ML20086K474

Cycle 6 Plant Transient Analysis

ML20086K474
Person / Time
Site:	Grand Gulf
Issue date:	10/31/1991
From:	Garner N, Hibbard M SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML19353B429	List: ML20086K407 ML20086K440 ML20086K461 ML20086K465 ML20086K474 ML20086K499 ML20086K511 ML20086K519
References
EMF-91-168, NUDOCS 9112130132
Download: ML20086K474 (48)

Text

..

o 3

.;3.-

)

SIEMENS

)

EMF-91-168 1

1 1

1

-]

Grand Gulf Unit 1 Cycle 6

]

Plant Transient Analysis

~

]

]

]

]

October 1991 l

N s Siemens Nuclear Power Corporation

((

P l I (1

)

}d

'F PDR

SIEMENS EMF-91 168 issue Date: 1 G/ 31 ^ 1 GRAND GULF UNIT 1 CYCLE 6 PLANT TRANSIENT ANALYSIS Prepared by

/

Llc '$&

M. J. Hibbard BWR Fuel Engineering Fuel Engineering and Ucensing

..$5W W

• / l l

l N. L Gainer i

BWR Fuel Engineering Fuel Engineering and Ucensing October 1991 I

Siemens Nuclear Power Corporation E ; ee g a-c va,vacu ~; ;a: ',

2 ' O ' Mf' PdO C5 Q000 PC COs 13)

PC"'YO tiA 39352f)130

'e < 5C95 375 8'CC

D 5:e T'iC

De I

CtfSTOMER DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THis -

00CWEN.,,1 SEASE RE.AD CAREFULLY S.emens Nuocer Power Corporaton's warrantes and representatons concoming me subsect maner of this docurnent are mose set form m the Agro6,nent between Siemens Nudeer Power Corporoton and me Cusemer pursuant a which mis document ie issued Accor$ngty, except as oeerwise expressly provu3ed in such

. Agreement neseer Siemens Nuocer Power Corporacon not any person actng on its borwf makes any warranty or representagon. expressed or imphed, wtm respect a the accuracy, comsnenness, or useuness of the eformacon contamed in mis docunent or that the use of any mformanon, apotratus. method or process desaceed m this accument wid not ofnnge pnveteey owred egnts, of assumes any liaodines une respect to the use of any mformaten, accaratus. method or crocess osaceed in trus document.

The clormaton contenned herein is for me som use of me Custoteer.

In order e and impeament of nghts of Siemens Nuclear Power Corporst.on m pasents or erwonsons wrect) may be encluded e the mtormaton contened e tras document, me reassent, by its accootenas of mas encument, agrees not to put$ sh or makeputscuse(inmepatentuseof the serm)of sucrimformasonunof soaumorced e wnene by Siemes Nucceer Power Corporoton or unst anor six (6) mones fotosing termmenon or emessaan of the aforeemd Agreemet and any estension tnereof. unises espreessy prended e me Agreement No nghts or konses e or to any poemte are impsed by me furrdehang of this cmcument

+-

,r--

w.

EMF-91 168 Pagei TABLE OF CONTENTS S ectio,,,0 Pace

1.0 INTRODUCTION

1 2.0

SUMMARY

4 30 THERMAL LIMITS ANALYSIS 16-3.1 Introduction.

-16 3.2 System Transients 16 3.2.1 Design Basis 17 3.2.2 Anticipated Transients..

17 3.2.2.1 Loss Of Feedwater Heating 18 3.2.2.2 Load Rejection No Bypass 18 3.2.2.3 Feetwater Controller Failure 19 3.2.2.4 Control Rod Withdrawal Error...

20 3.2.2.5 Power Dependent LHGR Umit.

20 3.3 Flow Excursion Analysis.

21 3.4 Safety Limit 22 3.5 Summary of Results 22

- 3.5.1 Power Dependent Thermal Umits and Values.

23 3.5.2 Flow Dependent Thermal Umits and Values 23 3.5.3 Exposure Dependern Thermal Umits.

23 4.0 MAXIMUM OVERPRES50RIZATION,...

33 33 4.1 Design Basis...................

4.2 Maximum Pressurization Transients 33 4.3 Results 34 35

5.0 REFERENCES

. APPENDIX A SINGLE-LOOP OPERATION,..

A-1

\\

s_.

EMF-91 168 Page si UST OF TABLES Table Pace 2.1 RESULTS OF ANALYSES 6

2.2 OPERATING UMIT COORDINATES e

3.1 GRAND GULF UNIT 1 CYCLE 6 LFWH DATA

SUMMARY

24 UST OF FIGURES Ficure Paco 1.1 POWER / FLOW MAP USED FOR GRAND GULF UNIT 1 MEOD ANALYSIS 3

2.1 EXPOSURE DEPENDENT MCPR UMITS FOR GRAND GULF UNIT 1 C'iCLE 6 11 2.2 POWER DEPENDENT MCPR LIMITS FOR GRAND GULF UNIT 1 CYCLE 6.

12 2.3 POWER DEPENDENT LHGRFAC VALUES FOR GRAND GULF UNIT 1 CYCLE 6 13 2.4 FLOW DEPENDENT MCPR LIMITS FOR GRAND GULF UNIT 1 CYCLE 6 14 2.5 FLOW DEPENDENT LHGRFAC VALUE FOR GRAND GULF UNIT 1 CYCLE 6 15 3.1 ANALYSIS OF LFWH INITIAL MCPR VERSUS FINAL MCPR,.......,,

25 3.2 LOAD REJECTION WITHOUT BYPASS (POWER AND FLOWS) 26 3.3 LOAD REJECTION WITHOUT BYPASS (VESSEL PRESSURE) 27 3.4 LOAD REJECTION WITHOUT BYPASS (WATER LEVEL ABOVE SEPARATOR SK!RT).

28 3.5 FEEDWATER CONTROLLER FAILURE (POWER AND FLOWS)..

29 3.6 FEEDWATER CONTROLLER FAILURE (DOME PRESSURE) 30 3.7 FEEDWATER CONTROLLER FAILURE (WATER LEVEL ABOVE SEPARATOR SKIRT) _..........

31 3.8 GRAND GULF UNIT 1 CYCLE 6 SAFETY UMIT DESIGN BASIS LOCAL POWER DISTRIBUTION,,

32 A.1 PUMP SElZURE EVENT SLG (POWER AND FLOWS)..

A-3 A.2 PUMP SElZURE EVENT SLO (VESSEL PRESSURE)

- A.4 l

A.3 PUMP SE12URE EVENT SLO (WATER LEVEL ABOVE SEPARATOR SKIRT)A-5 l

i l

l

)

E V F-91 -16S Page 1

1.0 INTRODUCTION

This report presents the results of analyses performed by Siemens Nuclear Pcwer Corporation (SNP) for reload fuelin Grand Gulf Unit 1 Cyde 6 for operation within the Maximum Extended Operating Domain (MEOD). In Cycle 1 (Reference 1) the NSGS vendor performed extensive transient analyses for Grand Gulf Unit 1 in conjunction with the extension of the power / flow operating map to the MEOD. These analyses established conservative operating limits for MEOD operation. The initial reload of SNP fuelin Grand Gulf Unit 1 occurred in Cycle 2.

In support of the initial reload of SNP fuel, extensise additional transient analyses were performed by SNP to justify the NSSS under operating limits and, where necessary, provide appropriate limits for SNP fuel using SNP methodo!ogies (Reference 2).

Cycle 6 for Grand Gulf Unit 1 will include the second reload of SNP 9x9-5 fuel (Reference 15). The nominal cycle energy is 1748 GWd and the cycle length remains 18 months.

The NRC approved methods ernployed for the Cyc!e 6 analysis include the CASMO 3G/MICROBURN B codes (Reference 7), COTRANSA2 system analysis methocs

-(Reference 5), safety limit methodology (Reference 9), and the use of the ANFB Critical Power Correlation (Reference 14) in XCOBRA and XCOBRA-T. The Cycle 6 transient analysis censists of recalculation of the limiting transients at state points having the least margin to operating limits to confirm that the effects of the Cycle 6 changes en transient resu!ts are small relative to available margin and/or establish appropriate limits. Reanalysis of the limiting transients for Cycle 6 assures that the less limiting transients which were previously addressed will continue to_ be protected by the _ established operating limits for Cycle 6.

The poweriflow conditions

- analyzed in Cycle 6 are presented in Figure 1.1. Analyses were performed at EOC 30 EFPD. at EOC, and at EOC+30 EFPD (Effective Full Power Days).

4 h

EMF 91-168 Page 2 These nnalyses establish the Grand Gulf Unit 1 Cycle 6 Technical Specification MCPR limits at rated conditions, establish MAPLHGR limits for Cycle 6 operation, and establish revisea

- thermallimits for off-rated conditions. Previous Grand Gulf reload analyses have demonstrated that the maximum vessel pressure for the most limiting pressurization event varies over a narrow range essentially independent of fuel design. The evaluation of these analyses shows that vessel integrity is protected during the most limiting Cycle 6 pressurization event.

The MCPR, and MCPR, limits have been revised to reflect Cycle 6 results using SNP methodology. The Grand Gulf Unit 1 power and flow dependent MCPR analyses for Cycle 6 were performed at limiting power / flow conditions. LHGR protection has been established for both

- 8x8 and 9x9-5 fuelin Cycle 6 at rated and off4ated conditions. Power and flow dependent LHGR limits have been established for Cycle 6 using SNP methodology.

s.

l

.F

~ :

4

+

12 0,

.0 State Points for: Cycle 6 Transient Analyses-10 0 o

.u e.

t y.o

-so

[

~

C Oo e

60 i/

e.

3:oa 40

/ o.

e L

i O

U a

y 20 -

i 0

a a

t i

e

}

O 10 20 30 40 50 60 70 80.

90 10 0 110 120 Core Flow, Pe' cent of Roted E

r m.

8 $!

• L.

FIGURE 1.1 POWEFVFLOW MAP USED FOR GRAND GULF UNIT 1 MEOD ANALYSIS a.

u*

t f

.t.

.__s 4

w

, i

..m.

EMF 91-168 Page 4 20

SUMMARY

The results of the Grand Gulf Unit 1 Cyt.le 6 transiont analyses support appropnate thermal limits for the Grand Gulf core including the ANF-1.5 9x9-5 reload. SNP thermal limits have been provided for MCPR, that are based on Control Rod Withdrawai Error (CRWE) analyses and analyses for Load Reject No Bypass (LRNB) and Feedwater Controller Failure (FWCF) transients. Additionally, MCPR, limits and LHGRFAC, values (Reference 12) have been established for only the ' loop manual

• mode of operation. The single loop mode of operation (SLO) is evaluated in Appendix A.

The 8x8 MAPLHGR (Reference 16) and a MAPLHGR limit for 9x9-5 tuel satist/ the requirements specified by 10CFR50.46 of the U.S. Code of-Fedoral Regulations. The 8x8 and 9x9-5 LHGR limits will be protected at off-rated conditions by applying LHGRFAC, and LHGRFACp multipliers on the Technical Specification LHGR limits.

Table 2.1 summanzes the transient analyses results applicable to Grand Gulf Unit 1 Cycle 6. These results, together with the Grand Gulf Unit 1 Cycle 6 calculated safety limit MCPR of 1.06. support use of a 1.20 MCPR operating limit (at rated conditions) for Cycle 6 operation between BOC and EOC-30 EFPD. The operating limit (at rated conditions) from EOC-30 EFPD to' EOC+30 EFPD is supported at 1.25. Figure 2.1 presents the MCPR, limit as a function of core average exposure. The calculated safety limit of 1.06 includes the assessment of the channel bow impact using appropriate SNP methods (Reference 9).

The plant transient an.: safety limit analyses results reported herein establish the power dependent Minimum Critical Power Ratio (MCPR ) limits. The power dependent Linear Heat p

Generation Factor (LHGRFAC ) is presented for Cycle 6 operation for SNP 8x8 and 9x9-5 fuel p

types. The MCPR limits, the LHGRFAC values, and the corresponding results of SNP's p

p analyses are presented in Figses 2.2 and 2.3.

The flow dependent Minimum Critical Power Ratio (MCPR,) limit and the results of SNP's analysis are presented in Figure 2.4. The flow dependent Unear Heat Generation Rate Factor (LHGRFAC,) is presented in Figure 2.5. These flow dependent LHGRFAC, values and MCPR, f

4 w

f EMF 91-168 Page5 limits have been established for Cycle 6 to support the "!oop manuat' mode of oporation. These curves are based on conservative maximum core flow rates. Table 2,2 shows tne coordinates

- 9 sed to construct Figures 2.1 through 2.5.

-The implementation of the MCPR operating limit-requires that the r'iost restrictive 1

operating limit be chosen from among the threo MCPR curves based on exposure, flow, and

. power, Thus, the greater value of MCPR as given by MCPR,, MCPR,, or MCPR is selected as p

the operating limit in accordance with the state point of operation (Figures 2.1,2.2, and 2.4).

The results of previous analyses for the maximum system pressurization event are ls

. presented in References 2,22,24, and 25. The results show that the Grand Gulf Unit 1 safety valves have sufficient capacity and performance to protect the ASME Boiler and Pressure Vessel 4

' Code, Section til, Class I, maximum vessel pressur3 transient limit of 1375 psig during Cycle 6.

The fuel' re' lated Technical Specification limits for Cycle 6 operation are included in the reload analysis report (Reference 3).

A w

4 J

6 Y

4 m,,

EMF-91 168 Page 6 TABLE 2.1 RESULTS OF ANALYSES THERMAL UMITS Transient Delta-CPR Loss of Feedwater Heating (all conditions) 0.09 Control Rod Withdrawal Error (Reference 4) 100% power 0.10 70% power (1 foot ganged rod withdrawals) 0.18 70% power (2 foot ganged rod withdrawals) 0.34 20% power 0.48 Feedwater Controller Failure Without Bvoass Delta CPR (Umiting Fuel Type)

% Power /% Core Flow EOC-30 EFPD EOC EOC+30 EFPD 104.2/108' O.13 0,15 0.16 40/10S 0.36 0.37 l.-

104.2% power /10S% core flow is used for the Reload Ucensing Analysis (RLA) conditioris to conservatively bound 100% power /105% core flow, i

t-

EMF 91168 Page 7 TABLE 2.1 RESULTS OF ANALYSES (CONTINUED)

Load Rejection Without 8voass Detta-CPB (Urniting Fuel Typo)

% Power /% Core Flow EOC 30 EFPD

JOQ_,

EOC+30 EFPD 104.2/108 0.14 0.16 0.18 40/108 0.16 0.17 40'108 0.22 0.74 25/73.5**

0.79 0.76 25/40 I 0.60 0.59 s

l I

I l

104.2% power /108% core flow is ured for the Reload Ucensing Analysis (RLA) conditions to conservatively bound 100% power /105% core flow.

Direct scram on turbine trip disabled.

i 1

1

~

EMF-91 168 PageS TABLE 2.2 OPERATING LIMIT COORDINATES 1

GRAND GULF UNIT 1 Cycle 6 MCPR(e) Umits (Figure 2.1)

Core Average Exposure GWd/MTU MCPR(e)

]

13.385 (BOC) 1.20 25.012 (EOC-30 EFPD) 1.20 25.012 1.25 25.831 (EOC) 1.25 26.650 (EOC+30 EFPD) 1.25 MCPR(o) Limits (Figure 2.2)

Percent of Rated Core Power MCPRio) 100 1.20 70-1.24 70 1.41 40 1.49 40 1.85**

40 2.10*

25 2.05 **

25 2.20 Core flow s 50%.

Core flow > 50%.

i

EMF-91-168 Page 9 TABLE 2.2 OPERATING UMfr COORDINATES (CONTINUED)

LHGRFAC(o) Limitg (Figure 2.3)

Percent of Rated Core Power LHGRFAC(o)

M M

100 1.00 1.00 70 1.00 1.00 40 0.69 0.75 25 0.69 0.75 MCPR(1) Umits (Figure 2.4)

Percent of Rated Core Flow MCPR(f) 20 1.28 30 1.28 65 1.20 105 1.20 l

i

l.

l EMF-91-168 Page 10

" ABLE 2.2 OPERATING LIMIT COORDINATES (CONTINUED)

LHGRFAC(0 Limits

[

(Figure 2.5)

Percent of Rated 9x9-5 Core Flow LHGAFAC(0 110.0 1.000 100.0 1.000 90.0 1.000 80.0 1.000 70.0 1.000 68.5 1.000 60.0 O.954 50.0 0.900 40.0 0.846 30.0 0.792 20.0 0.792 Percent of Rated 8x8 Core Flow LHGRFAC(0 110.0 1.000 100.0 1,000 84.3 1.000 80.0 _

0.977 70.0 0.928 60.0 0.880 50.0 0.837 40.0 0.794 30.0 0.752 20.0 0.752

1 7

l.5 l

j i,

O Cycle 6 Anolysis 1.4

1. 3

^e 1.25 i

v E

O-0 U

a y

1.20 g

1.2 g

8 1.1 i

1 1

I BOC EOC-30

'EOC EOC+30 mC Core Average Exposure

?!

c --

1,.

U$

FIGURE 2.1 EXPOSURE DEPENDENT MCPR LIMilS FOR GRAND GULF UNIT 1 CYCLE 6 j

2.50 Cycle 6 Analysis o

O CRWE Results Core Flow > 50 %

2.00 E

o y

Core Flow < 50 7.

o a

1.75 U

2 o

0 1.50 -

i 1.2 s -

~

g o

8 e

a i

e i

e e

1.00 O

to 20 30 40 50 60 70 80 90 10 0 110 12 0 Core Power, Percent of Roted m

^I m"

$91 FIGURE 2.2 POWER DEPENDENT MCPR LIMITS FOR GRAND GULF UNIT 1 CYCLE 6

.. = _...

EMF 91 168 Page 13

1. 6 9 s 91,M G R F A C(i..

O Cysse 6 Treasieht eneire,e j

J i

c 1

1.2

~

I I

o g

a D

D d

j i.o 2

i O:.J D

0 I

3.8 o

i o.e o.*

o io 30 30 40 to 60 70 to to too no

'2e Core Power, Percent of Roted

1. 6 i

8 s 81.e sR F AC(p)

,,.. O Cysie 6 frene.snt Analysie i

s.3 o

a

^

3v 8

b

1. 0 0

=0m o

o.O 0

0 0

o.e o.4 o

to 20 30 40 So so to so to 10 0 tio 12 0 Core Power, Percent of Roted 1

L FIGURE 2.3 POWER DEPENDENT LHGRFAC VALUES FOR GRAND GULF UNIT 1 CYCLE 6 L

I

1 i

l-i i.6 o

cycle 6 Anotysis l

i t.s i

l i.4 O

va a

1.3 D

0 6

s.2 o

t o

i o

O l

t o

r t.:

L r

,,n O

10 20 30 40 50 60 70 80 90 10 0 110 12 0 i

Core Flow, Percent of Roted y

f u?

.5 9 e -

-m FIGURE 2.4 FLOW DEPENDENT MCPR LIMITS FOR GRAND GUL12 UNIT 1 CYCLE 6

g?2 $

vee 3 s

0

~

2 1

6 o

E i

L t

m CY C

0 1

0 T

1 IN U

0 FL 9

U d

3 e

t D

o N

0 R A

8 R

f G

o R

t O

0 n i

7 F

ec E

r U

e L

0P AV 6

w CA g

o F

D s

0F G

l R

5 9+

e H

4 r

L o

T C

N 0

E 4l D

a t

N o

E T

P 0

E 3

D W

O L

0 F

2 5

2 E

0 R

U 1

G I

F O

o.

9 e.

7 6

3, i

1 o

0 o

o o

OvUdO' 0I'-

llIlljrll lll!l!

l

EMF 91163 Page 16 30 THERMAL LIMITS ANALYS!G 3.1 Introduction The scope of the thermallimits analysis includes syst6m transients. localized core events.

and safety limit analysis. Results of these analyses are used to establish po^er, flow. and exposure dependent MCPR limits and LHGRFAC values as appropriate.

COTRANSA2 (Reference 5), XCOBRA T (Reference 6), XCOBRA (Reference 18), and MICROBURN B (Reference 7) are the major codes used in the thermal limits analyses as described in SNP's THERMEX Methodology Report (Reference 8) and Neutronics Methodology Report (Reference 7). COTRANSA2 is a system transient simulation code which includes an axial one-dimensional neutronics model. XCOBRA T is a transient thermal hydraulic code used in the analysis of thermal margins of the limiting fuel assembly. MICROBURN.B is a three dimensional steady state core simulation code which is used for Control Rod Withdrawal Error (CRWE), Loss of Feedwater Heating (LFWH), and flow excursion events (LHGRFAC,). XCOBRA is a steacy state thermal hydraulic code used in the analysis of slow flow excursion events (MCPR,). The ANFB Crillal Power Correlation (Reference 14) evaluates the thermal margins of the fuel assemblies. This correlation has been generically approved by the NRC (Reference 14).

3.2 System Transients Thermal limits have been appropnately revised based upon SNP methods used in the Cycle 6 analysis. Figure 1.1 shows the four power / flow conditions that were analyzed in support of the Cycle 6 reload System response for pressurization transients from these state points was ancyzed for Cycle 6 using COTRANSA2. The Load Reject No Bypass (LRNB) pressurization transient analysis was performed at each of the four state points. The Feedwater Controller Failure (FWCF) analysis, without credit for bypass valve operation, was performed at 104 2%!108% and 40%/108% ASME pressuitzation analyses were performed for Cycles 2,3, 4, and 5 and wem not repeated for Cycle 6. LFWH analyses were performed with MICROBURN B for a large number of exposure points for Cycles 1 through 5 (Reference 22) as well as for Cycle 6. Analyses have been performed considering the SNP 9x9-5 fuel to assure that the power dependent limits supported by analyses for control rod withdrawal error remain applicable to l

[

i L

EMF 9t 168

[

Page 17 Grand Gulf Unit i Cycle 6. These analyses show less restrictive resu!!s of little change from the Cycle 5 analyses due to Cycle 6 changes, thus justifying that the less limiting transients not analyzed for C e 6 will continue to be protected. The pump seizure event was analyzed for single-loop operation for Cycle 6, the results are presented in Appendix A.

i 3.2.1 Desion Basis The LRNB and FWCF transients have been determined to be most limiting at end of full i

power capability when control rods are fully withdrawn from the core. Between BOC and EOC 30 EFPD, the CRWE translept is most limiting. From nominal EOC 30 EFPD to EOC+30 EFPD, the LANB and FWCF transients are limiting. The delta CPR calculated for EOC 30 EFPD, EOC, and EOC+30 EFPD is conservative for cases where control rods are partially inserted. The analysis for Grand Gutt Unit I with MEOD was performed using conservative analytical limits for trips and setpoints. Events initiated at core powers below 40% rated were analyzed with the direct scram due to turbine control and stop valve fast closure disallowed, and with the recirculation pump high to low speed transfer disabled. Recirculation pump inp on high dome pressure was enabled for events initiated at core powers below 40% rated.

t t

' 42.2 : Anticiosted Transients

. s SNP's transient methodology report for jet pump BWRs (Reference 19) considered eight categories of anticipated transients. The most limiting transients were evaluated at various power / flow points within MEOD to verify the power dependent thermal margin for Grand Gulf Unit 1 Cycle 6. The limiting transients analyzed for Grand Gulf Unit 1 Cycle 6 were:

Loss of Feedwater Heating Load Rejection No Bypass Feedwater Controller Failure No Bypass

+

Other transients are inherently nonlimiting or bounded by one of the above as shown in the NSSS vendor MEOD' analyses for Cycle 1 and the SNP Grand Gulf Unit 1 Cycle 2 analyses.

Control rod withdrawal error is an exception in that it has been analyzed generically.

EMF 9116B Page 18 32.2.1 Loss Of Feedwater Heatmo Analysis of the les of feedwater heating event was performed to reflect reactor operation over the MEOD operating power versus flew map and conditions anticipated dunng actual Grand Gulf reactor operation.

Calculations performed for Cycles 1 thf 369h S aaN t ;onservative reducuen of 100'F in the feedwater temperature. Results for C,jejas ) threc Qh 5 are provided in Table 3.1 of Reference 22. Table 3.1 provides the conditions for up, cases analyzed in Cycle 6 ir :erms of cycle exposure, core power, and core flew. Tne ir"ht and final MCPR values are presented for each case.

Analysis of the data from previous cycles revealed a strong correlat:on between the initial and final MCPR. A least squares fit of these data resulted in a linear relationship such that:

MCPR(initial) = -0.04974 + 1.1021

• MCPR(final)

In order to conservatively bound all of the calculated data, the largest deviation between the calculated and fitted results were applied to the least squares fit such that the LFWH MCPR operating limit is defined by:

OLMCPR(LFWH) = -0.02386 + 1.1021

• Sl>ACPR This bounding relationship and the Cycle 6 data are presented in Figure 3.1. Substituting the SLMCPR of 1.06, the MCPR operating limit for the LFWH event for all operating conditions analyzed is 1.15.

3 2.2.2 Load Rejection No Oveass The Load Rejection No Bypass (LRNB) event is the mostlimiting of the class of transients characterized by rapid vessel pressurization for Orand Gulf Unit 1. The load rejection causes a fast closure of the turbine control valves.. The resulting compression wave travels through the steam lines into the vessel and creates the rapid pressunzation condition. A reactor scram and

-p T-m 4

y ymy rg---,

--w-

-mrw--

,e wm-r-

ey

- - + + - -

c +

3-e r

EMF 91169 Page 19 a recirculation pump transfer from high to low speed are irutiated by fast clo valves. Condenser bypass flow, which can mitigate the pressurization effe excursion of the core power due to void collapse is primarily terminated by void growth due to the recirculation pump high to low speed transfar.

Figures 3.2, ~3.3, and 3.4 present the response of various reactor an to the MNB event hilated at the Reload Ucensing Analysis condition (104.

flow). The MCpR operating limit of 120 is bounding for all exposures up t MCPR operating limit of 1.25 is bounding for all exposures between EO conditions. Table 2.11.ists the delta-CPRs for this transient at the power / flo exposure conditions considered.

3.2.2.3 Feedwater Controller Failu.r,,g, e

The failure of the feedwater controller to maximum demand (FW of the vesselinventoryincrease transients. Failure of the feedwater contro demand would result in an increase in the coolant level in t feedwster flow results in lower temperatures at the core inlet, which in tu core power level.

n If the feedwater flow stabilizes at the increased value, the core pow stabilize at a new, higher value. If the flowincrease continues, the water le will eventuaaf reach the high level setpoint, at which time the turbine stop avoid damage to the turbine from excessive liquid inventory in the stea level trip also initiates reactor seram, and subsequent turbine trip iea to low speed transfer. The core power excursion is terminated by the sa g

end the LRNB transient.

a Figures 3.5,3.6, and 3.7 present the response of various reactor and to the FWCF without bypass 'eyent initiated at the Reload Ucensed Ana power /108% core flow). The delta CPR for this event was calculated to be 0.1 delta-CPR is bounded by the LRNB delta-CPR. At EOC 30 and EOC+30 for this event is also bounded by the LRNS. The cases of FWCF with by r

i l

EMF 9116B Page 20 heaters out of se vice ( 100 'F) were analyzed in previous cycles and shown to be bounded by FWCF without bypass case.

1 3 2.2.4 Contrel Aod W1hAayval Error Reference 4 documents G!C 's generic CRWE analysis for Grand Gulf Unit 1 cperation within the MEOD. The applicability of these analyses was confirmed by performing CRWE analyses with MICROBURN B using SNP's ANFB entical power correlation.

Based on Reference 4 operating conditions and analytical procedures. one and two foot CRWE events were simulated. Designs using 9x9-5 fuel were also analyzed (Reference 22). The results of these analyses were statistically combined to produce a 95/95 upper limit for various power levels. This upper limit is bounded by the generic analysis results. Figure 2.2 shows tne operating limit curve for protecting the Cycle 6 fuel under CRWE conditions based on SNP's generic CRWE analysis and the Cycle 6 MCPR safety limit of 1.06.

3.2.2.5 Power Decendent LHGR Umit Transient analyses have been performed to define appropriate multipliers on the fuel design limit LHGR for part power operation. The purpose of these multipliers is to protect fuel from failure due to centerline melt and exceeding tr's 1% plastic strain mechanical performance d'esign criteria during off-rated condition transient events Analyses were performed for the Lead Rejection No Bypass (LRNB) and Feedwater Controller Failure (FWCF) pressurization event transients and the Control Rod Withdrawal Error event which is a localized event. Analyses performed for Cycle 2 showed the LANB and FWCF transients to be limiting relative to MCPR and LHGR increases. CRWE analyses performed at various off-rated conditions on the power / flow map gave results which were less restrictive than for the LRNB and FWCF events. The LRNB and FWCF transients were evaluated for Cycle 6 considering a variety of exposure and operating conditions, The results of these analyses are provided in Figure 2.3 and demonstrate adequate margin to the operating limit: Separate limits are established for SNP 8x8 and 9x9-5 fuel types bas ~ed upon the appropriate transient LHGR limit for each fuel type.

l

EMF 91-168 Page 21 i

33 Aow Excursion Analysis The flow excursion transient is analyzeo to determine the flow dependent thermallimits and values (MCPR, and LHGRFAC,). This transient is analyzed by assuming a failure of the recirculation flow control system such that the recirculation flow increases slowly to the physical maximum attainable by the equipment. The mode of operation analyzed for Grand Gulf Unit 1 Cycle 6 is " loop manual" oni. This mode of operation corresponds to a single recirculation loop f

flow excursion event.

The results of the flow excursion transient analyses were used to establish new flow dependent thermallimits of MCPR,. For these analyses the change in entical power along the flow ascension path was calculated with XCOBRA (Reference 18) Peaking factors were selected such that the bundle with the least margin would reach the safety limit MCPR of 1.06 at the maximum achievable flow. Figure 2.4 presents the MCPR, limit for maximum achiev'able core flow, conservatively assuming that the recirculation system equipment is capable of 110% of rated flow on the limiting rod line. For flow rates less than 30% rated flow, the recirculation system operates at low speed restricting the maximum possible flow. Because of this restriction.

the MCPR, curve conservatively remains fixed between 20% flow and 30% flow.

The Cycle 6 LHGRFAC, analysis was performed with the CASMO-3G/MICROBURN.B neutronic codes assuming a single pump runup flow excursion. The analysis assumes that the recirculation flow increases slowty along the limiting rod line (Reference 2) with a maximum core flow capacity of 110% of rated. A series of flow excursion ana'yses were performed starting from different initial power / flow conditions. Variations in the cycle exposure and control rod patternt were also considered. The final conditions are conservatively determined based on the maximum attainable core flow rate. Xenon is conservatively assumea to remain constant during the event.

The operating limits were established to bound the limiting results and are shown in Figurs 2.5.

Separate limits are established for 0NP 8x8 and 9x9 5 fuel types based upon the appropriate j

transient LHGR limit for each fuel type. Because of restrictions in flow rates attainable for operation with core flows less than 30% of rated, the LH'.R/AC, conservatively remains constant for core flow rates between 20% and 30%.

-~.

EMF 91 168 Page 22 3.4 Safety Limit i

The safety limit MCPR is defined as the minimum value of the critical power ratio at which j

the *uel could be operated, with the expected number of rods in boiling transition not exceeding 0.1% of the fust rods in the core. The safety limit is the minimum critical power ratio which would be permitted to occur during the limiting anticipated operational occurrence. The safety 1*,it MCPR for all fuel types in Grand Gulf Unit 1 Cycle 6 operation was calculated to be 1.06 using the methodology presented in References 9 and 11. The determination of the safety !antit i

expkcitly includes the effects of channel bow and relies on the following assumptions:

l 1.

Cycle 6 wiH not contain channels used for more than one fuel bundle lifetime.

2.

The channel exposure at dischttge will not exceed 40.000 mwd /MTU based on the fuel bunole average exposure.

3.

The Cycle 6 core will contain GE and Cartech supplied channels.

4.

The limiting moc'ule contains a conservative exposure configuration.

The input parameter values for uncertainties used in the safety limit MCPR analysis are unchanged from the Cycle 2 analysis presented in Reference 2 except for the uncertainties associated with the ANFB correlation, its implen entation in the safety limit evaluation. cha 'el bow, and the uncertainties appropriate for CASMO/MICROBURN analysis. The limiting local power distribution used to determine the safety limit MCPR is shown in Figure 3.8. The effects of channel bow were modeled in the safety limit evaluation.

3.5 Summary of Aesults The results of the Grand Gulf Unit 1 Cycle 6 thermallimits analysis show a Cycle 6 safety limit MCPR of 1.06 and a MCPR operating limit of 1.20 at rated conditions for exposures below EOC-30 EFPD. A MCPR operating limit of t.25 at rated conditions is shown from EOC 30 to EOC 4 30 EFPD. These exposure dependent limits are shown in Figure 2.1. The MCPR operating limit considers the effects of exposure (MCPR), flow (MCPR,), and power (MCPR ).

The p

operating limit ofinterest is the larger of the three values for a given reactor operating condition.

EMF 91 168 Page 23 3 5.1 Power Dependent Thermal Umits and Values The power dependent MCPR limit (MCPR ) protects against exceeding the safety limit p

MCPR during anticipated oper'ational occurrences from off rated conditions. The MCPR, limit bounds the sum of the delta CPR for the limiting event and the calculated safety limit MCPR.

The power dependent LHGRFAC (LHGRFAC )is used to protect against both fuel melting p

and 1% clad strain during anticipated system transients from off rated conditions. The conservative LHGR values for protection' against fuel failure during anticipated operational occurrences are given in References 10 and 13. The results are presented in a fractional form for application to the LHGR operating limit. The flow dependence of the LHGRFAC, at low power has been conservatively removed.

The MCPR, limits and LHGRFAC, values for Cycle 6 are shown in Figures 2.2 and 2.3.

respectively. Results from the Cycle 6 transient analyses and the SNP generic CRWE analyses establish the MCPR, operating limit for Cycle 6. The Cycle 6 LHGRFAC values establish the p

applicable operating !imits for SNP 8x8 and 9x9-5 fuel,

'3'5.2 Flow Dependent Thermal Umits and Values The flow dependent MCPR iimit (MCPR,) protects against exceeding the safety limit MCPR for slow flow excursion events. The results of the MCPR, analysis for Grand Gulf Unit 1 Cycle 6 are presented in Figure 2.4. The flow dependent LHGRFAC (LHGRFAC,) protects against both fuel melting and 1% clad strain. The LHGRFAC, values for SNP 8x8 and 9x9 5 fuel to be used in Cycle 6 are presented in Figure 2.5.

3.5.3 Exoosure Deoendent Thermal Umits The exposure dependent MCPR limit (MCPR,) protects against exceeding the safety limit MCPR during the operation of the core. The results of the exposure dependent analysis for Grand Gulf Unit 1 Cycle 6 are presented in Figure 2.1.

i I

l

EMF.91 168 Page 24 TABLE 3.1 GRAND GULF UNIT 1 CYCLE 6 LFWH DATA

SUMMARY

Initial State Final State Cycle Total Core Total Core Core Total Core Total Core Core Exposure Power Flow Minimum Power Flow Minimum fGWd/MT) fMWt)

(M1 b/hr_L CPR fMWt)

__ f M1b/hr)

CPR 0.00 3833 106.88 1.472 4333 106.88 1.385 1.00 3833 102.38 1.460 4332 102.38 1.376 2.00 3833 103.50 1.476 4335 103.50 1.382 3.00 3833 96.75 1.445 4336 96.75 1.352 4.00 3833 101.25 1.427 4332 101.25 1.346 5.00 3833 104.63 1.430 4333 104.63 1.342 6.00 3833 99.00 1.397 4324 99.00 1.301 7.00 3833 100.13 1.386 4328 100.13 1.298 8.00 3833 99.00 1.316 4314 99.00 1.235 9.00 3833 100.13 1.328 4296 100.13 1.251 10.00 3833 105.75 1.336 4283 105.75 1.265 11.00 3833 105.75 1.369 4272 105.75 1.296 12.00 3833 103.50 1.382 4260 103.50 1,314 12.45 3833 112.50 1.409 4266 112,50 1.338 13.27 3833 112.50 1.423 4261 112.50 1.353 l

l

-u-

3.5 Cycle 6 LFWH Analysis 4

3.0 9

h f+

o s

2.5 (o

o g

2 o-

/

~

o

[s 3

?

2.0 q$

o*

t.s I

.~

1.0 1.0 1.5 2.0 2.5 3.o 3.5 m

Final MCPR I

m '"

E*

O!E FIGURE 3.1 ANALYSIS OF LFWH INITIAL MCPR VERSUS FINAL MCPR 7

a 0-a

4 1

4 350 1

300-

+=

Relative Core Power i

x=

Relative Heat Flux Relative Recirc Flow o=

250-V=

Relative Steam Flow a= Relative Feed Flow 1

200-j 15 0 -

10 01 l

50-d t

0-

. l 10 0 m

0 0.5 1

15 5

2.5 3

3.5 4

4.5 5

m '"

Time. seconds aT

• a FIGURE 3.2 LOAD REJECTION WITHOUT BYPASS (POWER AND FLOWS) j

f i

1500 1400-1 P-I m

CL T-1300-om (fl b

a.

~em 1200-mo>

110 0-I m

7 1000 0

0.5 1

1.5 2

25 5

'3S A

45 5

=o Time, seconds WI

"'I E FIGURE 3.3 LOAD REJECTION WiiHOUT BYPASS (VESSEL PRESSURE) e

s s

O e

4.50

/

. k Y

k 7e f

)

1. 4+

• po-

'e f

7

$4

\\

\\

(

y

~

~

~

s' l'

sk b

o oN 9

ee 'ec##'

e, s q#3 j s $ soa #$

1. # p f', # # #

e/**

4 i

250 Relative Core Power Relative Heat Flux 200-

........e..&..G..n.... ile..s....i..~.6..&.e.... ~.

~ ~ ii ~l Relotive_ Steam Flow_

~

Relative Feed Flow 33 o _

.)

,~.

e g

10 0-

_ _ _ m,. _

_ - - - - -.- y m m :. - E., -, ',

g

.~,

I O

y

, ~

c 5

h 50-jj e

i. I t

\\

"q n.

I I N-0-

1 i,

J 3

e y

m 4

-10 0 E

i' O

2.5 5

7.5 10 12.5 15 17.5 20 g,P Time, seconds SE om e <=

1 FIGURE 3.5 FEEDWATER CONTROLLER FAILURE (POWER AND FLOWS)

(

4 e

I j

i

._~

,, _ - ~. _ _

i.

4 1t-1500 t

i i

1 i

1400-L

.O l.

O 1300-

.e t.

3 r

m mc 1-d q.

i

, 1200-m ee>

1 t

4 1100-I l

mE 1000 s

I O

2.5 5

7.5 10 12.5 15 17 5 20 57' me

. Time, seconds WI i

a m.

o FIGURE 3.6 FEEDWATER CONTROLLER FAILURE (DOME PRESSURE) t 3

s 1

L b

e s

r rw..

w-

m%9:.g

,a= 9 02

)TR ik S

[

R O

TAR APE S

I EVOeA LEVE L

R g

E d

T n

A o

W C

(

=o e E

t S

Ru O

t M

A l

F R

5 L

s7 L

O R

TN OC R

s5 ETA WD EE 5

(

=2 7,

3 ERU O

5 6

5 5

5 4

5 3

6 5

4 3

C 'I>$ t2ON 3mM$

i

,4 i!

u

EMF 91168 Page 32 L

CONTROL R00 0

N 0.986 1.025 1.018 1.030 1.063 1.030 1.018 1.025 0.986 T

R 1.025 0.967 1.,047 0.989 0.014 0.989 1.047 0.966 1.025 0

L 1.018 1.047 1.028 0.970 0.994 0.968 1.027 1.047 1.019 R

1.030 0.989 0.970 0.897 0.000 1.150 0.970 0.990 1.031 0

O 1.063 0.814 0.994 0.000 0.000 0.'000 0.999 0.814 1.064 1.030 0.989 0.968 1.050 0.000 0.889 0.982 0.993 1.032 1.018 1.047 1.027 0.970 0.999 0.982 1.035 1.051 1.020 1.025 0.966 1.047 0.990 0.814 0.993 1.051 0.967 1.027 0.986 1.025 1.019 1.031 1.064

-1.032 1.020 1.027 0.987 FIGURE 3.8 GRAND GULF UNIT 1 CYCLE 6 SAFETY LIMIT DESIGN BASIS LOCAL POWER DISTRIBUTION

a a

aa.4A-a+4+4m b444-A4&4-.---+.m 4-

1. A-++IM 1-A A 4-8-#- - - *

• h=m3--1idi**--W-' ' '

M'py

4. 4 J4

-64 2 4 4 3 # M

---

e-h-Cd.D zi r_ b4Ae sie -6M

s. MwW-l h. aim 5+&W-M e's

&i a+Mh4+44W pt6-&

aAmE.m44.-.3, i

i I

f i

P i

t r

o P

6 4

l I

i t

h l

i f

s s

l T

I

-fs f

1 k

t I

b I

t T

t h

h t

I 6

h A

a n

6 4

I l

i l

I I

1 i

- l1 a

4 I

-W

% -rW webfAt pwaW seur g W -r-e g 4 g

• rais-w

+gg.--r,M iyWg ge H w my-)n m c Wyyr ger g7y-F

• rw-yy-g ywW i emia-t m w eg y

1e Tes-yv--

--v g--

m me y, e p 3 g gp-vaig.w-A-

maSevFr-*-M8 Cat ae9 g whty v-g T'P"WIW *Irv w-

d EMF 91 168 Page 33 40 MAXIMUM OVERPRESSURIZATION Max, mum system pressure has been calculated for the containment isolation event (rapid closure of ah main steam isolation valves) with an adverse scenario as specif;ed in the ASME Pressure Vestel Code for Cycles 2 through 5 (References 2,22,24, and 25). These analyses demonstrate that the Grand Gulf Unit 1 safety valves have sufficient capacity and performance to prevent pressuto from reaching the established transient pressure limit of 110% of design pressure (1,1 x 1l?50 = 1375 psig).

4.1 Desian Basis During the transient, the most critical active component (direct scram on MSIV closure) was assumed to fail. The event was terminated by the high flux scram. Credit was taken for actuation of only *,? of the 20 safety / relief valves: 6 in the relief mode and 7 in the safety mode.

The safety valve analysis setpoints for these calculations included a conservative 6% tolerance.

4.2 Maximum Pressurization Transientg t

Scoping analyses desenbed in Reference 19 found the closure of all main steam isolation valves (MSIVs) without direct scram to be limiting The MSIV closure was found to be limiting when alltransients are evaluated on the same basis (without direct scram) because of the smaller st'eam line volume associated with MSIV closure. Though the closure rate of the MSIVs is substantially slower than turbine stop or control valves, the compressibility of the additional fluid in the t'eam lines associated with a turbine isolation causes these faster closures to be less severe. Once the containment is isolated, the subsequent core power production must be absobed in a smaller volume compared to that of a turbine isolation resulting in higher vessel pres Nres.

9

~ - -., -

y

E i

EMF 91 168 Page 34 4.3 Re5Vits The maximum vessel pressures at the most limiting power / flow point for il

• previous cycles analyses demonstrate that the maximum vessel pressure varies over a very narrow range (1271 psig to 1298 psig) Independent of fuel and core design and that sufficient margin, more than 75 psid,is available to accommodate the minor changes represented by the Cycle 6 reload.

t k

,..++-4+4

,n.

4 W*

d

.4

EMF 91169 Page 35 50 REFERENCES 1.

Lester L Kintner, USNRC, Letter to O. D. Kingsley, Jr., MP&L, Technical Spec:fication Changes to Allow Operation with One Recirculation Loop and Extended Operat;ng Domaan" August 15,1986.

2.

" Grand Gulf Unit 1 Cycle 2 Plant Transient Analysis," XN.NF 86 36. Revision 3. Euen Nuclear Company, Inc., Richland, WA, August 1986.

3.

" Grand Gulf Unit 1 Cycle 6 Peload Analysis," EMF 91169. Siemens Nuclear PcAer Corporation Richland, WA, October 1991.

4.

"BWR/6 Generic Rod Withdrawal Error Analysis: MCPR, for Plant Operations Within the Extended Operation Domain," XN NF 825(P)(A), Supplement 2, Euon Nuclear Company.

Inc., Ricnland, WA, October 1986.

5.

COTRANSA2; A Computer Program for Boiling Water Reactor Transient Analysis" ANF 913(P)f A), Volume 1 Revision 1 and Supplements 2. 3, and 4, August 1990.

6.

"XCOBRA T: A Computer Code for BWR Transient Thermal Hydraulic Core Analysis."

XN NF-84-10%LP1(Al, Volume 1,

Euon Nuclear Company. Inc., Richland. WA.

February 1987.

7.

"Euon Nuclear Methodology for Boiling Water Reactors: Neutronics Methods for Design and Analysis," XN NF 8019(A). Volume 1 and Supplement 3, Euon Nuclear Company, Inc., Richland, WA, March 1983.

S'.

"Enon Nuclear Methodology for Boiling Water Reactors THERMEX: Thormal Limits Methodology Summary Description," XN-NF 80-19(P)(A), Volume 3. Revision 2. Exxon Nuclear Company, Inc., Richland, WA, January 1987.

9.

" Advanced Nuclear Fuels Cntical Power Methodology for Boiling Water Reactor,"

ANF 524(P)(A), Revision 2, and Supplements, Advanced Nuclear Fuels Corporation.

Richland, WA, April 1989.

10.

" Grand Gulf Urit 1 Reload XN 1.3 Cycle 4 Mechanical Design Report," ANF 88183(P),

Supplement ', Advanced Nuclear Fuels Corporation, Richland. WA, August 1991.

11.

"Euon Naclear Methodology for Boiling Water Reactors: Application of the ENC Methooology to BWRMalcads," XN NF-80-19(P)(A), Volume 4, Revision 1, Euon Nuclear Company, Inc., Richland, WA, June 1986.

12.

" Grand Gulf Nuclear Station Unit 1 Revised Flow Dependent Thormal Limits."

NESDO-88-003, MSU System Services Inc., November 1988.

I

i EMF 91 168 Page 36 13.

" Generic Mechanical Design for Advanced Nuclear Fuels 9x9-5 BWR Reload Fuel?

ANF 88152fP)fA) with Amendment 1 and Supplement 1, Advanced Nuclear Fuels Corporation, Richland, WA, November 1990.

i 14.

"ANF8 Critical Power Correlation," ANF 1125(P)(A) and Supplements 1 and 2, Apnl 199 15.

" Grand Gulf 1 ANF 1.5 Design Report, Mechanical, Thermal Hydraulic, and Neutronic Design for Advanced Nuclear Fuels 9x9-5 Fuel Assemblies," ANF 91-080(P), July 1991.

16.

" Grand Gulf Unit 1 LOCA Analysis," XN-NF.86-38, Jurie 1986.

17.

Not used, l

18.

"XCOBRA Code Users Manual," XN NF-CC-43, Revision 1, January 1980.

19.

" Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors," XN-NF 79-71(P Revision 2, including Supplements 1, 2, & 3(A), Exxon Nuclear Company, Inc.,

Richland, WA, Novemoer 1981, 20.

Not used.

21.

Not used.

22.

" Grand Gulf Unit 1 Cycle 5 Plant Transient Analysis Report." ANF 90-021, Revision 2.

Advanced Nuclear Fuels Corporation, Richland, WA, August 1990.

23.

" Grand Gulf Unit 1 Cycle 6 Single Loop Operation LOCA Analysis Report," EMF-91172, Siemens Nuclear Power Corporation, Richland, WA, A.ugust 1991.

24.

" Grand Gutt Unit 1 Cycle 3 Plant Transient Analysis Report," ANF 87-66, Revision 1, Advanced Nuclear Fuels Corporation, Richland, WA, August 1987.

2S. -

" Grand Gutt Unit 1 Cycle 4 Plant Transient Analysis Report? ANF 88-150, Advanced Nuclear Fuels Corporation, Richland, WA, November 1988.

9

EMF.9t 168 Page A 1 APPENDIX A SINGLE LOOP OPERATION Analyses have been provided by the NSSS vendor tnat demonstrate the safety of plant operation with a single recirculation loop out of service for an extended period of time. These analyses confirm that during single-loop operation. the plant cannot reach the normal cuncle power levels and nodal power levels that are possible when teeth recirculation systems are in operation. The physical-interdependence between core power and recirculation flow rate inherently limits the core to less than rated power, Because the SNP 9x9 5 fuel was designed to be compatible with the co-resident 8x8 fuel in thermal hydraulic, nuclear, and mechanical design performance, and because the SNP methodology has given results which are consistent with those of the previous analyses for two-loop operation, the analyses performed by the NSSS supplier for single loop operation are also applicable to single loop operation with fuel and analyses provided by SNP.

A.1 PUMP SElZURE ACCIDENT The pump seizure is a postulated accident where the operating recirculation pump suddenly stops rotating. This causes a rapid decrease in core flow, a decrease in the rate at "w'hich heat can be transferred from the fuel rods and a decrease in the critical power ratio.

COTRANSA2 and XCOBRA T are used to calculate the MCPR for SNP fuel during a pump seizure from single-loop operation.

COTRANSA2 was used to simulate system response to a pump seizure in single loop operation at the power flow point of 70.6% rated power and 54.1% rated flow. The operating recirculation pump rotor was stopped quickly causing a sudden decrease in the active jet pump drive flow.

During the event, the inactive jet pump diffuser flow went from negative flow to positive fiow. Figures A.1, A2, and A.3 show the graphical representation of important system parameters during the accident.

_. _ _ _ _ _. _ __ _. ~. _. _ _ _. _ _ _ _ _

EMF 91168 Page A 2 Thcrmal hydraulic analysis using SNP safety limit methodology has shown that the two-loop MCPR limit provides the required protection below 70% of rated core power such that any p

postulated fuel failures would not result in exceeding a small fraction (<10%) of the 10CFR100 requirements.

A.2 MCPR SAFETY UMIT For single 4oop operation, SNP has determined that a safety limit of 1.07 provides sufficient protection to account for increased TIP uncertainties and increased flow measurement uncertainties associated with single-loop Operation, S.) has evaluated the effocts of these uncertainties using SNP safety limit methodology and determined that augmenting the %c loop safety limit MCPR by 0.01 is appropriate for SNP fuel during single-loop operation for Cycle G.

A.3 FLOW DEPENDENT AND POWER DEPENDENT THERMAL LIMITS 11 is apptcpriate to uso the reduced flow and power two loop operating MCPR rond

~

LHGRFAC limits for single loop operations. The reduced flow MCPR limit is to protect against boiling transition during f%w excursions to maximum flow. The reduced flow LHGRFAC is based on the heat flux increase associated with an excursion to maximum flow.- The flow dependent limits are bounding for single-loop conditions because of the limited core flow capacity in single-loop operations. The power dependent MCPR limit (MCPR ) protects against exceeding the p

safety limit MCPR during anticipated operational occurrences from off rated conditions. The power dependent LHGRFAC is used to protect against both fuel melting and 1% clad strain

' during anticipated system transients from off rated conditions. The power dependent limits established for two-loop operation are appropriate limits for single-loop operation because the limiting events are unaffected by the single-loop mode of operation.

4 A.4 MAPLHOR LIMITS SNP has established that the two loop MAPLHGR limits for SNP 8x8 and 9x9-5 fuels multiplied by a reduction factor of 0.86 may be conservatively applied for single-loop operation.

Application of this reduction factor ensures that tne peak clad temperature from a single loop operation LOCA is bounded by the two-loop LOCA analysis. The application of these limits is valid for average planar burnups of up to 50000 mwd /MTU and 55000 mwd /MTU for SNP ex8 and 9x9-5 fuels, respectively (Reference 23).

4 80)

Relative Core Power 70J

+=

x= Relative Heat Flux

'o=

Relative Recirc Flow v = Relative Steam Flow 60-a = Relative Feed Flow V

_v j

0 50-O' t

N

-d t

u 40-s t

i 30-3 20-10 i

O 1

2 3

4 5

6 7

8 9

10 11 m

E i

y j

Time, seconds g[

FICURE A.1

• I PUMP SElZURE EVENT FLO (POWER dND FLOWS)

?e ua f

s

e -

j l

1015 l

i 6

1010-l 1005J l

.9

$ 1000-ea my 995-990-985-980-m gf u

1 975 0

1 2

3 4

5 6

7 8

9 10 11 Time, seconds

E (b

FIGURE A.2 PUMP SEl20RE EVENT SLO (VESSEL PRESSURE)

9 6

5.75-5.50-C c

5.25-

__t L,

sc

[

5-

-e M

M e>

4.75-4.50-4.25 m

g g{

0 1

2 3

4 5

6 7

8 9

10 i:

y Time, seconds

• E

%' e FIGURE A.3 PUMP SElZURE EVENT SLO (WATER LEVEL ABOVE SEPARATOR SKIRT)

EMF 91 168 issue Date:

10/31.'31 GRAND GULF UNI - ('YCLE 6 PLANT TRANSIENT ANA Q,istribution O. C. Brown R. A. Copeland L J. Federico D L Garbor N. L Garner D. E. Hershberger M. J. Hibbard R. B. Macduff J. N. Morgan R. S. Reynolds C. C. Roberts C. J. Volmer G. N. Ward A. W. Will Entergy Operations /S. L Leonard (40)

Document Control (S) 4

. to GNRO-91/00186

____ - -_ _ _-