reload.
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15 ANF-87-66 .
Revision 1 l 1 3.2.2.4 Control Rod Withdrawal Error Reference 4 documents ANF's generic CRWE ' analysis for' Grand Gulf operating within the ME00. This generic analysis is applicable to Cycle 3. The results from Reference 4 and the Grand Gulf Technical Specification MCPR p limits are presented in Figure 2.4. These results demonstrate that the Grand Gulf Unit 1 p Technical Specification limits. bound the CRWE based MCPRp limits.
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MCPR 3.3 Flow Excursion Analysis
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The flow excursion transient is analyzed to determine the flow dependent thermal limits and values (MCPRf and MAPFACr). 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.
In the MCPRf analysis the power ascension associated with the flow increase that was determined to be conservative for the Cycle 2 analysis was confirmed to remain appropriate for the Cycle 3 analysis. The change in critical power for all fuel types along the ascension path 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 maxit..um flow. g Figure 2.5 presents the MCPRf limits for maximum flows of 102.5 and 107.0 percent rated.
Cycle 2 analysis of reduced flow LHGR limits was performed statistically based upon a wide variety of initial conditions. Calculations were performed for Cycle 3 at 0, 4, 7 and 10.7 GWd/MTV with XTGBWR to simulate the flow runup event from 40% of rated flow. The maximum percentage increase in the LHGR observed was 75% which is well below the 95/95 value of 83% from the Cycle 2 analysis. Figure 2.6 displays the MAPFACf values for the maximum flows of x 102.5 and 107.0 percent rated.
l 16 ANF-87-66 Revision 1 3.4 Safety Limit The safety limit MCPR is defined as the minimum value of the critical power ratio at which the fuel could be operated, with the expected number of rods in boiling transition not exceeding 0.1% of the fuel rods in the core. The safety limit is the minimum critical power ratio which would be permitted to occur during the limiting anticipated operational occurrence. The safety limit MCPR for all fuel types in Grand Gulf Unit 1 Cycle 3 operation was determined to be 1.06 using the methodology presented in Reference 10.
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 3 specific design basis radial and local power distributions are shown in Figures 3.5 to 3.9.
3.5 Results The results of the Grand Gulf Unit 1 Cycle 3 thermal limits analysis show that all current Cycle 2 thermal limits are applicable to Cycle 3 operation. A safety limit MCPR of 1.06 and an operating limit MCPR of 1.18 at rated conditions are supported.
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 part power conditions. The MCPR p limit is determined by adding the delta CPR for the limiting event to the calculated safety limit MCPR.
The power dependent MAPFAC (MAPFAC p ) is used to protect against both fuel melting and 1% clad strain during anticipated system transients from part power conditions. The conservative LHGR values for protection against fuel failure during anticipated operational occurrences are given in Reference 11.
l.
17 ANF-87-66 Revision 1 1
1 The results are then presented in a fractional form for application to the l MAPLHGR value.
The current Technical Specification MCPRp limits and MAPFACp values are shown to bound the results of ANF's analysis in Figures 2.2 and 2.3 respectively.
The results of ANF's generic RWE analysis (Reference 4) and the existing MCPRp limits are presented in Figure 2.4. These results support the continued use of the current Technical Specification power dependent limits for Cycle 3 operation.
3.5.2 Flow Deoendent Thermal limits And Values The flow- dependant MCPR limit (MCPR f ) protects against exceeding the safety limit MCPR for a flow excursion event. The results of the MCPRf analysis for Grand Gulf Unit 1 Cycle 3 are presented in Figure 2.5.
The flow dependent MAPFAC (MAPFAC f ) protects against both fuel melting and 1%
clad strain. The results presented in Section 3.3 demonstrate that the
, statistical analysis performed for Cycle 2 is applicable to Cycle 3. The Cycle 3 MAPFACf values are presented in Figure 2.6.
These flow dependent MAPFACf values and MCPRf limits are the same as those presented in the current Grand Gulf Unit 1 Technical Specifications.
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24 ANF-87-66 Revision 1 1
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25 ANF-87-66 .
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26 ANF-87-66 Revir. ion 1
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' ~ 27 - ANF.87-66
-. Revision 1-
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28 ANF 87-66 Revision 1
. I i
4.0 MAXIMUM OVERPRESSURIZATIbN. -
l I 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 Pressur.i Nssel Code. This-analysis showed ;
that the Grand Gulf Unit 1 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 system pressures at the most limiting power / flow point (104.2 percent power /108 percent flow) tre shown in Table 2.1.
4.1 Desian Basis _
i During the transient, the most critical active component (direct scram on MSIV closure) was assumed to fail. The event is terminated by the high flux scram.
Credit is 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.
4.2 Maximum Pressurization Transients Scoping analysis descr;t,ca 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 when all transients are evaluated on the same basis (without direct scram) because of the smaller steam line volume associated with MSIV closure. Though the closure rate of the MSIVs is substantially slower than turbine stop or control valves, the compressibility 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 29 ANF-87-66
}
Revision l' 4.3 Results
)
The results of the maximum system pressurization analysis are presented in 1
, Table 2.1. Figures 4.1 ar.d 4.2 present the respense of critical reactor and l plant parameters during the MSIV closure event from 104.2% power /108% flow.
These results show that the Grand Gulf Unit I safety valves have sufficient capacity and performance to protect the previously established vessel pressure safety limit for Cycle 3.
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32' ANF-87 Revision 1 a
1 5.0 . REFERENCES-l 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," dated August 15, 1986.
L
- 2. 1" Grand Gulf -. Unit 1 Cycle '2 Plant Transient Analysis," XN-NF-86-36,-
Revision 3,. Exxon Nuclear Compt ',, Inc., Richland, WA, August 1986. 4
- 3. " Grand Gulf Unit 1 Cycle 3 . Reload Analysis," ANF-87-67, - Revision 1, Advanced Nuclear Fuels Corporation, Richland, WA, August 1987.
' 4. "BWR/6 Generic Rod Withdrawal . Error Analysis; MCPRo for Plant.0perations within the Extended Operation Domain," XN-NF-825(P)(A), Supplement 2,-
Exxon Nuclear Company, Inc., Richland, WA, October 1986.
- 5. " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactor,"
XN NF-79-71(P), Revision 2, Supplements 1, 2& 3(A), Exxon Nuclear Company,. Inc., Richland, WA, November 1981.
h 6. "XCOBRA-T: . A - Computer Code for BWR Transient Thermal Hyd raul i c ' . Core -
Analysis," E 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 Method l
for. Design -and Analysis," XN-NF-80-19(A), Volume 1, Exxon Nuclear Company, Inc., Richland, WA, March 1983.
- 8. " Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Thermal-Limits Methodology Summary Description," XN-NF-80-19(P)(A), Volume 3, Revision 2, Exxon Nuclear Company, Inc., Richland, WA, January 1987.
- 9. Letter, O. D. Kingsley (MP&L) to H. Denton (NRC), " Loss of Feedwater Heating Analysis Using XTGBWR," dated September 5, 1986. j
- 10. " Exxon Nuclear Critical Power Methodology for Boiling Water Reactor," XN-NF-524(A), Revision 1, Exxon Nuclear Company, Inc., Richland, WA, j November 1979. j
- 11. " 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, j September 1986. 1 l
l J
i l
l !
,- . . -- - . - - ~ _ .
1 c;;v ; -
n ANF-87-66'
-Revision 1 <
' Issue Date: 8/10/87 ~
{. .
GRAND GULF UNIT 1 CYCLE 3 PLANT TRANSIENT ANALYSIS-Distribution M. J. Ades
-D.'A.-Adkisson R.'E..Co11ingham-L.-J. Federico J.' D. Floyd K.- P. Galbraith B. J. Gitnick (ENSA)
R. G.'Grummer R..L. Gulley
. J. G. Ingl.am S. E. Jensen
, T. L. Krysinski b
J. N. Morgan L- C. J. Volmer H..E. Williamson SERI/J. D. Floyd (40) .
Document Control (5)
- - - - _ - _ _- = . _ . - -