ML20207G761

Rev 2 to Grand Gulf Unit 1 Cycle 2 Plant Transient Analysis

ML20207G761
Person / Time
Site:	Grand Gulf
Issue date:	06/30/1986
From:	Collingham R, Morgan J, Ward G SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML20207G733	List: AECM-86-0214, Application for Amend to License NPF-29,consisting of Proposed Change 86-15,revising Tech Specs Re Cycle 2 Reload. Fee Paid ML20207G736 ML20207G753 ML20207G761 ML20207G773
References
TAC-61930, XN-NF-86-36, XN-NF-86-36-R02, XN-NF-86-36-R2, NUDOCS 8607230161
Download: ML20207G761 (72)

Text

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' F-1 XN-NF-86-35 .

Revision 2 -

APPENDIX F COMBINED LOCA-SEISMIC EVALUATION t

The structural response of the ENC XN-18x8 fuel is the same as the structural response of the GE6 8x8 fuel it replaces in the Grand Gulf Unit I core.

Therefore, the LOCA-seismic structural response evaluation performed in support of the initial core remains applicable and continues to provide '

assurance that control blade insertion will not be inhibited following the ,

occurrence of the design basis LOCA-seismic event.

l The physical and geometric properties of the ENC and GE 8x8 fuel types are summarized in Table F-1. The close agreement between the important parameters f for the two fuel types indicates that the structural response would be very i similar for both fuel types.

4 Similarity in the natural frequencies of the two fuel types is further assured by the stiffness of the fuel assembly channel box. Both fuel types use the

same fuel assembly channel box, and the channel box dominates the overall dynamic response of the incore fuel. ENC calculations show that approximately 97% of the stiffness of a BWR fuel assembly is attributable to the stiffness of the channel box. For this reason, the dynamic structural response of the reload core is essentially that of the initial core, and the original LOCA-seismic analysis remains applicable. Deformation of the channel to the i point that control blade insertion is iahibited is not predicted to occur.

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F-2 XN-NF-86-35 '_

Revision 2 TABLE F-1 g

COMPARISON OF PHYSICAL.AND GEOMETRIC CHARACTERISTICS 3 0F ENC AND GE 8X8 ASSEMBLIES I

PROPERTY ENC FUEL GE FUEL Fuel rod pitch, inch 0.636 0.636 Number of spacers 7 7 '

Assembly weight, lb 596 600 -

Assembly length, inch 176.16 176.16 [

Fuel rod diameter, inch 0.484 0.483 m Cladding thickness, inch 0.035 0.032 I

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I* XN-NF-86-35 Revision 2 Issue Date: 6/30/86 GRAND GULF UNIT 1 CYCLE 2 RELOAD ANALYSIS l ,

Distribution I

J.C. Chandler R.E. Collingham i T.P. Currie .

R.A. Decker #

J.D. Floyd '

I S.F. Gaines K.P. Galbraith R.G. Grummer R.L. Gulley i

T.H. Keheley l S.E. Jensen j

T.L. Krysinski J.N. Morgan L.A. Nielsen

, A. Reparaz i G.L. Ritter i G.N. Ward H.E. Williamson I

MP&L/JD Floyd (40)

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L f G9A\lD GOLF ONIT 1LCYCLE 2 i PLANT TRANSIENT ANALYSIS i "

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XN-NF-86-36 Revision 2 l Issue Date: 6/30/86 i

GRAND GULF UNIT 1 CYCLE 2 PLANT TRANSIENT ANALYSIS

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. i Prepare: B.J. Gitnick and J.Q. Chandler BWR Safety Analys

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j Concur: #I .4*c, lejr

R.E. Collingffam, Manager BWR Safety Analysis Concur

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G.N. Ward, Manager Reload Licensing Concur: M MA/b6 a C

' J.jef Morgan, Manager '

Customer Services Engineering i e Approve: , h l

H.E.'WIITi'&mson, Manager Licensing and Safety Engineering Approve: . I. (,/30/14 I

G.L. Ritter,' Manager '

Fuel Engineering and Technical Services i

tare 1

l ERON NUCLEAR COV 3ANY l\C.

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I i XN-NF-86-36 Revision 2 -

TABLE OF CONTENTS SECTION PEE I

1.0 INTRODUCTION

.......................................... 1 4 2.0 5UNHARY............................................... 3 1

3.0 ANALYTICAL METHODS AND COMPUTER MODELS................. 12 l

3.1 THERMEX Thermal Limi ts Methodol ogy. . . . . . . . . . . . . . . . . . . . . 12 3.2 COTRANSA System Mode 1.................................. 12 3.2.1 Plant-Specific Modifications to C0TRANSA............... 12 3.2.2 Treatment of Uncertainties in Data..................... 13 I

3.3 Cri tical Power Methodol ogy. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 4 1 4 4.0 MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT.............. 15 4.1 Design Basis Power Distribution........................ 15 4.2 Calculation Of The Number Of Rods In Boiling Transition.................................. 16 i

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5.0 GENERIC TRANSIENT ANALYSES APPLICABLE TO GRAND GULF.... 20 5.1 Loss Of Feedwater Heating.............................. 20 i

5.2 Control Rod Wi thdrawal Error. . . . . . . . . . . . . . . . . . . . . . . . . . . 20 l 6.0 TRANSIENT ANALYSES FOR THERMAL MARGIN.................. 23 1

6.1 Design Bases For Thermal Margins. . . . . . . . . . . . . . . . . . . . . . . 23 6.2 Analysis Of Plant Transients At Rated Conditions....... 25 6.2.1 Load Rejection Without Bypass.......................... 25 6.2.2 Feedwater Controller Failure........................... 26

6.3 Analysis At Reduced Flow Operating Conditions.......... 27 6.4 Analyses At Reduced Power Operating Conditions......... 28 6.5 Operation At Reduced Feedweter Temperature............. 28 l

L 11 XN-NF-86-36 Revision 2 _

TABLE OF CONTENTS (Continued)

SECTION P,AAGE 7.0 ANALYSIS OF EXTENDED OPERATING CONDITIONS.............. 35 7.1 Design Bases For Thermal Margins. . . . . . . . . . . . . . . . . . . . . . . 35 7.2 Determination Of Analytical Statepoints. . . . . . . . . . . . . . . . 35 7.3 Analysis Of P'. ant Transients Under ELL Conditions...... 36 7.4 Analysis Of Plant Transients Under ICF Conditions...... 37 7.5 Validation Of Off-Nominal MCPR Limits. . . . . . . . . . . . . . . . . . 37 -

8.0 ASME OVERPRESSURIZATION ANALYSIS....................... 52 8.1 Design Basis........................................... 52 8.2 MSIV Cl o s ure An alys i s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 52 8.3 Steam Dome Pressure Safety Limit....................... 53 f

9.0 SPECIFICATION OF OPERATING LIMITS...................... 57 9.1 MCPR Operating Limits.................................. 57 9.1.1 Flow-Dependent MCPR Limi t. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 g 9.1.2 Power Dependent MCPR Limit............................. 58

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9.2 MAPLHGR Operating Limits............................... 58 9.2.1 Flow-Dependent MAPLHGR Limi t Mul tiplier. . . . . . . . . . . . . . . . 58 9.2.2 Power-Dependent MAPLHGR Limit Multiplier............... 59

10.0 REFERENCES

............................................. 65 I .

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Revision 2 -

LIST OF TABLES

.TMLE .T.11LL EME 2.1 Summary of Transient Analysi s Resul ts. . . . . . . . . . . . . . . . . . 5 2.2 Overpressurization Analysi s Resul ts. . . . . . . . . . . . . . . . . . . . 6 6.1 Reactor and Plant Conditions, Transient Analyses Supporting Operation at Rated Conditions............... 29 J

6.2 Resul ts of Transient Analyses at Rated Flow. . . . . . . . . . . . 30 6.3 Resul ts of MCPR( f) Analys is. . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 6.4 Results of Transient Analyses with Feedwater Heaters

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Ou t o f Se rv i ce . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 i -

7.1 Analytical Statepoints for ELL and ICF Evaluation...... 39 7.2 Results of Transient Analyses at ELL and ICF Statepoints................................ 40 8.1 Reactor and Plant Conditions, ASME l Overpressurization Analysis............................ 54

. 8.2 Results of ASME Overpressurization Analysis. . . . . . . . . . . . 55 9.1 MAPFAC(p) Analysi s Resu l ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 60 I

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Revision 2 -

LIST OF FIGURES FIGURE IIILE EAGE 2.1 Operati ng Power-Fl ow Map. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 1 2.2 Flow-Dependent MCPR Limit............................. 8 2.3 Power-Dependent MCPR L1mit............................ 9 2.4 Flow-Dependent MAPLHGR Factor......................... 10 2.5 Power-Dependent MAPLHGR Factor........................ 11 4.1 Design Basis Radial Power Distribution. . . . . . . . . . . . . . . . 17 4.2 Design Basis Local Power Distribution, ENC XN-1 8x8 Fuel..................................... 18 l 4.3 Design Basis Local Power Distribution, GE 8x8 Fuel.... 19 '

5.1 Power-Dependent MCPR Limit from Generic CRWE Analysis................................. 22 6.1 System Traces...................................... 33 6.2 System Traces......................................... ... 34 7.1 Operati ng Power-Fl ow Map . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 1 7.2 System Traces......................................... 42 7.3 System Traces......................................... 43 7.4 System Traces......................................... 44 7.5 System Traces......................................... 45 7.6 System Traces......................................... 46 7.7 System Traces......................................... 47 7.8 System Traces......................................... 48 7.9 System Traces......................................... 49 7.10 Flow-Dependent MCPR Limit Validation.................. 50 7.11 Power-Dependent MCPR Limi t Val idation. . . . . . . . . . . . . . . . . 51 l 8.1 System Traces, ASME Overpressurication Analysis....... 56 9.1 Fl ow-Dependent MCPR Limi t. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 61 l 9.2 Power-Dependent MCPR Limit............................ 62 9.3 Fl ow-Dependent MAPLHGR Factor. . . . . . . . . . . . . . . . . . . . . . . . . 63 9.4 Power-Dependent MAPLHGR Factor........................ 64 l

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Revision 2 ACKNOWLEDGEMENT i

The following individuals made significant contributions to the effort reported in this document:

S.E. Jensen g T.H. Keheley 7

L.A. Nielsen .

8.J. Gitnick (ENSA) l I

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1 XN-NF-86-36 Revision 2 -

1.0 INTRODUCTION

I This report describes the plant system transient analyses performed by Exxon g Nuclear Co., Inc., in support of the Cycle 2 (XN-1) reload for Grand Gulf Unit i 1. This cycle is scheduled to commence operation in Fall 1986.

l Cycle 2 is the first cycle during which the Grand Gulf Unit I core will contain ENC 8x8 fuel. In addition to the ENC 8x8 fuel, the Cycle 2 core will contain a significant number of 8x8 assemblies fabricated by General Electric.

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Operating limits for these fuel types during Cycle 2 operation are established in this report.

l In order to improve plant operational flexibility, transient analyses were performed to support operation in extended regions of the operating power flow ,

map submitted during Cycle 1. Transient evaluations cover operation in the Extended Load Line (ELL) and Increased Core Flow (ICF) regions and operation with one or more feedwater heaters out of service.

1 In addition to the analysis of system transients, this document also contains

, the results of antlyses for the MCPR Fuel Cladding Integrity Safety Limit and l the Maximum Overpressurization Accident for qualification in compliance with the ASME code.

l The combination of these analytical results into a framework of operating ll limits is discussed ir. Section 9.0. In addition to the thermal margin limits I

determined from the plant transient analyses, this section also addresses l, power- and flow-dependent factors for modification of the MAPLHGR limits when operating at other than rated power and flow conditions. In general, i

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2 XN-NF-86-36 '

l l Revision 2 l

1 l off-nominal operating limits have been retained from the Cycle 1 analyses performed by the NSSS supplier (Ref. 9) except in instances where the ENC analyses indicated a need for more restrictive limits.

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2.0 SUfmARY l This report provides the results of the simulations of the limiting plant g transients for Cycle 2 rperation of Grand Gulf Unit 1. These transients are a the generator load rejection with concurrent failure of the condenser bypass system (LRNB) and the failure of the feedwater flow control system to maximum l demand (FWCF). Results of detailed transient analyses performed for determination of operating margin requirements are summarized in Table 2.1.

These analytical results were combined to support the operating limitations determined in Section 9.0 using the operating power-flow map shown in Figure 2.1.

The Loss of Feedwater Heating (LFWH) transient and the Control Rod Withdrawal Error (CRWE) transient, which are covered in generic technical reports described in Section 5.0, require nominal MCPR operating limits of at least

, 1.14 and 1.16, respectively. The Fuel Loading Error (FLE) event, which is reported in the reload analysis (Ref.12), requires a nominal MCPR limit of at least 1.17.

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The flow-dependent MCPR limits for Cycle 2 operation of Grand Gulf Unit I are shown in Figure 2.2. These limits were retained from Cycle 1 and validated for all fuel types by the conservative analysis of a hypothetical recirculation flow increase transient. The Rated Conditions MCPR operating limit, which is to be retained from the Cycle 1 operating limits, appears as

part of the flow-dependent MCPR limit.

, The power-dependent MCPR limit for Cycle 2 operation of Grand Gulf Unit 1 is shown in Figure 2.3. This limit was retained from Cycle 1 and validated for continued use during Cycle 2 for all fuel types by the analyses reported in this document.

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d 4 XN-NF-86-36 Revision 2 The flow-dependent MAPLHGR limit factors for Cycle 2 operation of Grand Gulf I

l Unit I are shown in Figure 2.4. These limit factors were retained from Cycle

! I for GE fuel and calculated separately for ENC fuel.

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l l The power-dependent MAPLHGR limit factor for Cycle 2 operation of Grand Gulf I l Unit 1 is shown in Figure 2.5. This limit factor was also retained from Cycle j 1 and validated for continued use during Cycle 2 for all fuel types by the l analyses reported in this document.

The results of the maximum overpressurization accident analysis performed for

• compliance with the ASME code are shown in Table 2.2. These results g

demonstrate that design pressure margins are protected during the postulated worst overpressurization event in the plant design basis with seven g

safety / relief valves out of service. A The system transient analyses documented in this report were performed using ENC's COTRANSA computer code (Ref. 2), which includes a one-dimensional representation of the core for evaluation of the axial power shape behavior l during transient events. The delta CPR's obtained for the LRWB and FWCF transients for analyses to date were obtained with the updated hot channel g

delta CPR model in COTRANSA. NRC has informed ENC that this model is a l

unreviewed. As a result, ENC is reanalyzing the more limiting cases using the

'XCOBRA-T code (Ref. 15) which is currently under review by the NRC. The results from the delta CPR model of Reference 2 is to be used as a figure of l merit to aid in establishing the more limiting cases.

The limits that may be affected by this reanalysis include the MPCR operating y limit and MCPRp limit. The other limits as defined in the report should be A.

unaffected. This report will be revised to include detail results when these runs are completed. Sections 2 and 7, Tables 2.1, 6.2, 6.4, 7.1, 7.2, and 9.1, and Figures 6.1, 6.2, 7.1 through 7.11, and 9.4 will be updated to f

include these detail results.

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Revision 2 TABLE 2.1

'. SIMiARY OF A TRANSIENT ANALYSIS RESULTS INIT INIT PEAK PEAK PEAK DELTA EVENT F0WER FLOW POWER HT FLX PRESSURE CPR MODEL M ltid M 21d osia ANALYSES FOR RATED MCPR LIMIT l

TO BE SUPPLIED LATER l

1 ANALYSES FOR EXTENDED LOAD LINE OPERATION TO BE SUPPLIED LATER I

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i ANALYSES FOR INCREASED CORE FLOW OPERATION TO BE SUPPLIED LATER s

Peak heat flux value near time of minimum delta CPR.

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TA8LE 2.2 OVERPRESSURIZATION ANALYSIS RESULTS E

73.8% FLOW CASE 100% FLOW CASE 108% FLOW CASE Maximum Flux, % 384.9 336.4 357.6 Maximum Pressure l in Steam Dome, psia 1280 1264 1260 .- 5 Maximum Pressure in Lower Plenum, psia 1296 1288 1288 Maximum Pressure in Steamlines {

at S/R Valves, psia 1268 1252 1249 5 I

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(73.8,100) (105,100) j 100-l ELL Region . r-ICF l A g. / Region i

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CORE FLOW, % OF RATED l

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'GE FUEL 0.9 MAX FLOW = 102.5%

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Revision 2 -

3.0 ANALYTICAL METHODS AND COMPUTER MODELS This section describes the analytical methods used in the analysis of anticipated transient conditions for operation of the Grand Gulf Unit 1 i, reactor during Cycle 2.

3.1 THERMEX Thermal Limits Methodoloav i' Exxon Nuclear's THERMEX thermal limits methodology is described in Reference

1. This topical report sumarizes the analytical methods used for thermal

• l hydraulic design analysis and their interaction in the formulation of MCPR limits for the operating cycle.

3.2 COTRANSA System Model The COTRANSA model is described in Reference 2. COTRANSA is a combination of the system model developed for PTSBWR3, which incorporates steamline dynamics, g and the COTRAN core kinetics model, which includes a one-dimensional neutronic I representation of the core. Calculational uncertainties are bounded in the

analysis using the methods described in Reference 8.

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, 3.2.1 Plant-Snecific Modifications to COTRANSA l

Minor model changes. to COTRANSA were necessitated by plant configuration lt features unique to Grand Gulf Unit 1, such as the performance characteristics l

of the turbine control valve and modeling of the BWR/6 safety grade high water level scram.

l The adequacy of the Grand Gulf COTRANSA model was verified through prediction of feedwater and pressure controller tests and plant startup tests for recirculation pump coastdown and load rejection with bypass. The recirculation

l. pump coastdown input bounded by test data were used for analysis of the recirculation pump flow coastdown following trip. COTRANSA conservatively 1

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13 XN-NF-86-36 Revision 2 predicted the reactor power history in the load rejection startup test, I

confirming the adequacy of the Grand Gulf COTRANSA model for license basis analyses.

3.2.2 Treatment of Uncertainties in the Data For the evaluation of the system transients covered in this analysis, plant variables were considered to be at conservative values. For variables covered l

by the Technical Specifications, such as scram insertion speed and delay time, the technical specification limits were used in the license basis analysis.

• Calculational uncertainties associated with COTRANSA were bounded in the analysis through the use of a 10% uncertainty factor applied to the integral of the neutron flux. This procedure is in accordance with the treatment of uncertainties described in Reference 8.

3.3 Critical Power Methodology The operating critical power ratio (CPR) is calculated using the XN-3 critical power correlation (Ref. 4) and the analytical formulation described in Reference 3. This reference also describes the Monte Carlo procedure by which ENC calculates the MCPR Fuel Cladding Integrity Safety Limit.

In the ENC MCPR safety limit methodology, plant measurement uncertainties are combined with power distribution measurement uncertainties and the uncertainties inherent in the XN-3 prediction of critical heat flux phenomena using a Monte Carlo procedure and non-parametric tolerance limits. The safety limit MCPR is defined on the basis of local and radial power distributions representative of the operating cycle such that during sustained, steady state operation at the MCPR safety limit, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition.

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i The COTRANSA model used in Grand Gulf analyses used the hot channel delta CPR model of Reference 2 as a figure of merit to determine the more limiting system transients and flow conditions. The delta CPR for the more limiting i

transients and conditions are calculated with XCOBRA-T (Ref.15).

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, Revision 2 4.0 MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT

The MCPR fuel cladding integrity safety limit was calculated using the methodology and uncertainties described in Reference 3. In this methodology, a Monte Carlo procedure is used to evaluate plant measurement and power prediction uncertainties such that during sustained operation at the MCPR Fuel Cladding Integrity Safety Limit, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition. This section describes the I calculation and presents the analytical results. ~

During sustained operation at a MCPR of 1.06 with the design basis power distribution described below, at least 99.9% of the fuel rods in the core are expected to avoid boiling transition at a confidence level of 95%.

4.1 Desian Basis Power Distribution Predicted power distributions were extracted from the fuel management analysis for Grand Gulf Unit 1 Cycle 2. These radial power distributions were evaluated for perfonsance as the design basis radial power map, and the distribution at L 6000 MWD /MT cycle exposure was conservatively selected as the most severe 1

expected distribution for the cycle. The distribution was skewed toward higher power factors by the addition of bundles with a radial peaking factor f approximating an operating MCPR level of 1.17 at full power.

l The resulting design basis radial power distribution is shown in Figure 4.1.

, The fuel management analysis indicated that the maximum power ENC bundle in i the core at this statepoint was predicted to be operating at an exposure level of 7594 MWD /MT,. so a local power distribution typical of a nodal exposure of

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8000 MWD /NT was selected as the design basis local power distribution. This distribution is shown in Figure 4.2.

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16 XN-NF-86-36 Revision 2 A bounding (flat) local power distribution was selected for the coresident G.E. fuel. This distribution is shown in Figure 4.3.

Because the predicted power distributions during the cycle were all characterized by bottom peaked axial distributions, a center peaked cosine distribution was selected as conservative for purposes of the safety limit analysis.

l 4.2 Calculation Of The Number Of Rods In Boilino Transition The computer models described in Reference 3 were used to analyze the number of fuel rods in boiling transition. The XN-3 correlation (Ref. 4) was used to predict critical heat flux phenomena.1000 Monte Carlo trials were performed using the uncertainties identified in Reference 4. Consistent with Reference l

3, non-parametric tolerance limits (Ref. 5) were used in lieu of Pearson curve fitting.

A total of 2000 Monte Carlo trials were run in confirmation of the MCPR safety limit. At 1000 trials, 0.07% of the fuel rods in the core are expected to experience boiling transition with a confidence level of 95%. This result satisfies the criterion defining the MCPR safety limit.

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1.02 : 0.97 : 1.02 :

1.06 : 1.07 :

1.01 : 0.00 : 0.87 : 0.99 : 1.06 : 1.06 :

1.06 : 1.06 : 0.99 : 0.87 : 0.00 : 1.00 : 1.07 : 1.06 :

1.02 : 0.97 : 1.02 : 0.99 : 1.00 : 1.03 : 0.87 : 1.02 :

0.95 : 0.94 : 0.97 :: 1.06 :: 1.07 :: 0.87 :: 1.05 :: 0.95 ::

1.04 : 0.95 : 1.02 :: 1.06 :: 1.06 :: 1.02 :: 0.95 :: 1.04 :: I

......................................................................... [

FIGURE 4.2 f DESIGN BASIS LOCAL POWER DISTRIBUTION ENC XN-1 8X8 FUEL -

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• Fuel rod adjacent to control blade position

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I 10 XN-NF-86-36 Revision 2 f .

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1  : 0.99 : 0.99 : 1.02 :: 1.01 :: 1.02 :: 0.91 :: 1.03 :: 0.99 ::

I  :  :  :  :  :

0.99 : 1.02 : 1.01 : :0.91 :: 0.00 :: 1.02 : : 1.02 : :0.99 :  :

0.99 : 1.03 : 1.02 : 0.00 : 1.02 : 1.01 : 0.99 : 0.99 :

!I  :  :  :

0.99 : 1.03 : 0.91 : 1.02 : 1.01 : 0.98 : 0.99 : 0.99 :

,  : 1.00 : 0.99 : 1.03 :: 1.02 :: 0.99 :: 0.99 :: 0.97 :: 1.00 :

l .........................................................................

l

1.03 : 1.00 : 0.99 :: 0.99 :: 0.99 :: 0.99 :: 1.00 :

1.03 :

r FIGURE 4.3 DESIGN BASIS LOCAL POWER DISTRIBUTION G.E. 8X8 FUEL

• Fuel rod adjacent to control blade location 1

l

i 1

20 XN-NF-86-36 -

Revision 2 i 5.0 GENERIC TRANSIENT ANALYSES APPLICABLE TO GRAND GULF This section identifies the generic transient analyses which have been a submitted as topical reports by ENC' and are applicable to the Grand Gulf reload.

I 5.1 Loss Of Feedwater Heatina The Loss of Feedwater Heating (LFWH) transient has been analyzed on a generic .-

basis for a wide cross section of BWR configurations. This generic analysis is documented in Reference 6.

l* The Reference 6 analysis provides a statistical evaluation of the consequences of the LFWH transient for BWR/4, BWR/5, and BWR/6 plant configurations under conditions which cover the normal operating power flow map and the Extended j Load Line (ELL) and Increased Core Flow (ICF) regions.

The generic conclusions support a MCPR operating limit of at least 1.14 for l plants with a MCPR safety limit of 1.06. As noted in Section 4.0 of this report, the Grand Gulf Unit 1 MCPR safety limit is 1.06; hence the LFWH transient requires a MCPR operating limit of 1.14 or greater for Grand Gulf, f 5.2 Control Rod Withdrawal Error l! The Control Rod Withdrawal Error (CRWE) transient has been analyzed by ENC on l a generic basis for BWR/6 plants. This generic analysis is documented in l

, Reference 7.

The Reference 7 analysis provides a statistical evaluation of the consequences of the CRWE transient for 8WR/6 plant configurations under conditions which cover the normal operating power flow map and the ELL and ICF regions.

l l

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21 XN-NF-85-36 Revision 2 The generic conclusions support a power-dependent MCPR limit function as shown in Figure 5'.1. This limit was considered in determining the power-dependent MCPR limits documented in Section 9.0 of this report.

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I-1 23 XN-NF-86-36

, Revision 2 _

!i 6.0 TRANSIENT ANALYSES FOR THERMAL MARGIN 1his section covers the determination of the rated MCPR operating limit and the off-nominal MCPR limits associated with the normal power-flow map. Limits

!u for off-nominal conditions in the extended regions of the power-flow map are confirmed by analyses in separate report sections.

i 6.1 Desian Bases For Thermal Marains I Eight categories of anticipated operational occurrences are identified in ,.

. XN-NF-79-71(P) (Ref. 2) as potentially limiting system transients in terms of

. thermal margins. Of these categories, all but rapid pressurization, increase in vessel coolant inventory, and reduction in core inlet enthalpy (cold water transients) are identified as either inherently self-limiting or bounded by one of the other classes. Reactivity anomalies, which are also potentially 1 limiting events, are covered in the reload analysis report (Ref.12).

{

Transient analyses are performed for the limiting pressurization, coolant 4

inventory increase, and inlet enthalpy decrease events identified in the FSAR to determine operating thermal margin requirements for the fuel types in the Grand Gulf Unit I core during Cycle 2 operation. Thermal margin requirements are specified as limits on the Minimum Critical Power Ratio (MCPR) as j calculated using the XN-3 Critical Power Correlation (Ref. 4).

Analyses in this section address rapid pressurization events and vessel

! inventory increase events; generic analyses for cold water transients and one of the reactivity anomaly events are referenced in the determination of l

operating limits.

Flow dependent MCPR operating limits are established by evaluating a flow increase transient with a resulting power increase which continues until the physical maximum performance of the system is reached. These limits assure that the MCPR safety limit is protected following such an event.

24 XN-NF-86-36 ,

Revision 2

.I Power-dependent MCPR operating limits have been established generically for BWR/6 plants with a statistical evaluation of the control rod withdrawal error transient. The generic limit function is verified through analysis of the feedwater controller failure and generator load rejection transients at full flow and reduced power conditions and at additional points within the operating power-flow map. If necessary, the generic limits are adjusted upward to accommodate the results of the additional analyses.

For transient events whose consequences are dependent on the rapid insertion of control rods, the limiting point in the operating cycle has been determined f to be at the end of normal full power capability, at which time all of the control rods are fully withdrawn from the core. With the control rods fully '

withdrawn from the core, a longar control movement is required before the g

blade tips reach the axial level at which a significant reactivity effect is L realized. This effective scram delay makes the end of cycle exposure point the most severe in terms of thermal margin consequences for scram-sensitive transients such as LRNB and FWCF.

f In the Grand Gulf plant, events which are sensitive to scram performance exhibit relatively benign consequences throughout the operating cycle. In the g transient analyses supporting operation at rated conditions, scram performance L was assumed at the minimum acceptance values cited in the Technical Specifications.

f An additional analysis is provided to support plant operation with reduced feedwater temperature. Results of the LRNB and FWCF transients are used to demonstrate the continued applicability of the nominal MCPR operating limit at y l feedwater temperature values up to 100 degrees F lower than the nominal value. R l

The MCPR operating limit is determined by comparing the thermal margir, l requirements of all the anticipated events considered in the design basis. The f

operating thermal margin requirements for each anticipated event are determined from existing generic analyses and from plant specific analyses. {

E

25 XN-NF-86-36 4

Revision 2 _

l For rapid transients, such as those analyzed in this report, thermal margin requirements are dictated by the change in thermal margin (delta-MCPR) observed during the transient. In other events, the thermal margin q requirements are established as operating MCPR values directly. For comparison purposes, delta-CPR values are converted to MCPR limit requirements by adding the MCPR safety limit value.

Administration of the operating limits established by these analyses is l dependent on the operating power-flow state. At other than rated conditions, the power- and flow-dependent MCPR limit functions provide limit values, which are compared against the normal conditions limit. The highest of these.three yklues is the MCPR operating limit for the operating power-flow statepoint.

I 6.2 Analysis Of Plant Transients At Rated Conditions i Analyses were performed at analytical statepoints selected to protect allowed operation of the plant at up to rated power and flow conditions. Initial e

conditions for these analyses assumed a power level corresponding to 104.2% of

! rated thermal power. For analyses at extended power-flow operating points

, (ELL and ICF), refer to Sections 7.0 of this report.

4 Major input variables used in the analyses supporting operation at rated I conditions are listed in Table 6.1. Major analytical results are given 'in the first two lines of Table 6.2.

'1 6.2.1 Load Re.iection Without Bvoass The generator load rejection transient without bypass to the condenser (LRNB) is the most severe transient in the category of events characterized by rapid vessel pressurization. Other events in this category include turbine trip and containment isolation.

I i

.l 26 XN-NF-86-36 Revision 2

.I The transient is initiated by sudden rejection of the turbine generator output. The turbine controller commands a rapid closure of the turbine control valve, which in turn causes a scram trip, a recirculation pump trip, and a l

pressure wave in the steamlines. a R

The controller also generates a signal to open the steam bypass valves, but there is a comon mode failure which could disable the bypass system while triggering a rapid pressurization transient. The bypass is therefore assumed to fail before any of the valves open, allowing the pressure wave to continue up the steamlines unattenuated and to pressurize the core. ,.

l l

l When the pressure wave reaches the core, the core void level is reduced by the I increased pressure. The lower void inventory increases the reactivity, and the power rises.

The power excursion is terminated by the effects of control rod insertion, increased voiding caused by recirculation pump trip, and increased voiding

{

caused by direct moderator heating. Void effects associated with the increased heat flux also have a negative reactivity effect but are not realized until after the event is over. The rapid scram implemented by Grand Gulf limits most fast transients to relatively benign consequences.

The analytical results of the LRNB calculations are shown 'a Figure 6.1, which depicts the time variance of important reactor and plant parameters during the f

LRNB transient.

6.2.2 Feedwater Controller Failure The failure of the feedwater controller to maximum demand (FWCF) is the most limiting of the vessel inventory increase transients.

(

l I

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27 XN-NF-86-36 Revision 2 -

Failure of the feedwater control system to maximum demand would result in an increase in the 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 soiae extent.

The core power excursion is terminated by the same mechanisms that end the LRNB transient.

Important core and system parameters are plotted against event time in Figure 6.2.

6.3 Analyses At Reduced Flow Operatina Conditions The hypothetical failure of the recirculation flow control system such that the recirculation flow increases slowly to the physical maximum attainable by the equipment was evaluated. The power ascension associated with the flow increase was conservatively taken as the bounding ascension from a series of XTGBWR analyses of flow excursion events.

The thermal hydraulic conditions of the core were calculated by heat balance at several points along the assumed power ascension line. The change in critical power for all fuel types.along the ascension path was calculated with '

XCOBRA (Reference 1). Peaking factors were selected such that the bundle with the least margin would reach the MCPR safety limit of 1.06 at 108% of rated flow, which was conservatively assumed to be the maximum capability of the recirculation system at high pump speed.

l 1

28 XN-NF-86-36 I

Revision 2 The results of the flow-dependent MCPR limit MCPR(f) analysis are given in Table 6.3.

l 6.4 Analyses At Reduced Power Ooeratino Conditions The FWCF and LRNB transients were evaluated at reduced power conditions and E

full recirculation flow to verify the applicability of the power-dependent E MCPR limit curve developed in Reference 9. The results of these confirmatory analyses are given in Table 6.2. For points at or below 40% power, the scram .-

trips associated with turbine control valve position and turbine stop valve position and recirculation pump trip (RPT) were assumed to be overridden.

6.5 Ooeration At Reduced Feedwater Temoerature The FWCF and LRNB transients were reevaluated with COTRANSA with a revised set of initial conditions corresponding to reductions in feedwater temperature.

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29 XN-NF-86-36 Revision 2 -

TABLE 6.1 6

REACTOR AND PLANT CONDITIONS l

TRANSIENT ANALYSES SUPPORTING OPERATION AT RATED CONDITIONS Reactor Thermal Power 3994 MWt (104.2% of rated)

Total Core Flow .

112.5 M1b/hr (100% of rated) i Core Active Flow 100.7 M1b/hr I

Core Inlet Enthalpy 530.2 BTU /lbm Vessel Pressures Steam Dome 1060 psia Upper Pienum 1070 psia Core 1077 psia Lower Plenum 1095 psia Feedwater/ Steam Flow 17.3 M1b/hr Feedwater Enthalpy 403.1 BTU /lbm Recirculation Pump Flow (per pump) 16.0 M1bm/hr G

- . - . . _ , _ . _ _ . _ _ . . _ I

1 i I

l 30 XN-NF-86-36 1 Revision 2 TABLE 6.2 RESULTS OF TRANSIENT ANALYSES AT RATED FLOW INIT INIT PEAK PEAK PEAK DELTA EVENT POWER 1ttd FLOW 3rts POWER 3r.ti HT FLX frtd PRESSURE CPR osia MODEL f

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i 31 XN-NF-86-36 -

Revision 2 TABLE 6.3 RESULTS OF MCPR(f) ANALYSIS ANALYSES FOR 108[ MAXIMUM FLOW FEEDWATER DOME CORE CORE INLET POWER FLOW TEMPERATURE PRESSURE PRESSURE ENTHALPY MCPR A E (deareas F) (nsia) (nsia) (BTU /lb) 106.2 108.0 425 1053 1064 530.6 1.06 g 100.0 100.0 420 1040 1050 527.9 1.10 92.3 90.0 418 1026 1035 525.4 1.14

.! 84.6 80.0 414 1013 1020 522.4 1.18 76.9 70.0 404 1000 1006 518.4 1.24 69.2 60.0 389 988 992 513.6 1.31 l 61.5 50.0 371

' 977 980 507.7 1.39 53.8 40.0 347 968 970 500.0 1.50 l ANALYSES FOR 103.5[ MAXIMUM FLOW FEEDWATER DOME CORE CORE INLET

<t POWER FLOW TEMPERATURE PRESSURE PRESSURE ENTHALPY MCPR g A E (deareas F) (nsia) Insia) (BTU /lb) 102.7 103.5 423 1046 1056 529.2 1.06

, i 100.0 100.0 420 1040 1050 527.9 1.07 l 92.3 90.0 418 1026 1035 525.4 1.12 84.6 80.0 414 1013 1020 522.4 1.16 76.9 70.0 404 1000 1006 518.4 1.22 69.2 60.0 389 988 992 513.6 1.28 61.5 50.0 371 977 980 507.7 1.36

! 53.8 40.0 347 968 970 500.0 1.47 l-

• Conservatively support 107.0% and 102.5% flow settings respectively.

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32 XN-NF-86-36 Revision 2 TABLE 6.4 RESULTS OF TRANSIENT ANALYSES WITH FEEDWATEP. HEATERS OUT OF SERVICE EAPOSURE FEEDWATER MAXIMUM MAXIMUM MAXIMUM DELTA EVENT STATEP0 INT TEMPERATURE POWER HEAT FLUX PRESSURE CPR (MWD /MTU) (DEG F) (PCT) (PCT RTD) (PSIA)

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35 XN-NF-86-36 - '

Revision 2 i

7.0 ANALYSIS OF EXTENDED OPERATING CONDITIONS The NSSS supplier provided analyses which supported operation beyond the normal upper limits on power and flow assumed in the operating power-flow map.

This section provides supporting analyses to validate the continued use of these portions of the operating power-flow map during Cycle 2.

7.1 Desian Bases For Thermal Marains Analyses performed by the NSSS supplier were docketed during Cycle 1 operation

, to allow plant operation outside the normal operating power-flow map. The Extended Load Line (ELL) analysis and the Increased Core Flow (ICF) analysis

! t covered the operating power-flow map depicted in Figure 7.1. In support of

. Cycle 2 operation, Exxon Nuclear has performed confirmatory analyses to justify the continued use of the Figure 7.1 power-flow map and to verify the

! appropriate operating limits.

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7.2 Datemination of Analvtical Statenoints i

Specific power-flow statopoints were identified for analysis. These statepoints were selected at key operating points for validation of operating limits. Table 7.1 shows the thermodynamic state at each of the analytical statopoints in the ELL and ICF transient analyses. The ICF and ELL analytical statopoints are also shown in Figure 7.1.

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I 36 XN-NF-86-36 Revision 2 h) j 1

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7.3 Analysis Of Plant Transients Under ELL Conditions The LRNB and FWCF transients were evaluated at each of the statepoints j identified for further analysis. The LFWH and CRWE generic analyses sumarized in Section 5.0 already cover the extended operating power-flow map.

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37 XN-NF-86-36 -

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1 7.4 Analysis Of Plant Transients Under ICF Conditions Additional power-flow statepoints were identified and analyzed to support operation at greater than rated recirculation flow.

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i 7.5 Validation Of Off-Nominal MCPR Limits

( Figure 7.10 shows the Cycle 1 flow-dependent MCPR limits and the results of the ENC analyses supporting these limits. Results of the flow-dependent .MCPR i analysis reported in Section 6.3 and the plant transient analyses performed at off-nominal flow conditions are plotted on this Figure to demonstrate the

p continued applicability of the limits.

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38 XN-NF-86-36 -s Revision 2 [

Figure 7.11 shows the power-dependent MCPR limit retained from Cycle 1 operation. Plotted on this figure are the indicated MCPR operating limits for the transients evaluated under ELL and ICF. These limits are supported by the g ELL and ICF transient analyses. 5 I

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, 39 XN-NF-86-36 -

l Revision 2 TABLE 7.I ANALYTICAL STATEPOINTS FOR ELL AND ICF EVALUATION j FEEDWATER DOME CORE CORE INLET j POWER FLOW TEMPERATURE PRESSURE PRESSURE ENTHALPY A 1 (deareas F) fosia) (osia) (BTU /lbm)

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40 XN-NF-86-36 Revision 2 ,

TABLE 7.2 RESULTS OF TRANSIENT ANALYSES AT ELL AND ICF STATEPOINTS INIT INIT PEAK PEAK PEAK DELTA EVENT POWER FLOW POWER HT FLX PRESSURE CPR MODEL SttJL M 1rfdL %rtd osia g ELL ANALYSES TO BE SUPPLIED LATER y

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FIG. 7.10 FLOW-DEPENDENT MCPR LIMIT VALIDATION 1.7 -

+ Flow increase analyses, 107% max flow

1. 8 - x Flow increase analyses, 102.5% max flow A Cycle 2 transient analyses 1.5- +

- 1.4 - 107% MAXIMlM FLOW + m b 102.5% MAXIMlM FLOW x

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2- - - Core Flow <50%

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i 52 XN-NF-86-36 -

Revision 2 ,

8.0 ASME OVERPRESSURIZATION ANALYSIS Maximum system pressure has been calculated for compliance with the ASME Boiler and Pressure Vessel Code. This overpressure evaluation postulated the occurrence of a containment isolation event, which includes closure of the

Main Steam Isolation Valves (MSIVs), with concurrent failure of the direct scram trip activated by the MSIV position switches. The analysis demonstrated

.! that the safety valves have sufficient capacity and performance to prevent the L internal pressure from reaching component overpressure limits. The analysis .-

assumed that seven of the twenty safety / relief valves were out of service.

I 8.1 Desian Basis I

The reactor conditions used in the overpressure analysis are sumarized in 3

Table 8.1. Because not all of the phenomena contributing to the overpressure l effects are clearly conservative at the same power-flow statopoint, the overpressurization analysis was per'anned at three different statepoints at I which rated power may be attained. The results of all three analyses are I

contained in this report; the same conclusions are indicated in each instance.

I

The containment isolation event was selected for the overpressure analysis because the single failure assumption (the direct scram on MSIV closure) makes f{ the event more severe than the load rejection transient which was evaluated for thermal margins. If the direct scram is allowed to function, the MSIV closure event does not threaten pressure or critical heat flux margins. The

, overpressure analysis was performed with COTRANSA (Ref. 2).

, 8.2 MSIV Closure Analysis 1 '. The results of the three MSIV closure analyses are given in Table 8.2. The system performance during the MSIV closure event from 73.8% flow conditions is shown in Figure 8.1.

f l

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53 XN-NF-86-36 Revision 2 f

Because the direct scram trip was not allowed to function, reactor power was terminated by the high flux scram trip. The setpoint for the high pressure scram was reached shortly after the high flux trip was reached.

The maximum pressure calculated for the 73.8% flow case, the pressure inside I

the reactor vessel during this postulated event was 1281 psig (1296 psia),

which is less than 110% of the design pressure of the reactor vessel. The maximum pressure in the steamlines was calculated to be 1253 psig (1268 psia). g The maximum pressure calculated to occur in the steam dome was 1265 psig (1280 -

a psia).

8.3 Steam Dome Pressure Safety limit I

Extrapolation of these results indicates that an overpressurization transient which reaches the physical pressure limit inside the vessel pressure boundary g would have a maximum steam dome pressure of 1343 psig (1358 psia). Retention (

of the existing 1325 psig pressure safety limit is conservative relative to this comparison.

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54 XN-NF-86-36 -

Revision 2 TABLE 8.1 l

REACTOR AND PLANT CONDITIONS I I ASME OVERPRESSURIZATION ANALYSIS i I 100% FLOW CASE  !

4 Reactor Themal Power 3994 MWt (104.2% of rated)

Total Core Flow 112.5 M1b/hr (100% of rated) ,.

Core Active Flow 100.7 M1b/hr Core Inlet Enthalpy 530.2 BTU /lbm Vessel Pressures Steam Dome 1060 psia Upper Plenum 1070 psia Core 1077 psik Lower Plenum 1095 psia Turbine Pressure 975 psia Feedwater/ Steam Flow 17.3 M1b/hr Feedwater Enthalpy 403.1 BTU /lbm Recirculation Pump Flow (per pump) 16.0 M1bm/hr e

1 4

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55 XN-NF-86-36 g Revision 2 g TABLE 8.2 RESULTS OF ASME OVERPRESSURIZATION ANALYSIS MAXIMUM PRESSURE MAXIMUM POWER FLOW POWER VESSEL STEAMLINE DOME R A E (%) Iguij t) 1g111), (osia) 3 104.2 108.0 357.6 1288 1260 1249

~

104.2 100.0 336.4 1288 1264 1252 104.2 73.8 384.9 1296 1280 12ES I

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.. .-,m m.m

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ass. aseserc seres /es it.as.ar.

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Revision 2 i

9.0 SPECIFICATION OF OPERATING LIMITS i

i The operating limits for Cycle 2 operation of Grand Gulf Unit 1 are based on the Cycle 1 operating limits. These limits are based on limits provided in References 9 and 10 which have been modified as required.

9.1 MCPR Operatina Limits 1

The power- and flow-dependent effects which were assessed in the transient .-

analyses described in this report have been assembled into discrete limits on

operating MCPR values. At any allowable operating power-flow state, the MCPR operating limit is the greater of the flow-dependent MCPR(f) limit and the

'} power-dependent MCPR(p) limit.

The rated conditions MCPR result is included in the MCPR(f) curve.

t 9.1.1 Flow-Denendent MCPR Limit

,i il The flow-dependent MCPR operating limit is determined from the quasi-static

analysis of the recirculation flow increase transient. Observance of the limit for each fuel bundle in the core during operation at less than full flow  :

( conditions assures that during an uncontrolled flow increase transient which l teminates at the flow limit setpoint of the recirculation system (either  ;

102.5% or 107%), the MCPR Fuel Cladding Integrity Safety Limit will not be violated. The conservatively performed analysis assumed maximum flow runout capabilities of 103.5% and 1085, respectively.

The flow-dependent MCPR limits are given in Figure 9.1 and are applicable to all fuel types. These results are retained from the Cycle 1 flow-dependent

+

MCPR limit and validated for Cycle 2 operation by analyses in this report as described in Section 7.5.

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Revisien 2 -

9.1.2 Power-Decendent MCPR Limit The power-dependent MCPR operating limit is detemined from the generic CRWE analysis and the plant-specific analyses of LRNB and FWCF transients at representative conditions blanketing the operating power-flow map. Observance of this limit for each fuel bundle in the core during operation at less than full power conditions assures that the MCPR Fuel Cladding Integrity Safety

~

Limit will not be violated during anticipated operational occurrences.

The existing MCPR(p) function is supported by the ENC analyses. The power-dependent MCPR limit is given in Figure 9.2 and is applicable to all fuel types.

9.2 MAPLHGR Ooeratina Limits The transient design LHGR limit for the ENC XN-18x8 fuel is established and justified in Reference 11 to protect against fuel damage (1% uniform clad strain and centerline melting). Operating ENC MAPLHGR limits are established from the steady state design LHGR limit as described in References 12 and 13.

l

Power- and flow-dependent factors were assigned to the MAPLHGR limits for

protection against fuel damage mechanisms associated with the initiation of g

transients at off-nominal conditions. 3 The LOCA analyses reported in Reference 14 were performed in a bounding manner such that MAPFAC(f) and MAPFAC(p) adjustment is not required for satisfaction of 10 CFR 50.46. g

[

9.2.1 Flow-Decendent MAPLHGR Limit Multiolier The MAPFAC(f) operating limit factor was evaluated for ENC fuel in Grand Gulf I

through analysis of flow increase transients. The results of the ENC analysis i

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59 XN-NF-86-36 . 1 Revision 2 .

l which are reported in Reference 12 developed the flow-dependent MAPLHGR limit factor given in Figure 9.3.

c 5 9.2.2 Power-Denendent MAPLHGR Limit Multiolier The MAPFAC(p) operating limit factor was validated for Cycle 2 operation of j Grand Gulf Unit I through the analysi: of the FWCF, LRNB, and CRWE transients.

The transient analyses which were used to evaluate power-dependent MCPR limit effects were also used to evaluate power dependent LHGR effects. .

The MAPFAC(p) function was determined by observing the performance of the clad i

surface heat flux during the transient events. The maximum relative heat flux increase was determined from the COTRANSA hot channel model at each of the power-flow operating statepoints evaluated for ELL and ICF operation. The reciprocal of this heat flux increase factor is the fraction of the fuel i damage threshhold LHGR value at which the fuel may be operated prior to the transient without reaching the threshhold during the transient.

1 The ENC fuel damage threshhold LHGR limit is always at least 120% of the design LHGR limit (Reference 11). The appropriate power-dependent factor to be applied to the operating MAPLHGR limit for protection of the fuel damage j threshhold LHGR limit may be determined by increasing the LHGR fraction from the. previous paragraph by a factor of 1.20.

e The FWCF and LRNB results of the MAPFAC(p) determination are shown in Table 9.1 and compared with the Cycle 1 MAPFAC(p) limits in Figure 9.4. The CRWE results of Reference 7 were confirmed to also be bounded Figure 9.4. These analyses and comparisons validate the existing limits for Cycle 2 operation with ENC fuel, i

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60 XN-NF-86-36 Revision 2 TABLE 9.1 MAPFAC(p) ANALYSIS RESULTS gagig ELQM HEI HEAT FLUX FACTOR MAPFACfn)

TO BE SUPPLIED LATER I

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FIG. 9.1 FLOW-DEPENDENT MCPR LIMIT

1. 7 -

2 l 1. 6 -

l

1. 5. -

! 107% MAXIMtM FLOW 102.5% MAXIMtN FLOW

- 1. 4 -

ta_

m W W O_

O

., r 1. 3 -

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1. 2 -

i M >C 1.1 - Of t'i

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2. 4 -

l

2. 2 -

ore Flow >50%

2-Core Flow 550%

- 1. 8 -

(L m

E E n_

o r 1. 6 -

s

1. 4 -

1. 2 -

EE 1 i . . , , . . . . . . , S $,

0 10 20 30 40 50 60 70 80 90 100 110 120 mg CORE POWER, % OF RATED 9

g G M**

i e

FIG. 9.3 FLOW-DEPENDENT MRPLHGR FRCTOR 1.1 -

i 1-1 i

0. 9 -

l 4

, G_ m m

w c3 a- 0. 8 -

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a- ENC 107% MAX FLOW r

GE 107% MAX FLOW

0. 7 - ENC 102.5% MAX FLOW i

GE 102.5% MAX FLOW l

0. 6 -

E?E k

1 0.5 , , , , , , ,

88

, , , , , , 4, 0 10 20 30 40 50 60 70 80 90 100 110 120 m f CORE FLOW, */. OF RflTED f

)

d

i FIG. 9.4 POWER-DEPENDENT MRPLHGR FRCTOR 1.1 -

]

A Transient analyses i 1-i i

0. 9 -

o _ 0. 8 -

@ Core Flow 550% Power >40% g

b_ Power 540% All Core Flows n_

! a- 0. 7 -

E

0. 6 -

Core Flow >50%

Power 540%

0. 5 -

NE 0.4 , , , , , ,

60 70 80 90 100 110 120 Ef Ng 0 10 20 30 10 50 CORE POWER, */. OF RATED W

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Revision 2 i

g

10.0 REFERENCES

1. " Exxon Nuclear Methodology for Boiling Water Reactors: THERMEX Themal t Limits Methodology, Susunary Description," XN-NF-80-19(P), Volume 3,

, Revision 1, Exxon Nuclear Company, Richland, Washington (April 1981).

2. " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors,"

XN-NF-79-71(P), Revision 2, Exxon Nuclear Company, Richland, Washington (November 1981), as supplemented.

. 3. " Exxon Nuclear Critical Power Methodology for Boiling Water Reactors,"

! XN-NF-524(A), Revision 1, Exxon Nuclear Company, Richland, Washington '

(November 1983).

l 4. "The XN-3 Critical Power Correlation," XN-NF-512(A), Revision 1, Exxon

! Nuclear Company, Richland, Washington (March 1981).

5. Paul N. Senerville, " Tables for obtaining Non-Parametric Tolerance Limits," Annals of Nathematical Statistics, Vol. 29, No. 2 (June 1958),

pp. 599-601. .

6. "A Generic Analysis of the Loss of Feedwater Heating Transient for Boiling Water Reactors," XN-NF-900(P), Exxon Nuclear Company, Richland, Washington (February 1986).

7. BWR/6 Generic Rod Withdrawal Error Analysis," XN-NF-825(P), Exxon Nuclear Company, Richland, Washington (April 1985), as supplemented.

8. " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors,"

XN-NF-79-71(A), Supplements 1-3, Exxon Nuclear Company, Richland, Washington (March 1982).

9. "GGNS Maximum Extended Operating Domain Analysis," General Electric l Company, San Jose, California (March 1986).

l

10. "GGNS Single Loop Operation Analysis," General Electric Company, San t Jose, California (February 1986).

11. " Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel,"

XN-NF-85-67(P), Revision 1, Exxon Nuclear Company, Richland, Washington (April 1986).

12. " Grand Gulf Unit 1 Cycle 2 Reload Analysis," XN-NF-86-35, Revision 1, I

• Exxon Nuclear Company, Richland, Washington (June 1986).

13. " Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," XN-NF-80-19(A), Volume 4, Revision 1, Exxon Nuclear Company, Richland, Washington (May 1986).

f l

P 66 XN-NF-86-36 Revision 2 I

14. " Grand Gulf Unit 1 LOCA Analysis," XN-NF-86-38, Exxon Nuclear Company, Richland, Washington (May 1986).

15. "XCOBRA-T : Computer Code for BWR Transients Thermal-Hydraulic Core Analysis," XN-NF-84-105(P), Volume 1 and Revision 1 of Supplements 1 and 2, Exxon Nuclear Company, Richland, Washington (May 1985 and March 1986).

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XN-NF-86-36 e f Revision 2 -

Issue Date: 6/30/86 I

I GRAND GULF UNIT 1 CYCLE 2 PLANT TRANSIENT ANALYSIS I

DISTRIBUTION WC Arcieri (ENSA)

DJ Braun JC Chandler TP Currie RE Collingham 4

• RA Decker SF Gaines BJ Gitnick (ENSA)

JG Hwang (Nutech) .

JG Ingham SE Jensen TH Keheley JE Krajicek TL Krysinski JN Morgan LA Nielsen TW Patten DA Prelewicz (ENSA)

DR Swope HE Williamson MP&L/JD Floyd (40)

Document Control (5) l<

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