ML20212M735

Rev 3 to Grand Gulf Unit 1 Cycle 2 Plant Transient Analysis

ML20212M735
Person / Time
Site:	Grand Gulf
Issue date:	08/09/1986
From:	Chandler J, Collingham R, Ward G SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML20212M724	List: AECM-86-0255, Addendum to 860714 Application for Amend to License NPF-29, Forwarding Rev 3 to XN-NF-86-35, Cycle 2 Reload Analysis & Rev 3 to XN-NF-86-36, Cycle 2 Plant Transient Analysis. Proposed Tech Specs Re Detect & Suppress Region Also Encl ML20212M728 ML20212M732 ML20212M735
References
TAC-61930, XN-NF-86-36, XN-NF-86-36-R03, XN-NF-86-36-R3, NUDOCS 8608270105
Download: ML20212M735 (90)

Text

I l X \-N =-86-36 g REVISIO\ 3 I

I GRA\ J GU_: U\

1 CYC_E 2  :

l 3

_A\~~ - RA\SIE\~~ A\ A_YSIS I

E l AUGUS-~ 1986 I

I I 9 C _A\ J, WA 99352 EXXOh NUCLEAR COM PANY, lh C.

,g I

I !8PS8!n 8889!a'

J u

XN-NF-86-36 H Revision 3 L Issue Date: 8/9/86 7 GRAND GULF UNIT 1 CYCLE 2 PLANT TRANSIENT ANALYSIS Prepare: \ - M d dd J. C. Chandler, Senior Engineer ~

BWR Safety Analysis 3

Concur: $$4ha R. E. Colling)fam, Manager ssm hlb BWR Safety AWalysis F

Concur: / )/ J/7/PC G. N. Ward, Nanaber

[ Reload Licensing Concur: A #1/J4L !4 //, %f fi J. N. Morga}(, Manager / D/ ~ /r Customer S(rvices Enginetiring Approve: g1 d Mrd H. 7 Williamsnn, Manager

{ Lic6nsing and Safety Engineering I

j Approve:

G. L. Ritter, Mahager

[ 7

'/

Fuel Engineering and Technical Services I

tmrc ERON NUCLEA 9 COM 3A\ Y, i\ C.

I I

I NUCLEAR REGULATORY CO51511SSION REPORT DISCLAIS!ER I'

151PORTANT NOTICE REG ARDING CONTENTS AND USE Or TitIS DOCU5 TENT PLEASE READ CAREFULLY This technical report was derived through research and development programs sponsored by Exxon Nuclear Company,Inc. It is being submitted by Exxon Nuclear to the U1 Nuclear Regulatory Commission as part of a technical contribution to facilitate safety analyses by licensees of the US Nuclear Regulatory Commission which utilize Exxon Nuclear-fabricated reload fuel or other technical services provided by Exxon Nuclear for light \

water power reactors and it is true and correct to the best of Exxon Nuclear's knowledge,information, and belief. Ti.e information contained herein may be used by the U1 Nuclear Regulatory Commission in its review of this report, and under the terms of the respective agreements, by licensees or applicants before the U1 Nuclear Regulatory Commission which are customers of Exxon Nuclear in their demonstration of compliance with the U1 Nucicar Regulatory Commission's regulations.

Exxon Nuclear's warranties and representations concerning the subject natter of this document are those set forth in the agreement tetween Exxon Nuc! car and the customer to which this document is issued. Accordingly, except as otherwise expressly provided in such agreement, neither Exxon 3 Nuclear nor any person acting on its behalf:

g A. Makes any warranty, or representation express or implir.d. with respect to the accuracy, completeness, or usefviness of the information contained in this document, or that the use of any information, apptratus, method, or process disclosed in this document will not infringe privately owned rights, or B. Ar,sumes any liabilities with respect to the use of,or for damages sesulting from the use of, any information, apparatus, method, or process disclosed in this document.

I 1

1 I'

1 XN-NF-86-36

% Revision 3 l-

SUMMARY

OF REVISIONS Revision 3 to XN-NF-86-36 is issued to incorporate the results of transient analyses with COTRANSA and XCOBRA-T. Major report changes are as follows:

( '

COTRANSA transient results are added to the body of the report.

These results are reinstated from the Revision 1 report. Revision 2 did not report these results because of a potential nonconservatism in the COTRANSA hot channel model.

k XCOBRA-T confirmation of the COTRANSA hot channel model results is added to the report as Appendix 8.

This document supersedes all previous revisions of XN-NF-86-36.

[

{

i XN-NF-86-36 Revision 3 TABLE OF CONTENTS SECTION A E.ASE

( 1.0 INTR 000CTION.......................................... 1 2.0

SUMMARY

............................................... 3 3.0 ANALYTICAL METH00S AND COMPUTER M00ELS................. 12 3.1 THERMEX Thermal Limits Methodology..................... 12 3.2 COTRANSA System Model.................................. 12 3.2.1 Plant-Specific Modifications to C0TRANSA............... 12 3.2.2 Treatment of Uncertainties in Data..................... 13 3.3 Cri ti cal Power Me thodol ogy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 3.4 Validation of COTRANSA Hot Channel Result .. . . . . . . . . . . . . 14

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 5.0 GENERIC TRANSIENT ANALYSES APPLICABLE TO GRAND GULF.... 20

{

5.1 Loss Of Feedwater Heating.............................. 20

( 5.2 Control Rod Withdrawal Error........................... 20 6.0 MARGIN..................

( TRANSIENT ANALYSES FOR THERMAL 23 6.1 Design Bases For Thermal Margins....................... 23

( 6.2 6.2.1 Analysis Of Plant Transients At Rated Conditions.......

Load Rejection Without Bypass..........................

25 26 6.2.2 Feedwater Controller Failure........................... 27 6.3 Analysis At Reduced Flow Operation Conditions.......... 27 6.4 Analn ns At Reduced Power Operating Conditions......... 28

)

{

ii XN-NF-86-36 Revision 3 TABLE OF CONTENTS (Continued) 6.5 Operation At Reduced Feedwater Temperature............. 28 7.0 ANALYSIS OF EXTENDED OPERATING CONDITIONS.............. 36 7.1 Design Bases For Thermal Margins....................... 36 7.2 Determination Of Analytical Statepoints. . . . . . . . . . . . . . . . 36 7.3 Analysis Of Plant Transients Under ELL Conditions...... 37 7.4 Analysis Of Plant Transients Under ICF Conditions... ... 38 h

7.5 Validation Of Off-Nominal MCPR Limits. . . . . . . . . . . . . . . . . . 38 8.0 ASME OVERPRESSURIZATION ANALYSIS....................... 53 I

8.1 Design Basis........................................... 53 8.2 MSIV Closure Analysis.................................. 53 8.3 S team Dome Pres sure Sa fety L imi t. . . . . . . . . . . . . . . . . . . . . . . 54 ,

9.0 SPECIFICATION OF OPERATING LIMITS...................... 58 9.1 MCPR Operating Limits.................................. 58 ..

9.1.1 Flow-Dependent MCPR Limit.............................. 58 9.1.2 Power-Dependent MCPR Limit............................. 59 9.2 MAPLHGR Operating Limits............................... 59 9.2.1 Flow-Dependent MAPLHGR Limit Multiplier................ 59 9.2.2 Power-Dependent MAPLHGR Limit Multiplier............... 60

10.0 REFERENCES

............................................. 66 APPENDICES A. COTRANSA HOT CHANNEL M0 DEL............................. A-1 -

B. Validation of COTRANSA Hot Channel Delta-CPR Results with the XCOBRA-T Transient Thermal-Hydraulic Model.... B-1 I

i

I jjj XN-NF-86-36 Revision 3 LIST OF TABLES TABLE TITLE PM 2.1 Summary of Transient Analysi s Resul ts. . . . . . . . . . . . . . . . . . 5 2.2 Overpressurization Analysis Results.................... 6 6.1 Reactor and Plant Conditions, Transient Analyses 6 6.2 Supporting Ope-ation at Rated Conditions...............

Results of Transient Analyses at Rated Flow. . . . . . . . . . . .

30 31 6.3 Resul ts of MCPR(f) Analysi s. . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 I-6.4 Results of Transient Analyses with Feedwater Heaters Out of Service......................................... 33

$3 7.1 7.2 Analytical Statepoints for ELL and ICF Evaluation......

Results of Transient Analyses at 40 ELL and ICF Statepoints................................ 41 8.1 Reactor and Plant Conditions, ASME Overpressurization Analysis, 100% Flow Case............ 55 8.2 Resul ts of ASME Overpressurization Analysi s. . . . . . . . . . . . 56 9.1 MAPFAC(p) Analysi s Resul ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 61 I

I  :

I I

I I

I I

I

I I iv XN-NF-86-36 Revision 3 LIST OF FIGURES FIGURE TITLE PE 2.1 Operati ng Power-Fl ow Map . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 2.2 Flow-Dependent MCPR Limit............................. 8

,. 2.3 Power-Dependent MCPR Limit............................ 9 I 2.4 2.5 Flow-Dependent MAPLHGR Factor.........................

Power-Dependent MAPLHGR Factor........................

10 11 4.1 Design Basis Radial Power Distribution. . . . . . . . . . . . . . . . 17 I 4.2 4.3 Design Basis Local Power Distribution, ENC XN-1 8x8 Fue1.....................................

Design Basis Local Power Distribution, GE 8x8 Fuel....

18 19 5.1 Powar-Dependent MCPR Limit from Generic CRWE Analysis................................. 22 l

6.1 System Traces, LRNB 104.2/100......................... 34 6.2 System Traces, FWCF 104.2/100......................... 35 I~

7.1 7.2 7.3 Operati ng Powe r- Fl ow Map . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

System Traces, LRNB System Traces, FWCF 104.2/73.8........................

70/40.............................

42 43 44 7.4 System Traces, LRNB 25/73.8........................... 45 7.5 System Traces, LRNB 25/40............................. 46 7.6 System Traces, LRNB 104.2/108......................... 47 7.7 System Traces, FWCF 104.2/108......................... 48 I 7.8 7.9 7.10 System Traces, LRNB System Traces, FWCF Flow-Dependent MCPR Limit Validation..................

40/108............................

40/108............................

49 50 51 7.11 Power-Dependent MCPR Limit Validation................. 52 8.1 System Traces, ASME Overpressurization Analysis....... 57 I 9.1 9.2 9.3 Flow-Dependent MCPR Limit.............................

Power-Dependent MCPR Limit............................

Flow-Dependent MAPLHGR Factor.........................

62 63 64 9.4 Power-Dependent MAPLHGR Factor........................ 65 I

I I

I ,

y XN-NF-86-36 Revision 3 ACKNOWLEDGEMENT

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

S. E. Jensen

( T. H. Keheley T. R. Lindquist L. A. Nielsen

[

i J. A. White B. J. Gitnick (ENSA)

J. G. Hwang (ENSA)

(

(

l

(

{

1 XN-NF-86-36 Revision 3

(

1.0 INTRODUCTION

This report desc.-ibes the plant system transient analyses performed by Exxon

{ Nuclear Co., Inc., in support of the Cycle 2 (XN-1) reload for Grand Gulf Unit

1. This cycle is scheduled to commence operatic.n in Fall 1986.

Cycle 2 is the first cycle during which ENC 8x8 fuel will be irradiated in

( Grand Gulf Unit 1. In addition to the ENC 8x8 fuel, the Cycle 2 core will contain a significant number of 8x8 assemblies fabricated by General Electric.

Operating limits for these fuel types during Cycle 2 operation are established

{ in this report.

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 during transient events.

(

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.

t In addition to the analysis of system transients, this document also contains the results of analyses for the MCPR Fuel Cladding Integrity Safety Limit and the Maximum Overpressurization Accident for qualification in compliance with the ASME code.

1 The combination of these analytical results into a framework of operating limits is discussed in Section 9.0. In addition to the thermal margin limits determined from the plant transient analyses, this section also addresses power- and flow-dependent factors for modification of the MAPLHGR limits when operating at other than rated power and flow conditions. In general, f - -

I 2 XN-NF-86-36 Revision 3 off-nominal operating limits have been retained from the Cycle 1 analy es I

! performed by the NSSS supplier (Ref. 9) except in instances where the ENC analyses indicated a need for more restrictive limits.

I I l I

I I

i l

I

' I I

I I

I l

I I

I 3 XN-NF-86-36 Revision 3 2.0

SUMMARY

I This report provides the results of the simulations of the limiting plant transients for Cycle 2 operation of Grand Gulf Unit 1. These transients are I the generator load rejection with concurrent failure of the condenser bypass system (LRNB) and the failure of the feedwater flow control system to maximum 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 plant-specific transient analyses reported in this document require a rated MCPR operating limit of at least 1.10. 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 I Fuel Loading Error (FLE) event, which_ is reported in the reload analysis (Ref.

12), requires a nominal MCPR limit of at least 1.17.

The flow-dependent MCPR limits for Cycle 2 operation of Grand Gulf Unit 1 are shown in Figure 2.2. These limits were retained from Cycle 1 and validated l for all fuel types by the conservative analysis of a hypothetical recirculation flow increase transient. The 1.18 Rated Conditions MCPR limit, I 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 l continued uso during Cycle 2 for all fuel types by the analyses reported in this document.

I I

E

4 XN-NF-86-36 I

Revision 3 '

The flow-dependent MAPLHGR limit factors for Cycle 2 operation of Grand Gulf Unit 1 are shown in Figure 2.4. These limit factors t:ere retained from Cycle 1 for GE fuel and calculated separately for ENC fuel.

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

The results of the maximum overpressurization accident analysis performed for I

compliance with the ASME code are shown in Table 2.2. These results demonstrate that design pressure margins are protected during the postulated worst overpressurization event in the plant design basis with seven safety / relief valves out of service.

I I

I I

I I

I I

I I

. . . . . . . . . . . . _J

I 5 XN-NF-86-36 Revision 3 TABLE 2.1 I

SUMMARY

OF TRANSIENT ANALYSIS RESULTS I

INIT INIT PEAK PEAK PEAK DELTA EVENT POWER FLOW POWER HT FLX PRESSURE CPR MODEL

_7s_td itM _5fM %rtd osia I ANALYSES FOR RATED MCPR LIMIT

'I LRNB FWCF 104.2 104.2 100 100 131.4 109.9 100.5 107.3 1229 1194 0.02 0.04 COTRANSA COTRANSA LFWH See generic analysis, Ref. 6) 0.08 XTGBWR CRWE See generic analysis, Ref. 7) 0.10 XTGBWR ANALYSES FOR EXTENDED LOAD LINE OPERATION LRNB 104.2 73.8 191.8 101.5 1231 0.05 COTRANSA LRNB 70 100 103.8 69.1 1192 0.04 COTRANSA I LRNB LRNB LRNB 70 40 25 40 100 73.8 70.0 142.4 60.9 61.3 48.7 44.4 1185 1143 1180 0.16 0.27 1.05 COTRANSA COTRANSA COTRANSA LRNB 25 40 86.6 38.3 0.80 I

1174 COTRANSA FWCF 104.2 73.8 109.5 106.5 1189 0.06 COTRANSA FWCF 70 100 73.3 70.9 1071 0.03 COTRANSA I FWCF FWCF FWCF 70 40 25 40 100 73.8 74.9 69.6 32.5 71.7 42.8 26.9 1064 1009 974 0.15 0.07 0.08 COTRANSA COTRANSA COTRANSA FWCF 25 40 27.8 25.5 966 0.07 COTRANSA ANALYSES FOR INCREASED CORE FLOW OPERATION LRNB 104.2 108 144.6 101.5 1232 0.02 COTRANSA LRNB 40 108 153.3 48.7 1137 0.28 COTRANSA FWCF 104.2 108 110.1 107.4 1198 0.04 I

COTRANSA FWCF 40 108 76.5 43.6 1012 0.09 COTRANSA Peak heat flux value near time of minimum delta CPR.

I I

I 6 XN-NF-86-36 I

Revision 3 TABLE 2.2 OVERPRESSURIZATION ANALYSIS RESULTS I

I 73.8% FLOW CASE 100% FLOW CASE 108% FLOW CASE Maximum Flux, % 384.9 336.4 357.6 Maximum Pressure E 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 g

l I

I I

I I

I I

I I

I

M M M M M M M M m M m M M M M M M M M l FIG. 2.1 OPERRTING POWER-FLOW MRP 120 -

i l

i (73.8,100) (105,100)

I I

ELL Region r-ICF O 80 - / Region 4

)

a: 100% Rod Line

! $ {

i a m-

\ ~

8 o.

w #- . /

105,40) 5 U

i (40,25) (75,25) 20 -

2%

o , , , , , , , , , , , ga i 0 10 20 30 40 50 60 70 80 90 100 110 I ww I CORE FLOW, % OF RATED i

l 1

I i

l

i l

1 FIG. 2.2 FLOW-DEPENDENT MCPR LIMIT 1.7 -

1. 6 -

l

1. 5 -

l 1

i

- 1. 4 - 107% MAXIMUM FLOW m i e l g 102.5% MAXIMtN FLOW i, o l r 1.3 -

i

1. 2 -

1.1 - a's.

24 1 . . . . . . . . , . .

N

-=

i ,

I O 10 20 30 40 50 60 70 00 90 100 110 120 i CORE FLOW, */. OF RATED t

i t

I i

l M M M M M M M M M M M M M M M M M M M

M M M M M M M M M M M M M M M M M M i

f FIG. 2.3 POWER-DEPENDENT MCPR LIMIT

2. 4 -

2 2 Core Flow >50%

' 2-Core Flow 550%

i

! - 1. 8 - *

Q-e i 0-

! E$ 1.s-1 I

l 1. 4 -

1. 2 -

l; 75 tm i

i II 1 ' ' ' ' "O*

b b b '

i i 1

O 10 4O 60 70 80 90 100 110 120 CORE POWER,-% OF RATED

FIG. 2.4 FLOW-DEPENDENT MRPLHGR FACTOR 1.1 -

1-GE FUEL MA:( FLOW = 102.5%

MA:( FLOW = 107%

O.9 -

l E o

0. 8 -

{

0-ENC FUEL MAX FLOW = 102.5%

E MAX FLOW = 107%

0. 7 -

0. 6 -

"i E En 0.5 , , , , , , , , , , , i =T 0 10 20 30 40 50 60 70 80 90 100 110 120 "g CORE FLOW, % OF RATED em aus use sua nun use sus una mas em uns aus num uma uma amm ama aus ses

W M M M M M M M M M M M M M M m m m M FIG. 2.5 POWER-DEPENDENT MRPLHGR FRCTOR 1.1 -

1- ~

a

0. 9 -

{ 0. 8 - Power >40%

c3 All Core Flows b

S E

0.7 - Core Flow 550% -

Power 540%

Core Flow >50%

0. 6 - Power 540%

0. 5 -

%'s 04 ' ' ' '

0 l0 b b 4O b b 7O b 9O 150 110 lb kh CORE POWER, 7. OF RflTED wk

r.

~

> l

~

12 XN-NF-86-36 L Revision 3 3.0 ANALYTICAL METHODS AND COMPUTER MODELS E This section describes the analytical methods used in the analysis of anticipated transient conditions for operation of the Grand Gulf Unit 1 7 reactor during Cycle 2.

L 3.1 THERMEX Thermal _Ljmits Methodoloov b

Exxon Nuclear's THERMEX thermal limits methodology is described in Reference

{ 1. This topical report summarizes the analytical methods used for thermal hydraulic design analysis t.nd 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, and the COTRAN core kinetics model, which includes a one-dimensional neutronic representation of the core. Calculational uncertainties are bounded in the analysis using the methods described in Reference 8.

3.2.1 Plant-Soecific Modifications to COTRANSA The COTRANSA model used in Grand Gulf analyses used the hot channel model originally applied to the Susquehanna Unit I plant. The COTRANSA modifications associated with the revised hot channel are described in Appendix A.

Several minor model changes were necessitated by plant configuration features unique to Grand Gulf Unit 1, such as the performance characteristics of the turbine control valve. Other changes allowed explicit modeling of the BWR/6

( safety grade high water level scram.

[

I 13 XN-NF-86-36 i

Revision 3 The adequacy of the Grand Gulf COTRANSA model was verified through prediction of feedwater and pressure contrciler tests and plant startup tests for recirculation pump coastdown and load rejection with bypass. The recirculation pump coastdown input bounded by test data were used for analysis of the recirculation pump flow coastdown following trip. COTRANSA conservatively predicted the reacter power history in the load rejection startup test, confirming the adequacy of the Grand Gulf COTRANSA model for license basis analyses.

3.2.2 Treatment of Uncertainties in the Data I

For the evaluation of the system transients covered in this analysis, plant variables were considered to be at conservative values. For variables covered i 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 vere bounded in the I

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.

l 3.3 Critical Power Methodolooy l 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 l

combined with power distribution measurement uncertainties and the l 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 l

I I

I 14 XN-NF-86-36 Revision 3 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.

3.4 Validation of COTRANSA Hot Channel Results The delta-CPR values calculated by the COTRANSA hot channel model were validated through confirmatory analyses with XCOBRA-T (Ref. 15). These analyses are described in Appendix B.

I I

I I

I il 5

m I

I I

,I lI

l I

i I 15 XN-NF-86-36 Revision 3 4.0 MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT I 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 E 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 L would be expected to avoid boiling transition. This section describes the calculation and presents the analytical results.

E 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%.

L 4.1 oesion eas4s Power oistribution

{ Predicted power distributions were extracted from the fuel management analysis for Grand Gulf Unit 1 Cycle 2. These radial power distributions were evaluated for performance as the design basis radial power map, and the distribution at 6000 MWD /MT cycle exposure was conservatively selected as the most severe expected distribution for the cycle. The distribution was skewed toward higher power factors by the addition of bundles with a radial peaking factor approximating an operating MCPR level of 1.17 at full power.

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 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 8000 MWD /MT was selected as the design basis local power distribution. This

( distribution is shown in Figure 4.2.

C

16 XN-NF-86-36 I

Revision 3 I-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 g analysis. 5 4.2 Calculation Of The Number Of Rods In Boilina 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 g using the uncertainties identified in Reference 4. Consistent with Reference 5 3, non-parametric tolerance limits (Ref. 5) were used in lieu of Pearson curve fitting.

i A total of 1000 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.

I I.

1 I

I I

I I

1

M M e ?n.ha i

? 0" m 8 u '1 e

6

'1 m N O

I T

U .

m B I -

.' l 4

R _

T .

S _

m I D

R O

R .

E T W - ,

2.

1 C m O P

R F

L .

G A

I -

N m D A

R

- ,1 I

K R

S - E P

m I _

S -

A _ L B

N .

,0 8.AI D

m G I

S _

A R

E -

D .

m 1

,0 6

E 4

W G I _

F _

4

- ,0 M -

2 M 0 0 s 0 5 0 0

4 5

3 0

3 a 2 1 1 M

mJdgm t u.o "umz=

t M L '

n u

S M

.l  : 1 ' .l ll1l u

l 18 XN-NF-86-36 Revision 3 I

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

0.92 : 0.96 : . . .

0.96 : 0.97  : 0.91  : 1.07  : 1.06 : 0.97  : 0.94 : 0.95 :

1.02 : 0.97

1.02  :

! : 1.02 : 0.91 1.04 :

1.01  : 0.99 :

1.07  : 1.01  : 0.00 0.87  : 0.99 : 1.06 : 1.06  :

1.06 : .

l 1

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

1.06 : . . . . .

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

1.02  : 0.97 : . . . .

1.05 : 0.95

\

1.07  : 0.87 :  :

0.95  : 0.94  : 0.97  : 1.06 : .

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

1.04  : 0.95 : .

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

- E

• Fuel rod adjacent to control blade position I

I I

I I 19 XN-NF-86-36 Revision 3 I

I I _.... . _..............____..................._.......___........__... _.

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

I 1.00 :

I : 1.00  : 0.97

0.99 :

1.02  :

1.03  :

1.03 : 0.99 :

I

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

1.03 : 0.99 :

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

0.99 : . . . . .

I  : 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  : 0.58 : 0.99 : 0.99 :

I 1.03  : 1.02  : 0.99 : 0.99 : 0.97 : 1.00 :

1.00 -

0.99 :

I ...................._.....___.... ___....__................._______......

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

1.03 :

I  :

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

• Fuel rod adjacent to control blade location j

i 1

1 I

L

[

r 20 XN-NF-86-36 L Revision 3

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

L 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.

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 Load Line (ELL) and Increased Core flow (ICF) regions.

7 The generic conclusions support a MCPR operating limit of at least 1.14 for 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.

5.2 Control Rod Withdrawal Error The Control Rod Withdrawal Error (CRWE) transient has been analyzed by ENC on a generic basis for BWR/6 plants. This generic analysis is documented in

( Reference 7.

The Reference 7 analysis provides a statistical evaluation of the consequences i of the CRWE transient for BWR/6 plant configurations under conditions which

{ cover the normal operating power flow map and the ELL and ICF regions.

I I

21 XN-NF-86-36 I

Revision 3 The generic conclusions support a power-dependent MCPR limit function as shown l in Figure 5.1. This limit was considered in determining the power-dependent i MCPR limits documented in Section 9.0 of this report.

I I

I I

I I

I I

3 l

l t

I I

I I

I

'; ' . i g l I l I I I I l I I l I I 3E I F - - - - -

i g

u r

e MEr z

5 ,- i Ill WE 5d0 'E O.o2 i 1

P .

o i w ,- i e

r D

p e

e . i n ,- i .

d .

e .

n t ( .

M C

d,

)

2 2

P B ,- i i

R d

L 3 i

m H t

i a i

,- i f

r o

m G . .

. e n ,- i . i e .

r i .

c .

C 1

. R .

W ,- i i

E .

A !IlijIiIIIlgIIIliIii11,III1 I 3Il RX n eN a . v-

- iN l

y a - - - - - -

- - - ~ - sF s ,- i-o8 i

s , s a R S 8 a R 3 S ~8 .

S e n6 C 33 u o: k w .a g $ g ^ 6

I 23 XN-NF-86-36 Revision 3 I 6.0 TRANSIENT ANALYSES FOR THERMAL MARGIN I This section covers the determination of the rated MCPR operating limit and the off-nominal MCPR limits associated with the normal power-flow map.

Analyses are reported to support a single value limit comparable to the 1.18 nominal limit used during Cycle 1. Limits for off-nominal conditions in the extended regions of the power-flow map are confirmed by analyses in separate report sections.

6.1 Desian Bases For Thermal Marains 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' limiting events, are covered in the reload analysis report (Ref.12).

I Transient analyses are performed for the limiting pressurization, coolant 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 1 core during Cycle 2 operation. Thermal margin requirements are specified as limits on the Minimum Critical Power Ratio (MCPR) as calculated using the XN-3 Critical Power Correlation (Ref. 4).

l Analyses in this section address rapid pressurization events and vessel inventory increase events; generic analyses for cold water transients and one l of the reactivity anomaly events are referenced in the determination of operating limits.

Flow dependent MCPR operating limits are established by evaluating a flow I increase transient with a resulting power increase which continues until the l

I

I 24 XN-NF-86-36 I

Revision 3 physical maximum performance of the system is reached. These limits assure I

that the MCPR safety limit is protected following such an event.

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 g feedwater controller failure and generator load rejection transients at full 5 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 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 longer control movement is required before the blade tips reach the axial level at which a significant reactivity effect is 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.

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

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 i feedwater temperature values up to 100 degrees F lower than the nominal value.

l I I

I

I I 25 XN-NF-86-36 l

I Revision 3 The MCPR operating limit is determined by comparing the thermal margin requirements of all the anticipated events considered in the design basis.

The operating thermal margin requirements for each anticipated event are determined from existing generic analyses and from plant specific analyses. I For rapid transients, such as those analyzed in this report, thermal margin  ;

requirements are dictated by the change in thermal margin (delta-MCPR) I observed during the transient. In other events, the thermal margin 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.

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

6.2 Analysis Of Plant Transients At Rated Conditions Analyses were performed at analytical statepoints selected to protect allowed I operation of the plant at up to rated power and flow conditions. Initial conditions for these analyses assumed a power level corresponding to 104.2% of rated thermal power and 100% of rated recirculation flow. For analyses at extended power-flow operating points (ELL and ICF), refer to Sections 7.0 of this report.

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.

I .

I I

I 26 XN-NF-86-36 Revision 3 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.

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 puitp trip, and a pressure wave in the steamlines.

The controller also generates a signal to open the steam bypass valves, but I

there is a common 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.

When the pressura wave reaches the core, the core void level is reduced by the 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 g increased heat flux also have a negative reactivity effect but are not 5 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 in Figure 6.1, which depicts the time variance of important reactor and plant parameters during the LRNB transient. B 5

I

I

L .-

I L

XN-NF-86-36

{ I 27 Revision 3 The delta-CPR calculated for the LRNB transient was 0.02, including the calculational uncertainty factors identified in Reference 8. This result indicates a MCPR operating limit requirement of 1.08 for this event.

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.

Failure of tne 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 th'e 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 sone extent.

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

Important core and system parameters are plotted agairst event time in Figure 6.2. The delta-CPR for this event was calculated to be 0.04, indicating a

{ MCPR operating limit requirement of 1.10 for the event.

6.3 Analyses At Reduced Flow Ooeratina Conditions l 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

{

I

l I

28 XN-NF-86-36 I1 Revision 3 increase was conservatively taken as the bounding ascension from a series of I

XTGBWR analyses of flow excursion events.

The thermal hydraulic conditions of the core were calculated by heat balance g at several points along the assumed power ascension line. The change in 5 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.

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

6.4 Analyses At Reduced Power Ooeratina Conditions I

The FWCF and LRNB transients were evaluated at reduced power conditions and full recirculation flow to verify the applicability of the power-dependent MCPR limit curve developed in Reference 9. The results of these confirmatory 5

5 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 50 and 100 degree F reductions in feedwater temperature. Both end of cycle and 2000 MWD /MTU before end of cycle exposures were evaluated. Results of these analyses show an insignificant degradation in overall event consequences with decreasing feedwater temperature. As expected, end of cycle conditions were more limiting yielding slightly higher delta-CPRs. Main analytical results are included in Table 6.4.

I I

I 29 XN-NF-86-36 Revision 3 I The analyses indicate that the thermal margin requirements for operation with feedwater temperatures as low as 320 degrees F are conservatively protected by the existing operating limits, which require a MCPR penalty of 0.01 when I operating with a 50-degree or greater feedwater temperature reduction. Margin requirements remain well below those defined by CRWE considerations.

I I

I I

I I

I I

I I

I I

I

I

I 30 XN-NF-86-36 Revision 3 TABLE 6.1 I

REACTOR AND PLANT CONDITIONS TRANSIENT ANALYSES SUPPORTING OPERATION AT RATED CONDITIONS Reactor Thermal Power 3994 MWt 3 (104.2% of rated) g b*. f rated)

Core Active Flow 100.7 Mlb/hr Core Inlet Enthalpy 530.2 BTU /lbm Vessel Pressures Steam Dome 1060 psia Upper Plenum 1070 psia Core 1077 psia Lower Plenum 1095 psia Turbine Pressure 975 psia Feedwater/ Steam Flow 17.3 Mlb/hr l

Feedwater Enthalpy 403.1 BTU /lbm Recirculation Pump Flow (per pump) 16.0 Mlbm/hr I

I I

I-I I

i I

I I 31 XN-NF-86-36 Revision 3 TABLE 6.2 I RESULTS OF TRANSIENT ANALYSES AT RATED FLOW INIT INIT PEAK PEAK PEAK DELTA I EVENT POWER

%rtd FLOW gr_td POWER

%rtd HT FLX

%rtd PRESSURE CPR osia MODEL I LRNB FWCF 104.2 104.2 100 100 131.4 109.9 100.5 10'.3 1229 1194 0.02 0.04 COTRANSA COTRANSA I LRNB FWCF 70 70 100 100 103.8 73.3 69.1 70.9 1192 1071 0.04 0.03 COTILNSA COTRANSA LRNB 40 100 142.4 48.7 1143 0.27 COTRANSA FWCF 40 100 69.6 42.8 1009 0.07 COTRANSA I

I I

I

I I

I Peak heat flux value near t.ime of minimum delta CPR.

I I

I l

I 32 XN-NF-86-36 I

Revision 3 l

TABLE 6.3 RESULTS OF MCPR(f) ANALYSIS ANALYSES FOR 108% MAXIMUM FLOW l FEEDWATER DOME CORE CORE INLET l

POWER FLOW TEMPERATURE PRESSURE PRESSURE ENTHALPY MCPR

( (%) .{3)_. (dearees F) (osia) (osia) (BTU /lb) 106.2 108.0 425 1053 1064 530.6 1.06 100.0 100.0 420 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 61.5 50.0 371 977 980 507.7 1.39 53.8 40.0 347 968 970 500.0 1.50 ANALYSES FOR 103.5% MAXIMUM FLOW FEEDWATER DOME CORE CORE INLET POWER

(%)

FLOW 11)_

TEMPERATURE (dearees F)

PRESSURE (osia)

PRESSURE (osia)

ENTHALPY (BTV/lb)

MCPR 3 3

l 102.7 103.5 423 1046 1056 529.2 1.06 100.0 100.0 420 1040 1050 527.9 1.07 92.3 90.0 418 1026 1035 525.4 1.12 84.6 80.0 414 1013 1020 522.4 1.16 g 76.9 70.0 404 1000 1006 518.4 1.22 3 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 I

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

I I

t

[

[ 33 XN-NF-86-36 Revision 3 TABLE 6.4 RESULTS OF TRANSIENT ANALYSES WITH FEEDWATER HEATERS OUT OF SERVICE L

EXPOSURE FEEDWATER MAXIMUM MAXIMUM MAXIMUM DELTA r EVENT STATEPOINT TEMPERATURE POWER HEAT FLUX PRESSURE CPR L (MWD /MTU) (DEG F) (PCT) (PCT RTD) (PSIA) 1

[ LRNB LRNB EOC2 E0C2-2000 320 320 142.8 104.2 102.3 100.4 1226 1221

-0.02 0.01 LRNB [0C2 370 140.7 101.7 1224 0.02 LRNB E0C2-2000 370 104.2 99.9 1224 0.01 FWCF E0C2 320 114.3 106.9 1171 0.04 FWCF E0C2-2000 320 115.3 107.3 1162 0.04 FWCF E0C2 370 113.6 107.5 1182 0.04 FWCF EOC2-2000 370 114.0 107.8 1178 0.04

[

(

(

l .

34 XN-NF-86-36 Revision 3 l

I5C .. =6 ==r6u 66,u 2 HE FLUX

3. REC RCULAf!DS: FL3W m 4. VES EL stent FLOW 12,w a. rtt.ivAlta FLWu

,2 jh3 10-5 15 i

k ~

N A  %

e .

g i

I I

25 J

[ Y , , , , , ,

~v

-26.0 0.5 10 1.5 20 25 3J 3.5 4.0 45 5.0 TIME. SEC M0.RoewF 07/o$/06 ca.4i.34 I

I7" .. 16nu. rnu==w ws wai ,

2 VESEEL WATER LEVEL tina 150

[k "

l 82 s

/ N  % ,

N 10G I

15 t t 25 0 00 J.5 0.5 10 1.5 2.0 25 TINE. SEC 30 40 45 50 l 3

l no. n wr eues/ .......

I Figure 6.1 System Traces, LRNB 104.2/100 -

i I

I XN-NF-86-36 I 35 Revision 3 I

14c i. nsvinun ru n 66166 y y y 2 HEAR FLui

]

3 RECIRCULAfl0N FLOW 4 WESSEL STE4t FLOW I20 a. ratzum an rLuu

, . ,,

• W *

. . . s h

E80 60 g ly 1 =

I *

'kL. "

k T u . ,

Y .s s s O O 36 40 I

4 8 12 16 20 24 28 32 f!NE. SEC s(0. eroaoss es/os/se on.ee.3s.

I .

I 16C i. 16 .66 ruunun6 wnmws irm 2 VESEEL WATER LEVEL (!NI n

12C 80

\

I t

, , , , _)

~

I -4C

\

s

-8c

00 4 8 12 16 20 24 28 32 36 40 TIME. SEC 3(0. OF02039 De/Os/M 09.44.34 I Figure 6.2 System Traces, FWCF 104.2/100

I 36 XN-NF-86-36 Revision 3 7.0 ANALYSIS OF EXTENDED OPERATING CONDITTONS I 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.

I 7.1 Desian Bases For Thermal Marains I 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 covered the operating power-flow map depicted in Figure 7.1. In support of I 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.

Eleven power-flow statepoints were defined for validation of the operating power-flow map. These statepoints are described in Section 7.2, below.

Representative. system transients were evaluated at each of these confirmatory I statepoints to demonstrate the applicability of the existing power- and flow-dependent limits.

7.2 Determination Of Analytical Stateooints

'I Specific power-flow statepoints 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 statepoints in the ELL and ICF transient analyses. The ICF and ELL analytical statepoints are also shown in Figure 7.1.

Base analyses were performed at 104.2% of rated power and 100% of rated recirculation flow. These analyses are discussed earlier in this report.

I

37 XN-NF-86-36 I

Revision 3 Additional analyses were performed at 104.2% of rated power and 108% of rated fl ow. These analyses define the power and flow extremes attainable during normal operation. Analyses were also performed at 104.2% of rated power and 73.8% of rated flow to assess margin effects at the lowest flow-rated power statepoint.

Analyses were performed at 100% of rated flow and 70% and 40% of rated power to confirm the power-dependent MCPR limits established by the generic control rod withdrawal error analysis. Analyses were also performed at 108% flow and 40% power to verify that the trends from the reduced power analyses at 100%

flow remain consistent at increased flow conditions.

Analyses at 25% of rated power were performed to validate the power-dependent MCPR limit function. Analytical statepoints were chosen at 73.8% flow and 40%

flow to verify both branches of the MCPR(p) curve.

Analyses were perforned at 70% power and 40% flow to characterize the upper t

left corner of the power-flow map.

I 7.3 Analysis Of Plant Transients Under ELL Conditions The LRNB and FWCF transients were evaluated at each of the statepcints g

identified for further analysis. The LFWH and CRWE generic analyses 3 summarized in Section 5.0 already cover the extended operating power-flow map.

Results of the LRNB analyses at the ELL statepoints are shown in Table 7.2.

These analyses are consistent with the analyses performed in support of operation at rated conditions. For the events initiated from 40% power and Lelow, the scram trips associated with turbine control valve position and g

turbine stop valve position and recirculation pump trip (RPT) were assumed to 5 be overridden.

I I

I

l l

, l 38 XN-NF-86-36 Revision 3 I Results of the FWCF analyses at the ELL statepoints are shown in Table 7.2.

l l

These analyses are , consistent with the analyses performed in support of operation at rated conditions.

System behavior during representative analysis in the ELL region is illustrated in Figures 7.2 through 7.5.

I 7.4 Analysis Of Plant Transients Under ICF Conditions I Additional power-flow statepoints were identified and analyzed to support operation at greater than rated recirculation flow. Although the major effect of increased core flow operation is in the area of the LOCA evaluation, the power-dependent MCPR operating limits were investigated for validity at I increased core flow.

The LRNB and FWCF transients were evaluated at 108% flow at the highest and low'est power levels allowable under the increased core flow operating map.

Analyses were undertaken at 104.2% of rated power to support full power operation and at 40% of rated power to verify the lower bound of the limit curve.

I The results of these analyses were compared with the power-dependent MCPR limits determined in Reference 9.

Analytical results are summarized in Table 7.2 along with the ELL results.

Representative system traces are given in Figures 7.6 through 7.9.

I 7.5 Validation Of Off-Nominal MCPR Limits I Figure 7.10 shows the Cycle I flow-dependent MCPR limits and the results of

! the ENC analyses supporting these limits. Results of the flow-dependent MCPR 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 continued applicability of the limits.

lI l

lI

T B \

39 XN-NF-86-36 I

Revision 3 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 l

ELL and ICF transient analyses.

1 I l

I I

l I

l 1

II l

I-I 3

I!

I Il Il Il L

I l

I 40 XN-NF-86-36 Revision 3 r

I TABLE 7.1 I ANALYTICAL STATEP0INTS FOR ELL AND ICF EVALUATION FEEDWATER DOME CORE CORE INLET I POWER

(%)

FLOW 171)_

TEMPERATURE (dearees F)

PRESSURE (osia)

PRESSURE fosia)

ENTHALPY (BTU /lbm)

ELL STATEPOINTS I 104.2 104.2 70.0 100.0 73.8 100.0 425 425 391 1060 1060 989 1071 1071 994 530.2 522.0 525.2 70.0 40.0 391 989 994 499.2 I 40.0 40.0 40.0 100.0 50.0 40.0 306 306 306 954 954 954 956 956 956 524.0 510.8 504.3 I 25.0 25.0 73.8 40.0 254 254 944 944 945 945 522.9 512.3 ICF STATEPOINTS I 104.2 40.0 108.0 108.0 425 306 1060 954 1071 956 531.9 524.9 I

~

I I

I

. i

I 41 XN-NF-86-36 I

Revision 3 TABLE 7.2 RESULTS OF TRANSIENT ANALYSES AT ELL AND ICF STATEP0INTS l

INIT INIT PEAK PEAK PEAK DELTA

! EVENT POWER FLOW POWER HT FLX PRESSURE CPR MODEL

%rtd 5ttd d Ett_d_ %rtd osia ELL ANALYSES l LRNB 104.2 73.8 191.8 101.5 1231 0.05 COTRANSA i

LRNB 70 100 103.8 69.1 1192 0.04 COTRANSA LRNB 70 40 70.0 67.3 1185 0.16 COTRANSA LRNB 40 100 142.4 48.7 1143 0.27 COTRANSA l LRNB 40 50 189.7 49.1 1126 0.25 COTRANSA 3 LRNB 40 40 207.4 50.9 1170 0.29 COTRANSA LRNB 25 73.8 60.9 44.4 1180 1.05 COTRANSA LRNB 25 40 86.6 38.3 1174 0.80 COTRANSA FWCF 104.2 73.8 109.5 106.5 1189 0.06 COTRANSA FWCF 70 100 73.3 70.9 1071 0.03 COTRANSA E FWCF 70 40 74.9 71.7 1064 0.15 COTRANSA 5 FWCF 40 100 -

69.6 42.8 1009 0.07 COTRANSA FWCF 40 50 48.0 42.4 989 0.13 COTRANSA FWCF 40 40 42.0 40.8 987 0.09 COTRANSA FWCF 25 73.8 32.5 26.9 974 0.08 COTRANSA FWCF 25 40 27.8 25.5 966 0.07 COTRANSA ICF ANALYSES LRNB 104.2 108 144.6 101.5 1232 0.02 COTRANSA LRNB 40 108 153.3 48.7 1137 0.28 COTRANSA FWCF 104.2 108 110.1 107.4 1198 0.04 COTRANSA FWCF 40 108 76.5 43.6 1012 0.09 COTRANSA I

I Peak heat flux value near time of minimum delta CPR.

I I

FIG. 7.1 OPERATING POWER-FLOW MRP

) 120 -

i v v v

) 100 -

ELL o Region

1. ~ e ion a /

1 y 100% Rod Line i r 3 N

o l b4 d

W 2

W 40 - " V '

MV e

O Low Pump Speed i

Minimum Flow 20 -

\ "

/ Low Pump Speed Maximum Flow a?

$ la do 3o lo s'o s'o 7o do do 150 1Io Sh wm CORE FLOW, % OF RATED

43 XN-NF-86-36 l Revision 3 24C . navinun rLim 6trt6 2 MEA r FLUX

3. rec lRCULAll[pN FLOW

4. VESSEL STEM 1 FLOW 200 a. ettmita eLuv l

16G e

7 El2C

) I

$i i . 2 <

l 3

1 \ m,\ ~

W W

. i T

s mA ,N

^

40

(

8

( ^j , , , , , ,

i I

~4h.0 05 1.0 1.5 20 25 30 35 40 45 5.0 Tit 1E. SEC Ko. st03crf erros/es o,.os.oi.

Il lIS . vtapt6 rrstapunt 6ninwt tr a i J

2. VESSEL WATER LEVEL (110 M '

l 159 i

e  ; sg 120

, K. i l

l l

• l l

l i l '

10G' g l

l 75 ,

/  : ,

I

' i 2 .

50 2 2 2 2 i f f 25 ,

' l

, t .

5.0

~

0 0,0 0.5 i.0 i.5 20 2.5 3.0 35 4.0 45 l l

TIr1E. SEC Ste. st030Ff 07/05/84 0,.09 4 .

Figure 7.2 System Traces, LRNB 104.2/73.8 1

I I 44 XN-NF-86-36 Revision 3 ISC i. rc ww= rL- 661u 2 HE FLUX 3 RE RCtJt. Af fCN FLOW y 3

4. VEsbEL STEMt FLOW 12- f a. rusva i r.m eLW

.I /

IOC I n 5

ETS .a J ,

g '

l I 7 j'

N (itf%

% ~, ,

^

I

~

t s , , ,

~ . , ,

I 66 68 20 I -25 a 4 6 8 60 62 14 T!?tE. SEC sto. weseas is/es/es os.,s.se.

I I IOC 2

i u. asu rnsevns wn--w6 41.SBEL WATFR LEVEL tlNI tr a i ,

73 ,T m s M l

lI 25

/ 3 A

l ,

\

l

-$C

• I6 2 4 6 8 to 62 14 is 18 20 i

firt. SEC

" sto. wescas asies/es s3.ne.re.

Figure 7.3 System Traces, FWCF 70/40 i

I

45 XN-NF-86-36 Revision 3 62C ,, ,,y,,,, ,,, ,,,,,

2 HEAR Flux

3. RECIRCULATION FLOW

4. WESSEL STE&l FLOW 3 FLLJMA I LfL rLWu I

o NW&A Eso -' r NY

~

b

- t 3 40

M 2-fs,

,0 \

QAwdA

-26 2 4 6 e i0 i2 i4 e6 is 20 TIrtE. SEC sco. aiocar ..iosia . . .. ..

28C . vs>>sk russavns wrv . s tra i s

2. vtSEEL WATER LEVEL (IN3 3 24C 200 l

16G l

t2C I

80 40 -- ww --- w 0 0 2 4 6 8 to 12 14 16 18 20 firt. SEC W us. a iocar isiosion oi.ss as.

I Figure 7.4 System Traces, LRNB 25/73.8 I

1

I I

L 46 Mi-NF-86-36 Revision 3 12G i. navinun r6u Ltict 2 NE M FLUX 3 RECIRCULAflDN FLois

4. WESEEL STEM FLOW 100 a. retsustn r tvu 80 o

i I E60 L

/\ ssh_1 ? ,

W y ~tY" s .

s s 5 5

% l' yw -

~26.0 1.0 2.0 30 4.0 5.0 6.0 1.0 8.0 9.0 10.0 fir 1E. SEC sto. stiions is/osias io.*i.rs.

F L

28C 1. 1suc6 rn6 m e s wne,. w c ir a . ,

r 2. VESEEL WATER LEVEL (IN)

L 24C b

L 20u- 7

, /

I6C

/

12C ,

, /

- /

Y

(

I ,-  ?

• e t  : a ~

40

. 0 0.0

/ 1.0 20 3.0 4.0 S.0 60 1.0 8.0 9.0 80.0 fir 1E. SEC sto. seisons is/os/es is.*i.as.

Figure 7.5 System Traces, LPNB 25/40

I -

l 47 XN-NF-86-36 l Revision 3 -

15C i.

l cuinun r6u u v u.

2 MEAI FLUX

3. REclRCULAftDes FLOW l 4.

g i

12" VESSEL STE S FLOW

a. rusvason rLuu 10G s _ 4, o i N

s 275 i for =h_

)

r -

~

?

h50

//  %$

=

l

-i , , , , ,

I

~ b.0 05 10 15 2.0 25 3.0 35 40 45 50 Tir1E. SEC sto. steinen oF/os/as oo.ss.si.

I 17; i. vam6 rnueva6 wnmws ir a di

2. VESBEL HATER LEVEL (110 ISC

_t 123 N 10C 75 i

5

~ ~

. , , , , , , , I 25 l

0 00 0.5 1.0 l.5 20 25 30 3.5 4.0 4.5 5.0 g TIP1E. SEC g sto. eLet004 otros/se oo.ss.ss.

l Figure 7.6 System Traces, ' RNB 104.2/108 .

I l

I L

I I 48 XN-NF-86-36 Revision 3 14C -c o n., esws g6,66 3

g , 2 MEAf Ft.UX J. RECIRCUi Af! Die FLOW

4. VESBEL STEst r.0W 14C 3 ,

reti.m en iLuv

/ s s , _ ' '

fa. -

100 -

a esl0 b

l ;

p , ,o I \y 3

\

40 -

(

AO N

b . .

, Yd s g I -0 4 9 12 it 23 firT. SEC 24 SES. WetGet 28 32 Is/01/86 JG 43

22. e t . t F.

160- ,, , , , , , , , , , , ,,,,, ,,,,,,,,73,,

2 VEstEi DATER LEVEL (IM) 12-90 ,

I 40 p M i

p

, , , . I L T e

4C

\

-80 "Ib

• 8 IJ lu 20 24 28 I

J2 J6 40 TIME. SEC sto. weises esios/es 22.es.ir.

I Figure 7.7 System Traces, FWCF 104.2/108

I 49 XN-NF-86-36 Revision 3 24G i. mswinun r 6u 5 61 r.6 2 HEAR FLUX

3. REclRCULATIDN FLOW .

4. VESSEL STDM FLOW 200 a. retJuasta rLum 16C 8 h 2 ' )

1. 12C g N 80 t / _ s -

40 --

%g ,

~

\A _

_'_, d. _ A ___ . f '

~46.0 1.0 2.0 J.0 4.0 50 60 7.0 80 90 10.0 T!f1E. SEC KO. SLeeces 14/06/M 10.04.82 15C ,, ,c au rm.m wri,m vm ,

BEL WATER LEVEL tlu) 12_. -

IOC 75 I

50 25 s '

\

~ I

-26.0 00 2.0 30 40 50 6.0 1.0 8.0 9.0 10.0 fil1E. SEC =

KO. OLeece4 16/03/M 10.04.52 Figure 7.8 System Traces, LRNB 40/108

I 50 XN-NF-86-36 Revision 3 I ISO J.

HE F,QX l J. aEco cAar a n o,

a. vE 9Eu stE m r. os I ,,3.

'* PLLJ d Ti LM PLVd 8

Ah% + 3 100-e 7

I =75 U

E I I $"$3 a i i/r i e dj (I'b' 25 I ~

_M.  ;

b. C _ , a E '#b d 8 le 16 23 24 28 32 J6 40 f!ME. SEC I

I ;00 . vo>t. -n c nun. vne,vc tr a i, 2 fE55E*. WATER 6 EVE ~. (INI A

g

. CO - s M

40 20

. . -h I -20 s

\ i ...L -.*

, i_

g gg, -

46 4 8 12 16 20 24 28 $2 36 40 flFE. SEC ts.36.32 i Ste. sFosoir I:/0*/86 g Figure 7.9 System Traces._fRCE. 40/108 I

FIG. 7.10 FLOW-DEPENDENT MCPR LIMIT VALIDATION

1. 7 -

l + Flow increase analyses, 107% max flow

1. 6 - x Flow increase analyses, 102.5% max flow A Cycle 2 transient analyses

1. 5 - +

- 1. 4 - 107% MAXIMLN FLOW + us

~

l b 102.5% MAXIMUM FLOW e

Q_

+

r 1.3 -

X

+

1.2 - k

+

x 3 3

A + 3 g x 1.1 - A A $is.

x+ 74 I$

1 0

20 30 40 50 60 70 80 90 100 110 120 wk 10 CORE FLOW, % OF RATED aus Ims una sus een ams man um age sus aus aim amt am um suu m num ses

M M W M M M M M M M M M M M M M M M M l

l

! FIG. 7.11 POWER-DEPENDENT MCPR LIMIT VALIDATION

2. 4 -

a Transient analyses

2. 2 - Core Flow >50%

a 2-

\ -

core riow <50%

a N

- 1. 8 - m cc Q_

O 1. 6 -

1. 4 -

A EE

1. 2 - A 5.1 A $

a 8?

"W 1 i , , , , i i i i i i i i 0 10 20 30 40 50 60 70 80 90 100 110 120 CORE POWER, 7. OF RATED -

I 53 XN-NF-86-36 Revision 3 i 8.0 ASME OVERPRESSURIZATION ANAI.YSIS I Maximum system pressure has been calculated for compliance with the ASME Boiler and Pressure Vessel Code. This overpressure evaluation postulated the I 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 g internal pressure from reaching component overpressure limits. The analysis E assumed that seven of the twenty safety / relief valves were out of service.

8.1 Desion Basis The reactor conditions used in the overpressure analysis are summarized in Table 8.1. Because not all of the phenomena contributing to the overpressure effects are clearly conservative at the same power-flow statepoint, the overpressurization analysis was performed at three different statepoints at 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.

The containment isolation event was selected for the overpressure analysis because the single failure assumption (the direct scram on MSIV closure) makes I 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 i overpressure analysis was performed with COTRANSA (Ref. 2).

8.2 MSIV Closure Analysis 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.

I I

I

I 54 XN-NF-86-36 I

Revision 3 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 the reactor vessel during this postulated event was 1281 psig (1296 psia), g which is less than 110% of the design pressure of the reactor vessel. The a maximum pressure in the steamlines was calculated to be 1253 psig (1268 psia).

The maximum pressure calculated to occur in the steam dome was 1265 psig (1280 psia).

8.3 Steam Dome Pressure Safety Limit 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 3 of the existing 1325 psig pressure safety limit is conservative relative to this comparison.

l 8

I I

I I

I I

I

I 55 XN-NF-86-36 Revision 3 i TABLE 8.1 I REACTOR AND PLANT CONDITIONS ASME OVERPRESSURIZATION ANALYSIS 100% FLOW CASE Reactor Thermal Power 3994 MWt (104.2% of rated)

b. f rated)

Core Active Flow 100.7 M1b/hr Core Inlet Enthalpy 530.2 BTV/lbm Vessel Pressures I Steam Dome Upper Plenum Core 1060 psia 1070 psia 1077 psia Lower Plenum 1095 psia I 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 I

I I

I I

I I

. I 56 XN-NF-86-36 Revision 3 TABLE 8.2 I

RESULTS OF ASME OVERPRESSURIZATION ANALYSIS MAXIMUM PRESSURE MAXIMUM POWER FLOW POWER VESSEL DOME STEAMLINE

(%) (%) (%) (osia) (osia) (osia) 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 1268 I

I I

I g.

! I I

I I

I I

57 XN-NF-86-36 Revision 3  ;

I 60C s.

utvi n un r um 6Lru I

2. HEAR Flux

3. REclRcutAf!Du Flow 50C ** 'E55E' 8IE" FLOW 3= P LL JUAI LM rLyg 40C S

7 ac30C ci E

I 0200

,, .. . . , ,, .. 2m .,m I ww.,

. ( , , , ,

I ~

0 10 20 3.0 4.0 50 60 7.0 40 90 10.0 ste. easaerc os/os/se er.as.37.

I I 20G s. 1samu ruunun6 wn, 2 VEsnL WATER LEVEL ws tr a i tlui s

200 g

[ a 16C N

E 12C S0 i f

40 - "

0 0.0 I 10 20 30 40 50 fir 1E. SEC 60 7.0 see. saasers 00 oe/os/se 9.0 10 0 it.os.sr.

Figure 8.1 System Traces, ASME Overpressurization Analysis I _

I 58 XN-NF-86-36 Revision 3 9.0 SP_ECIFICATION OF OPERATING LIMITS I The operating limits for Cycle 2 operation of Grand Gulf Unit I are based on the Cycle 1 operating limits. These limits are based on limits provided in I References 9 and 10 which have been modified as required.

9.1 MCPR Ooeratina Limits The power- and flow-dependent effects which were assessed in the transient analyses described in this report have been assembled into discrete limits on I operating MCPR values. At any allowable operat.ing 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.

9.1.1 Flow-Decendent MCPR Limit I 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 terminates 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 runaut capabilities of 103.5% and 108%, 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.

I I

lI

I 59 XN-NF-86-36 Revision 3 9.1.2 Power-Decendent MCPR Limit I

The power-dependent MCPR operating limit is determined from the generic CRWE analysis and the plant-specific analyses of LRNB and FWCF transients at E

representative conditions blanketing the operating power-flow map. Observance 5 of this limit for esch 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 g

fuel types. 5 9.2 MAPLHGR Ooeratino 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 MAPL}iGR limits are established from the steady state design LHGR limit as described in References 12 and 13.

Power- and flow-dependent factors were assigned to the MAPLHGR limits for protection against fuel damage mechanisms associated with the initiation of transients at off-nominal conditions.

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.

9.2.1 Flow-Decendent MAPLHGR' Limit Multiolier I

The MAPFAC(f) operating limit factor was evaluated for ENC fuel in Grand Gulf through analysis of flow increase transients. The results of the ENC analysis l l I

60 XN-NF-86-36 Revision 3 I which are reported in Reference 12 developed the flow-dependent MAPLHGR limit factor given in Figure 9.3.

9.2.2 Power-Decendent MAPLHGR Limit Multiolier The MAPFAC(p) operating limit factor was validated for Cycle 2 operation of Grand Gulf Unit 1 through the analysis of the FWCF, LRNB, and CRWE transients.

The transient analyses which were used to ev-luate 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 increar 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 damage threshhold LHGR value at which the fuel may be operated prior to the transient without reaching the threshhold during the transient.

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 threshhold LHGR limit may be determined by increasing the LHGR fraction from the previous paragraph by a factor of 1.20.

The FWCF and LRNB results of the MAPFAC(p) determination are shown in Table I 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 l analyses and comparisons validate the existing limits for Cycle 2 operation with ENC fuel.

I lE

I

I 61 XN-NF-86-36 Revision 3 TABLE 9.1 I

MAPFAC(p) ANALYSIS RESULTS E0XEH ELOR EYIRI HEAT FLUX FACTOR MAPFACfo) 104.2% 108% FWCF 1.04 >1.000 104.2 100 FWCF 1.02 >l.000 104.2 73.8 FWCF 1.03 >1.000 70 100 FWCF 1.01 >1.000 70 40 FWCF 1.03 >1.000 40 108 LRNB 1.37 0.87 E 40 100 LRNB 1.37 0.88 3 40 50 LRNB 1.33 0.90 40 40 LRNB 1.36 0.88 25 73.8 LRNB 1.90 0.63 25 40 LRNB 1.66 0.72 I

I I

I I

I I

I I

I

som ase sus == sus som amm aus num aus == men sem uma uma amm aus am 1

1 l

FIG. 9.1 F' LOW-DEPENDENT MCPR LIMIT l

1. 7 -

i l

1. 6 -

I 1

i 1. 5 -

107% MAXIMUM FLOW 102.5% MAXIMui FLOW

1. 4 - cn

~

u e

Q.

o r 1. 3 -

1. 2 -

1i_ FM 5.1

! r. 7' 88 w a m

1 , , , , , , , , , , , ,

0 10 20 30 40 50 60 70 80 90 100 110 120 CORE FLOW, % OF RRTED

FIG. 9.2 POWER-DEPENDENT MCPR LIMIT

2. 4 -

• ~

ore Flow >50%

2-Core Flow $50%

-- 1. 8 - 0 5

25 r 1. 6 -

1. 4 -

1. 2 -

??

1 ' i i a i ' ' ' ' '

O 10 20 3O 40 50 60 70 80 90 100 110 120 wM CORE POWER, % OF RRTED ams uma uma em uns aus sus sem aus uma aus aus aus ums uma anal asas sem e55

ll lll ,

ll I ' I m

e '

uh?b su Rk8" s

s u 0 s 2 R 1 a O mu T 0 1

C 1 R WW m F W W L L

. O O os O

L L O F F 5

0 F F X X 1 A A s R X A

X A

M M u G M M % 5% 5 s H  %%

7 7 2 2 d

o L 0 0 0 0 1 1 1 1 s

u P C C N E N E a R E G E G doDE M T u A e

s T O R

N '7 F s

u E O a D 'o 7 N 6 s E W n

u P O E 0 L

's F s D E u - R s W oO dC O

L a F [ $

0 u

s 3 o

. d e

s 9 e a

. 'l n G a I F O 9 8 7 6 m 1 1

u 1 0 0 0 0

- S t tg (f

m s

a m

l l1

FIG. 9.4 POWER-DEPENDENT MRPLHGR FRCTOR 1.1 -

A Transient analyses 1- A

0. 9 - A e

a g 0. 8 - g

@ Core Flow 550% -

Power >40%

Power 540% 3 All Core Flows c 0. 7 -

r A

i 0. 6 -

Core Flow >50%

Power $40%

i 1

0. 5 -

E5 kh 0.4 , , , , , , , , , , , , S$

0 10 20 30 40 50 60 70 80 90 100 110 120 wg CORE POWER, 7. OF RATED l

aus sus som sum uma sus sus mas ame uma sus aus aus aus sus. ama amm amm aus

.=

66 XN-NF-86-36 Revision 3

10.0 REFERENCES

I 1. " Exxon Nuclear Methodology for Boiling Water Reactors: THERMEX Thermal Limits Methodology, Summary 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 I (November 1981), as supplemented.

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

I XN-NF-524(A), Revision 1, Exxon Nuclear Company, Richland, Washington (November 1983).

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

Nuclear Co~mpany, Richland, Washington (March 1981).

5. Paul N. Somerville, " Tables for Obtaining Non-Parametric Tolerance I Limits," Annals of Mathematical Statistics, Vol. 29, No. 2 (June 1958),

pp. 599-601.

"A Generic Analysis of the Loss of Feedwater Heating Transient for I 6.

Boiling Water Reactors," XN-NF-900(P), Exxon Nuclear Company, Richland, Washington (February 1986).

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

I

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).

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

Electric l 10. "GGNS Single Loop Operation Analysis," General Electric Company, San 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).

I 12. " Grand Gulf Unit 1 Cycle 2 Reload Analysis," XN-NF-86-35, Revision 3, Exxon Nuclear Company, Richland, Washington (August 1986).

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

I I -

I l 67 XN-NF-86-36 I

Revision 3 l

! 14. " Grand Gulf Unit 1 LOCA Analysis," XN-NF-86-38, Exxon Nuclear Company, I

l Richland, Washington (June 1986).

15. "XCOBRA-T: A Computer C-de for BWR Transient Thermal-Hydraulic Core Analysis," XN-NF-84-105(P), Exxon Nuclear Ccmpany, Richland, Washington (May 1985).

I

, I I

l l I l

I

. I I

I I

I l I l

I I

I

I I A-1 XN-NF-86-36 Revision 3 I

APPENDIX A I A-1.0 COTRANSA HOT CHANNEL MODEL INTRODUCTION AND

SUMMARY

I The original COTRANSA hot channel model was used to give a figure of merit delta-CPR used to determine the limiting transient. The limiting delta CPR was then determined by the user using a cumbersome XCOBRA-R0DEX2-HUXY manual iteration. This involved a time consuming calculation where the user I manipulated a considerable amount of data between codes. Furthermore, it also resulted in a transient condition being analyzed using steady state approximations. The COTRANSA hot channel model has been modified to automate the delta CPR calculation and to give a transient delta CPR. Each fuel type is modeled, and a delta CPR specific to that fuel type is determined. XCOBRA and RODEX2 are used to determine the input for each hot channel. COTRANSA then calculates the delta CPR for each time step. The largest delta CPR is I then reported.

A-2.0 MODICICATIONS TO THE HOT CHANNEL MODEL A-2.1 Flow Resoonse Surface I The modifications to COTRANSA include a time dependent calculation of the flow rate to the hot assembly of each fuel type. The initial and transient flow to the hot channel is determined using XCOBRA, ENC's approved subchannel code for BWR's. A steady state response surface for the hot assemblies' flow rates are determined for four key variables:

I I

I I

I A-2 XN-NF-86-36 I

Revision 3 o Relative assembly thermal power o Core average thermal power o Core average active flow o Core pressure A quadratic equation is then determined for each hot assembly flow.

A-2.2 Fuel Temperature fiqdal The fuel temperature model for the hot rod is as described in the approved HUXY (XN-NF-79-71, Rev 2) with the clad gap conductance based on RODEX2 calculations. Each fuel type is run to the end of cycle, at the end of cycle the power is increased and the relationship of gap conductance to average fuel temperature is then determined. This gap conductance is then used in the hot channel model.

A-2.3 CRITICAL POWER RATIO CALCULATIONS l The MCPR calculation model used in the hot channel model is the approved XN-3 I

correlation as described in XN-NF-512, Rev 1. The hot channel model calculations do not interact with the core average solutions since the impact of the hot assembly is so small. Therefore, the boundary conditions which g

drive the hot channel model are stored and used iteratively. These boundary 3 conditions are:

I I I I

' I I

l I

I I A-3 XN-NF-86-36 Revision 3 I o Power I o o

Core inlet enthalpy Pressure o Inlet flow rate o Outlet flow rate o Bulk fluid temperature o Clad to fluid heat transfer coefficient o Heat flux o Axial power distribution o Enthalpy rise The purpose of the calculation is to determine the maximum allowable assembly power which will not exceed critical heat flux conditions during the transient.

The initial power used in the calculation is only an estimate. After the completion of the transient simulation the lowest calculated CPR is compared to 1.0 and the power of the fuel rod is modified. This new power is assumed I as an initial condition. The flow to the limiting assembly is determined from the response surface and the enthalpy rise is adjusted to be consistent with the new conditions. The hot channel model calculations are repeated and the

> lowest CPR is again compared to 1.0. The process is then repeated until the l lowest CPR is 1.0. The initial CPR minus the lowest CPR is the delta CPR for the transient consistent with ENC's reported methodology.

A-3.0 VERIFICATION OF THE HOT CHANNEL MODEL Two different checks were made to insure the adequacy of the COTRANSA hot channel model's delta CPR calculation. The standard XCOBRA - RODEX2 - HUXY iteration was performed, and steady state conditions were input into the hot

! channel. The XCOBRA - RODEX2 - HUXY iteration resulted in delta CPR's that I were more conservative than those for the hot channel model. These results were as expected because of the steady state nature of the method. When I

I .

I l

t A-4 XN-NF-86-36

! Revision 3 quasi-steady state conditions were forced into the hot channel, the resbits of I

the comparison were in close agreement.

l l

I I

I 1 I I

l l

I I

I I

I I

I I

I I B-1 XN-NF-86-36 Revision 3 I APPENDIX B CONFIRMATION OF COTRANSA HOT CHANNEL DELTA-CPR RESULTS WITH THE XCOBRA-T TRANSIENT THERMAL-HYDRAULIC MODEL I B-

1.0 INTRODUCTION

I During the internal review of the analyses reported in the body of this document, a potential nonconservatism was identified in the formulation of the I COTRANSA Hot Channel model used to calculate delta-CPR values. The impact of this potential nonconservatism was evaluated through application of XCOBRA-T (Reference B-1). This evaluation confirmed the COTRANSA Hot Channel results.

This appendix describes the confirmatory analyses and presents their results.

I B-2.0 XCOBRA-T BASES This section is a summary of Reference B-1.

XCOBRA-T has been designed to be a state of the art analysis tool for the

, transient thermal hydraulic response of BWR fuel bundles. As such, it employs fuel rod modeling that can be as detailed as in RODEX2 (Reference B-4) l analyses, if necessary, and it employs transient hydraulic modeling totally lg consistent with ENC's COTRANSA code (Reference B-5). A major assumption is 5 that the empirical XN-3 CHF correlation (Reference B-3) can be used with the local conditions that occur in transients. Sonic effects are not modeled, which is consistent with the non-LOCA transient events for which XCOBRA-T is intended.

I The basic analytical models in XCOBRA-T, as described in Reference B-1 are:

the transient fuel rod conduction model, and the transient mass, momentum and I energy equations for hydraulic response. These models and equations are basically the standard textbook type models used elsewhere by ENC and throughout the nuclear industry. The empirical relationships which are I

.I 1

I B-2 XN-NF-86-36 Revision 3 options of the code are: (1) the pellet-to-clad gap heat transfer which is I

modeled after ENC's state of the art RODEX2 code, (2) the subcooled void and void-quality hydraulic models which are used elsewhere in ENC's BWR methodology, and (3) the XN-3 critical power correlation.

Code-to-code benchmarking of XCOBRA-T with COTRANSA has verified the coding in XCOBRA-T from a transient hydraulic standpoint. Similar benchmarking against ENC's HUXY code (Reference B-6) has verified the transient fuel rod conduction model and benchmarking against XCOBRA has verified proper implementation of the XN-3 correlation. In all cases, the code-to-code benchmarks have shown excellent agreement.

Independently, XCOBRA-T has been benchmarked against transient critical heat flux tests performed at Columbia University for ENC. The results show a consistent measure of conservatism for the code with respect to the prediction of transient critical heat flux, and hence justify the application of the XN-3 correlation to transients.

B-3.0 XCOBRA-T METHODOLOGY FOR EVALUATION OF DELTA-CPR The following text describes how XCOBRA-T is used to evaluate the limiting assembly delta-CPR for BWR transient events that are analyzed with EfjC's COTRANSA code. The COTRANSA code is used to determine overall BWR system response for anticipated operational occurrences. The methodology for this system response evaluation for various transient events is detailed in Reference B-5. For the purpose of determining overall system response, the core hydraulic mcdel in COTRANSA consists of two hydraulic channels: the core average active hydraulic channel and the bypass channel. These are parallel l flow path chanc.els that share common lower plenum and upper plenum boundary I

conditions on pressure and enthalpy but otherwise are independent of one another. The hot channel model does not exert feedback on the average core model.

l l I I

I I B-3 XN-NF-86-36 Revision 3 I Conceptually, the core average active channel in ENC's transient methodology I can be decomposed into a large number of parallel flow channels, each of which are associated with the active flow region of individual fuel assemblies.

Each of these channels would share the common lower and upper plenum boundary conditions of the BWR system model but would otherwise be independent of one another. This decomposition is not important from the standpoint of overall BWR system response and hence is not done in COTRANSA. However, because of power level and fuel design differences, this decomposition is necessary for I an accurate evaluation of limiting assembly thermal hydraulic response during a transient. This decomposition is accomplished using XCOBRA and XCOBRA-T in the evaluation of delta-CPR.

The basic physical models governing steady state and transient hydraulic response are identical between XCOBRA-T and COTRANSA. For steady state response this hydraulic modeling is also identical to other codes in ENC's BWR i methodology, namely XCOBRA and XTGBWR (Reference B-7). In particular, the i.evy subcoooled void modeling and the Zuber-Finley void quality relationships are common to all the above codes. There are differences in numerics between l the various codes; however, very good agreement has been shown in code-to-code benchmarking. Fuel rod to coolant heat transfer is accomplished via a conduction model in XCOBRA-T in which the fuel rod is divided into radial rings and the temperature dependent thermal conductivity of the urania fuel pellets and the thermal conductivity of the zircalloy clad are considered.

Also considered are the pellet-to-clad gap and fuel rod surface heat transfer coefficients. These features are common with the COTRANSA and HUXY codes.

!I Numerics for the conduction solution are different between XCOBRA-T, COTRANSA and HUXY, but code-to-code benchmarking has shown very good agreement.

lI l

ENC's BWR transient methodology assumes the core radial power distribution to be constant during core wide transient events. This has been shown to be j acceptable based on the Peach Bottom turbine trip tests. Thus, the initial l radial power distribution, along with the transient variation in total core I

power and the axial power distribution determined in the COTRANSA BWR system

'I

I l

I B-4 XN-NF-86-36 Revision 3 response evaluation, determine the power response of the hot assembly.

Because of this and because the lower and upper plenum boundary conditions are common to the individual fuel assemblies of the active core, it is possible to evaluate hot assembly transient response in an analysis that is separate from the overall BWR system response analysis. This hot assembly analysis is accomplished using XCOBRA-T. The basic conditions for the analysis, namely the transient power and lower and upper plenum boundary conditions, are saved on a computer file by COTRANSA and this file is read by XCOBRA-T.

Aside from the COTRANSA generated boundary conditions other inputs to the XCOBRA-T calculation include:

1. Fuel hydraulic characterization.

I This includes bare rod flow area, wetted perimeter and I

component loss coefficients. These inputs are basically consistent with XCOBRA detailed core hydraulic analysis inputs.

XCOBRA-T also requires the channel entrance and exit areas t E

be specified. These areas should correspond to the flow area E at which point the upper and lower plenum pressure boundary conditions are applied. It is common to consider the entrance and exit in the XCOBRA-T calculation to be the bottom and top of the active fuel column and hence bare rod flow area would be l specified for both the entrance and exit areas. Because the l bypass flow is not modeled in XCOBRA-T, the inlet orifice loss E

coefficient is adjusted to give flow results identical to 5 XCOBRA.

2. Fuel rod characterization.

I I

Fuel rods are characterized with results from RODEX2 analyses. I In addition to the pellet and clad dimensions, a key factor is the pellet-to-clad gap coefficient. This corresponds to RODEX2 I

I

I B-5 XN-NF-86-36 Revision 3 predictions for the pellet-to-clad gap coefficient at the start I of the transient.

radial peaking.

In general, it is a function of assembly

3. Pressure bias.

Time varying boundary conditions saved by COTRANSA include the lower and upper plenum pressures at the middle elevation of I these volume . A hydrostatic correction is made to obtain the corresponding pressures at the entrance and exit boundaries of the XCOBRA-T model. The pressure corrections are constant with time and are based on COTRANSA time zero conditions. This constant correction is an approximation. Based on code-to-code comparisons of XCOBRA-T to COTRANSA, this approximation has not introduced appreciable error.

4. XN-3 parameters.

These are local peaking factors and S-factors that are used in both XCOBRA and XCOBRA-T.

iI I

The overall analysis method involves assuming an assembly radial peaking factor for the XCOBRA-T analysis. Next, the inlet loss coefficient is l

adjusted to achieve a time zero flow consistent with XCOBRA. Finally, the pellet-to-clad gap coefficient is input consistent with the assumed radial.

I XCOBRA-T is then run and the MDNBR (MCHFR) is determined.

B-4.0 GRAND GULF ANALYSES FOR CONFIRMATION OF HOT CHANNEL RESULTS 1

The transient conditions with the least margin to MCPR limits were reanalyzed with COTRANSA and XCOBRA-T. The ' choice of points to reanalyze was also g substantiated using the hot channel model of Reference B-5 as a figure of l5 merit determination. At the licensee's request, the COTRANSA reanalyses I

I . -

I B-6 XN-NF-86-36 Revision 3 included two changes in input parameters for as-built evaluation. These changes, which involved a decrease in the pressure drop in the main steamlines and a decrease in the enthalpy transport time in the downcomer, were made in the conservative direction relative to the methodology qualified in Reference B-5. With the exception of the analysis of the generator load rejection transient at 104.2% power, all of the recalculated delta-CPR values were within 0.01 of the original delta-CPR values tabulated in the main body of this report. All of the recalculated delta-CPR values were within the Cycle 1 limits which are being proposed as the MCPR operating limits for Cycle 2.

These limits are shown in Figures 9.1 and 9.2 of the main body of this report.

The XCOBRA-T analyses were directed toward confirming the delta-CPR values calculated by the COTRANSA Hot Channel model with these revised input parameters. This confirmation would also demonstrate the adequacy of the COTRANSA Hot Channel values calculated in the original analyses.

The reanalysis included load rejection (LRNB) evaluations at the 104.2/108, 104.2/73.8, 70/40, 40/108, 25/73.8, and 25/40 power-flow points and feedwater g l controller failure (FWCF) evaluations at the 104.2/108 and 104.2/100 E power-flow points. Feedwater controller failure transients were also evaluated without condenser bypass (FWCF/NB) at the 104.2/108 power-flow point and with a 100-degree feedwater temperature reduction (FWCF/RT) at the 104.2/100 power-flow point. These cases were the most limiting of the cases considered in the original analysis.

This confirmation of the COTRANSA Hot Channel delta-CPR values in the most limiting cases confirms that the less limiting cases analyzed with the COTRANSA Hot Channel model for Grand Gulf Unit I were acceptable.

The XCOBRA-T analysis was started by initializing the transient model to XCOBRA (Reference B-2) predictions of hot channel conditions at an initial critical power ratio (ICPR) consistent with the COTRANSA Hot Channel dalta-CPR value. The transient margin to critical heat flux was then determined through l I

I B-7 XN-NF-86-36 Revision 3 I XCOBRA-T analysis driven by boundary conditions calculated by the COTRANSA I average core model. If adequate margin to critical heat flux observed throughout the transient (CHFR ;tl.0), then the change in CPR predicted by XCOBRA-T is less than or equal to the COTRANSA Hot Channel delta-CPR. If critical heat flux were to be exceeded during the . transient (CHFR <1.0), then a higher ICPR would be assumed and the process repeated until adequate margin to critical heat flux was maintained throughout the transient.

I The results of the XCOBRA-T analysis are shown in Table B-1. This table identifies the transients analyzed and reports both the revised COTRANSA Hot Channel delta-CPR and the XCOBRA-T delta-CPR upper bound values for each event. The margin to critical heat flux is also provided as the difference between the lowest value of the critical heat flux ratio (CHFR) observed in the XCOBRA-T analysis and prediction of local critical heat flux (CHFR =

1.00). In this analysis, CHFR is the ratio of the critical heat flux as determined by the XN-3 correlation (Reference B-3) to the local heat flux as calculated by XCOBRA-T.

With the exception of the LRNB analysis at the 104.2/108 power-flow point, all -

of the XCOBRA-T analyses sh w margin to critical heat flux using the COTRANSA Hot Channel delta-CPR values. At the 104.2/108 power-flow point, the XCOBRA-T analysis was performed at a slightly higher (a value of .09 which was .01 higher) delta-CPR than the COTRANSA Hot Channel result, and margin to CHF was evident in the XCOBRA-T results. The COTRANSA Hot Channel delta-CPR results are conservative relative to the XCOBRA-T results for all of the analyses I except for the LRNB 104.2/108 analysis, which agrees within 0.01 delta-CPR (which is within the Cycle 1 limits). The COTRANSA Hot Channel results are supported by the XCOBRA-T analyses.

The tabulated results demonstrate close agreement between the COTRANSA Hot Channel model and XCOBRA-T for all of the FWCF cases. In the LRNB cases, the COTRANSA Hot Channel model shows conservatism as a function of the initial I power-to-flow ratio. At power-to-flow ratios falling above the nominal 100%

I I

I B-8 XN-NF-86-36 I

Revision 3 rod line (104.2/73.8 and 70/40), the XCOBRA-T analysis indicates considerable convervatism in the COTRANSA Hot Channel results. Analysis at statepoints at which the initial flow percentage exceeds the initial power percentage indicates less conservatism; in all cases, the XCOBRA-T results either confirmed the COTRANSA Hot Channel results as conservative or were within 0.01 delta-CPR.

The operating MCPR limits identified in Section 9.0 of the main body of this report are supported by the COTRANSA and XCOBRA-T analyses documented in this appendix; the COTRANSA Hot Channel model results of the Grand Gulf analysis were confirmed by the XCOBRA-T analysis.

B-

5.0 REFERENCES

B-1. "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," XN-NF-84-105(P), Volume 1, Exxon Nuclear Company, Richland, Washington (May 1985).

B-2. " Exxon Nuclear Methodology for Boiling Water Reactors: THERMEX 3 Thermal Limits Methodology, Summary Description," XN-NF-80-19(P), 3 Volume 3, Revision 1, Exxon Nuclear Company, Richland, Washington (April 1981).

B-3. "The XN-3 Correlation," XN-NF-512(A), Revision 1, Exxon Nuclear Company, Richland, Washington (October 1982).

B-4. "R0DEX2 Fuel Rod Thermal Mechanical Response Evaluation Model , "

XN-NF-81-58(Al, Revision 2, Exxon Nuclear Company, Richland, Washington (March 1984), as supplemented.

B-5. " 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.

B-6. "HUXY: A Generalized Multirod Heatup Code with 10 CFR 50 Appendix K Heatup Option," LN-CC-33 ( A) , Revision 1, Exxon Nuclear Company, a Richland, Washington (November 1975).

I B-7. " Exxon Nuclear Methodology for Boiling Water Reactors: Neutronics Methods for Design and Analysis," XN-NF-80-19(A), Volume 1, Exxon E Nuclear Company, Richland, Washington (May 1980), as supplemented. 5 I

I

I I B-9 XN-NF-86-36 Revision 3 TABLE 1 XCOBRA-T ANALYSIS RESULTS I EVENT INITIAL POWER

(%)

INITIAL FLOW

(%)

REVISED (4)

COTRANSA DELTA-CPR XCOBRA-T DELTA-CPR CHFR MARGIN LRNB(1) 104.2 108 0.08 <0.09 0.02 LRNB 104.2 73.8 0.10 <0.10 0.22 LRNB 70 40 0.16 <0.16 0.79 LRNB 40 108 0.29 <0.29 0.03 LRNB 25 73.8 1.05 <1.05 0.44 LRNB 25 40 0.80 <0.80 0.31 FWCF(2) 104.2 108 0.04 <0.04 0.06 FWCF 104.2 100 0.04 <0.04 0.05 FWCF/NB(3) 104.2 108 0.04 0.04 0.00 l FWCF/RT 104.2 100 0.05 <0.05 0.02 I

(1) LRNB - Generator load rejection without condenser bypass.

(2) FWCF - Feedwater flow controller failure to maximum demand.

/NB - transient without condenser bypass

/RT - transient with 100 degree temperature reduction (3) Original analyses did not include this event.

(4) Results of reanalysis at MP&L request.

I I

I I

E E XN-NF-86-36 L Revision 3 Issue Date: 8/9/86 F

L GRAND GULF UNIT 1 CYCLE 2 PLANT TRANSIENT ANALYSIS r

L DISTRIBUTION

{

DJ Braun JC Chandler TP Currie

{ RE Collingham RA Decker

,_ SF Gaines BJ Gitnick (ENSA)

JG Hwang (ENSA)

JG ingham SE Jensen .

{ TH Keheley JE Krajicek TL Krysinski r TR Lindquist L -

JN Morgan LA Nielsen TW Patten DA Prelewicz (ENSA)

DR Swope CJ Volmer JA White HE Williamson MP&L/JD Floyd (40)

Document Control (5)

I I