ML20115J614

Rev 2 to Grand Gulf Unit 1 Cycle 6 Plant Transient Analysis

ML20115J614
Person / Time
Site:	Grand Gulf
Issue date:	09/30/1992
From:	Garner N, Hibbard M SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML20115J613	List: ML20115J610 ML20115J614 ML20115J619
References
EMF-91-168, EMF-91-168-R02, EMF-91-168-R2, NUDOCS 9210280270
Download: ML20115J614 (44)

Text

.

SIEMENS EMF 91 168 Revision 2 lasue Date:

9/23/92 GRAND GULF UNIT 1 CYCLE 6 PLANT TRANSIENT ANALYSIS Prepared by Ij',O

, /( F ld.A$tr<2-

/

N'L. Garner BWR Nuclear Engirieering

/

C

$fi c N

' M. J. Hibbard BWR Nuclear' Engineering September 1992

/ pal 9210280270 921026 PDR ADDCK 05000416 P

PDR

c e

e CVSTOMER D!t OLAIMER IMPORTANT NOTICE REQAMDING CONTINTS AND USE OF THit Dr)CUMENT PLEAlt. MEAD CAMEFULLY Siernens Power Corporation's wwrentles and representauona concem6ng the subpect matter of this document are those set forth in the Agreement between Siemens Power Corporavon and the Customer pursuant to wNeh this socument le laeued, Accordingly, ehcopt as otherwise expreesty prov6ded in such Agrooment, neeer Semene Power Corporsuon not any person acung on its behalf makes any warranty or representagon, expressed of Implied, wfth respect to the accuracy, completeness or ueefulnese of the informahon contained 6n this dormment, of that the use of any informanon, inoparatus, method or p#ocess disclosed in thle document will not infringe privately owned rights; or assumes any liabilities wtth respect to the use of any information, apparatus, methoc or procais disclosed in tNe document.

The informabon contained here6n is for the solo use of the Custumor.

in order to avoid impairment W r6ghts of Siemens Power Corporation in patente or inventions wNch may be included in the information contasned 6n thle document, the recipient, by its acceptance of this document, agrees not to publish or make public use (in the patent use of the term) of such information until to authorued in writing by Siemene Power Corporacon or unul after six (6) months follow 6ng termination or expiration of the aforesaid Agreement and any extension i

thereof, unless expressy provided in the Agrcament. No r6gt,ts or licensee in or j

to any patente are impised by the fumisNng of this document.

-i J

t EMF 91 168 Revision 2 Pagei TABLE OF CONTENTS Stetion Peut 1.0 I NTR O D U CTI O N.............................................

1 3.0 S U M M A r.Y...............................................

4 3.0 THERMA'. LIMITS ANALYSIS 16 3.1 I n t r oef u c t i on..........................................

16 3.2 S ys t e m Tra nsie n t s.....................................

16 3.2.1 D e sig n Ba sis....................................

17 3.2.2 Anticipated Transients 17 3.2.2.1 Loss Of Feedwater Heating...................

18 3.2.2.2 Load Rejection No Bypass....................

19 3.2.2.3 Feedwater Controller Failure 19 3.2.2.4 Control Rod Withdrawal Error..................

20 3.2.2.5 Power Dependent LHGR Limit 20 3.3 Flow Excursion Analysis.................................

21 3.4 S a f e t y Lim i t..........................................

22 3.5 S umm ary of Re sults....................................

23 3.5.1 Power Dependent Thermal Limits and Values..............

23 3.5.2 Flow Dependent Thermal Limits and Values...............

24 3.5.3 Exposure Dependent Thermal Limits....................

24 4,0 MAXIMUM OVERPRESSURIZATION...............,,.............

34 4.1 D e si g n B a si s.........................................

34 4.2 Maximum Pressurization Transients.........................

34 4.3 R e s ul t s..............................................

35 5.0

-RdFERENCES..............................................

-36 APPENDIX A SINGLE LOOP OPERATION..............................

A1 l

l L

-l s1 l

EMF 91 168 i

Revision 2

.l Page il l

LIST OF TABLES Table gagg

.f 2.1 R E SULTS O F A N ALY S E S........................................

6 l

3.2 OPERATING LIMIT COORDINATES.................................

8 i

3.1 GRAND GULF UNIT 1 CYCLE 6 LFWH DATA

SUMMARY

25 t

LIST OF FIGURES Floure Egga

^

1.1 POWER / FLOW MAP USED FOR GRAND GULF UNIT 1-MEOD ANALYSIS.......

3-2.1 EXPOSURE DEPENDENT MCPR LIMITS FOR GRAND GULF 'JNIT 1 CYCLE 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 i

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...............

26 3.2 ~ LOAD REJECTION WITHOUT BYPASS (POWER AND FLOWS).............- 27 3.3 LOAD REJECTION WITHOUT BYPASS (VESSEL PRESSURE) 28 3.4 LOAD REJECTION WITHOUT BYPASS (WATER LEVEL ABOVE SEPARATOR SKIRT) 29 3.5 FEEDWATER CONTROLLER FAILURE (POWER AND FLOWC)...............

30 3.6 FEEDWATER CONTROLLER FAILURE (DOME PRESSURE),...............

31 3.7 FEEDWATER CONTROLLER FAILURE (WATER LEVEL ABOVE SEPARATOR SKIRT) 32 3.8 GRAND uULF UNIT 1 CYCLE 6 SAFETY LIMIT DESIGN BASIS

- LOCAL POWER DISTRIBUTION..............................,,

33 A.1 PUMP SElZURE EVENT SLO (POWER AND FLOWS)

A'.4 -

1'

' A.2 PUMP SEl2VRE EVENT SLO (VESSEL PRESSURE)

A.5

- A.3 PUMP SElZURE EVENT SLO (WATER LEVEL ABOVE SEPARATOR SKIRT)....

A6

l EMF 91 168 Revision 2 Pagelii l

An SPC investigation into the sensitivity of FWCF event severity to the water level in I the steam separator has led to the co'iclusion that the procedure used in past SPC FWCF j

l cnalyses is non-conservative relative to the benchmark cases for the SPC COTRANSA2 I methodology. SPC has established a new procedure which corrects this non-conservatism i end provides conformance with the approval basis for the methodology.

I l

Identification of the non conservatism required SPC to evaluate the impact upon I analyses performed for the Cycle 6 licensing c6mpaign for Grand Gulf Unit 1 as provided in l EMF 91 168 and EMF 91 169. The FWCF case with the least margin to MCPR operating i limits for Grand Gulf Cycle 6 operation (104.2%P/108%F st EOC 30 EFPD) has been I reanalyzed using the new procedure. The results of the analysis were used to assure that the 1 MCPR operating limit remains valid at the most limiting condition and to establish a bounding i increase in event delta CPR to be applied to results for all other cases. Based on_ this l reenalysis the event delta CPR at EOC 30 was shown to increase to 0.14 and a bounding i event increase of 0.02 was established. The updated results were incorporated into the l affected tables of results in Revision 1. The FWCF event results are now shown to be equal 1 in severity to the LRNB ea.nc at the EOC 30 and EOC+30 EFPD expesures used to sot the 1 exposure dependent MCPR operating limits. The FWCF is now shuwn to have a higher event I delta-CPR than the LRNB case at nominal EOC. However, the EOC transient results remain i less severe than EOC+30 EFPD results.

1 I

Revision 2 of this report is issued to reflect the changes in a more clear and complete l manner. Changes to text and tables from both Revision 0 and Revision 1 are indicated by l revision bars in the left margin of the report. Figure 2.1 has been revised to correct the l

l omission of a data point at EOC. Figure 2.2 has been revised to reflect only the EOC 30 l results at 104.2% powers Figures 3.5 through 3.7 have been replaced by results from the I reanalysis of the FWCF at 104.2%P/108%F and EOC 30 EFPD. The previous figures were I based on the original analysis procedure for the FWCF transient at nominal EOC.

r-

EMF 91 168 Revision 2 Page 1

1.0 INTRODUCTION

This report presents the results of analyses performed by Siemens Nuclear Power Corporation (SNP) for reload fuel in Grand Gulf Unit 1 Cycle 6 for operation within the Maxirnum Extended Operating Domain (MEOD). In Cycle 1 (Reference 1) the NSSS vendor performed extensive transient analyses for Grand Gulf Unit 1 in conjunction with the extension of the power / flow operating map to the MEOD. These analyses established conservative operating limits for MEOD operation. The initial reload of SNP fuel in Grand Gulf Unit 1 occurred in Cycle 2. In support of the initial reload of SNP fuel, extensive additional transient analyses were performed by SNP to justify the NSSS vendor operating limits and, where necessary, provide appropriate limits for SNP fuel using SNP methodologies (Reference 21.

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 employed for the Cycle 6 analysis include the CASMO 3G/MICROBURN B codes (Reference 7), COTRANSA2 system analysis methods (Reference 5), safetylimit methodology (Reference 9), and the use of the ANFB Critical Power Cntrelation (Reference 14) in XCOBRA and XCOBRA T.

The Cycle 6 transient analysis cor.sists of recalculation of the limiting transients at state points having the least margin to operating llmits to confirm that the effects of the Cycle 6 changes on transient results are small relative to available margin and/or establish appropriate limits. Reanalysis of the limiting transients for Cycis 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 power / flow 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).

l

- ~ -. - -. -

EMF 91 168 Revision 2 Page 2 Those analysos establish the Grand Gulf Unit 1 Cycle 6 Technical Specification MCPR limits at rated conditions, establish MAPLHGR limits for Cycle 6 operation, and establish revisod thermal limits for off rated conditions. Previous Grand Gulf reload analyses have demonstrated thst the maximum vossol pressure for the most limiting pressurization event varies over a narrow range essentially independent of fuel design. The evaluation of thoso analysos 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 i

were performed at limiting power / flow conditions. LHGR protection has been established for both 8x8 and 9x9 5 fuel in Cycle 6 at rated and off rated conditions. Power and flow dependent LHGR limits have been established for Cycle 6 using SNP methodology.

12 0 O State Points for Cycle 6 Transient Analyses a

10 0 "O3 o

O' 80 O

e 60 i

/

40

/ 0 eo U

=

20 0

O 10 20 30 40 50 60 70 80 90 100 110 12 0 m

Core Flow, Percent of Rated g5 2k3 28i una FIGURE 1.1 POWER / FLOW MAP USED FOR GRAND GULF UNIT 1 MEOD ANALYSIS

EMF 91 168 Revision 2 Page 4 2.0

SUMMARY

The results of the Grand Gulf Unit 1 Cycle 6 transient analyses support appropriate thermallimits for the Grand Gulf core including the ANF 1.5 9x9 5 teload. SNP thormallimits have been provided for MCPR, that are based on Control Rod Withdrawal 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 fuel satisfy the requirements specified by 10CFR50.46 of the U.S. Code of Federal Regulations. The 8x0 snd 9x9 5 LHGR limits will be protected at off rated conditions by applying LHGRFAO, and LHGRFAC, multipliers on the Technical Specification LHGR limits.

Table 2.1 summarizes 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 ilmit (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 asssssment of the channel bow impact using appropriate SNP methods (Refstence 9).

The plant transient and safety limit analyses results reported herein ostablish the power dependent Minimum Critical Power Ratio (MCPR,) limits. The power dependent Linear Heat

[

Generation Factor (LHGRFAC,)is presented for Cycle 6 operation for SNP 8x8 and 9x9 5 fuel types. The MCPR, limits, the LHGRFAC, values, and the corresponding results of SNP's analyses are presented in Figures 2.2 and 2.3.

The flow dependent Minimum Critical Power Ratio (MCPR,) limit and the results of l

SNP's analysis are presented in Figure 2.4. The flow dependent Linear Heat Generation Rate l

l

l I

i EMF 91 168 Revision 2 Page 5 Factor (LHGRFAC )is presented in Figure 2.5. These flow dependent LHGRFAC, values and MCPR, limits have been established for Cycle 6 to support the " loop,nanual" mode of operation. These curves are based on conservative maximum core flow rates. Table 2.2 shows the coordinates used to construct Figures 2.1 through 2.5.

The implementetion of the MCPR operating limit requires that the most restrictive operating limit be chosen from among the three MCPR curves based on exposure, flow, and power. Thus, the greater value of MCPR as given by MCPR., MCPR,, or MCPR,is selected as the operating limit in accordanca 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 presented in References 2,22,24, and 25. The results show that the Grand Gulf Unit 1 safuty valves have sufficient capacity and performance to protect the ASME Boiler and Pressure Vessel Code, Section lil, Class I, maximum vessel pressure transient limit of 1375 psig during Cycle 6.

The fuel related Teclinical Specification limits for Cycle 6 operation are included in the reload analysis report (Reference 3).

l l

l l

EMF 91 168 Revision 2 Page 6

.)

i TABLE 2.1 RESULTS OF ANALYSES i

t

)

l l

THERMAL LIMITS l

Transient Delta CPR Loss of Feedwater Heating (all conditions) 0.09 Control Rod Withdrawal Error (Reference 4) 100% power 0.10 I

70% power (1 foot ganged rod withdrawals) 0.18 70% power (2 foot ganged rod withdrawals) 0.34 20% power 0.48 l

l Feedwater Controller Failure Without Bvoans Delta-CPR (Limiting Fu6: Type)

% Power /% Core Flow EOC 30 EFPD EQC EOC + 30 EFPD l

104.2/108' O.14 0.17 0.18 I

40/108 0.38 0.39 l

l l.

104.2% power /108% core flow is used for the Reload Licensing Analysis (RLA) l-conditions to conservatively bound 100% power /105% core flow.

EMF 91 168 Revision 2 Page 7 TABLE 2.1 RESULTS OF ANALYSES (CONTINUED)

Load Relection Without Bvoass Delta-CPR (Limiting Fuel Type)

% Power /% Core Flow EOC 30 EFPD EOC EOC+30 EFPD 104.2/108' O.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" 0.60 0.59 104.2% power /108% core flow is used for the Reload Licensing Analysis (RLA) conditions to conservatively bound 100% power /105% core flow.

Direct scram on turbine trip disabled.

l I

EMF 91 168 Revision 2 -

Page 8 I

TABLE 2.2 OPERATING LIMIT COORDINATES GRAND GULF UNIT 1 Cvele 6 MCPR(e) Limits (Figure 2.1)

Core Average Exposure GWd/MTU MCPR{el 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(n) 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" i

Core flow s 50%

Core flow > 50%

.t

-r

-n

EMF 91 168 Revision 2 i

Page 9 TABLE 2.2 OPERATING LIMIT COORDINATES (CONTINUED) j 4

LHGRFAClo) Limits (Figure 2.3)

Percent of Rated Core Power LHGRFAC(n)

M M

i 100 1.00 1.00 70 1.00 1.00 40 0.69 0.75 i

25 0.69 0.75-MCPR(f) Limits (Figure 2.4)

Percent of Rated Core Flow MCPR(f) 20 1.28 130 1.28 65 1.20-105 1.20 t

I t

i e r,.

ea.-

-a---

.e-c-

,m

,w, gm-,

.. - -, - + - -.,1,#,_.:.y-,.-

w..,,,c---,-,-

.m-wm-....%

.-cc,,,,-

,-.e, r.e...<,,...ivwe.,

y,--

.L,,

EMF 91 108 Revision 2 Page 10 TABLE 2.2 OPERATING LIMIT COORDINATES (CONTINUED)

LHGRFAC(f) Limits (Figure 2.5)

Percent of Rated 9x9 5 J ote Flo w LHGRFAC(fl 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 0.954 50.0 0.900 40.0 0.846 30.0 0.792 20.0 0.792 Percent of Rated 8x8 Core Flow (HGRFActfl 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

1.5

~

0 Cycle 6 Anolysis 1.4 D

1.3 nS 1.25 e

a.

o t

Uy i.20 0

1.2 u

8 1.1 I

I I

I y

BOC EOC-30 EOC EOC+30 m

31 Core Average Exposure y13 e e. ;

e o _.

4 hp FIGURE 2.1 EXPOSURE DEPENDENT MCPR UMITS FOR GRAND GULF UNIT 1 CY'CLE 6

l 2.50 O

Cycle 6 Analysis a

CRWE Results 2.25 Core Flow > 50 %

2.00 2

o Core Flow < SO ?.

o y

a 1.75 U

2 O

1 50 3

1.25 o

e

- o O

1 1.00

^

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

,E m

?. da S.

e *_

• oo

_. :3 m FIGURE 2.2 POWER DEPENDENT MCPR LIMITS FOR GRAND GULF UNIT 1 CYCLE 6

EMF 91 168 Revision 2 Page 13 1.6 99 LHGAFAC(p)

,,,. O Cycle 6 Trenoient Anolysis D

4.2 3

o u

h t.o o

5 o

B O

o.s

' ^- '

o.4 o

to 20 30 40 so so to so to too tio 12 0 Core Power, Percent of Roted i

4 I

f Y

T T

I I

3 T

I 6 8 LHGRFAC(p) g,,, o Cycle 6 Tronelent Anolysie 1.2 o

3 U

t.o O

e 5

0.e O

o.s i

o 10 20 30 40 so so 70 so so too 110 12 0 Core P' v er, Percent of Roted FIGURE 2.3 POWER DEPENDENT LHGRFAC VALUES FOR GRAND GULF UNIT 1 CYCLE 6

1.6 r

G Cycle 6 Anclysis 1.5 1.4 O

v E

o_

1.3 U

1 0

0 1.2 D

O O

O O

1.1 1.0

^

^

^

^

O 10 20 30 40 50 60 70 80 90 100 11 0 12 0 Core Flow, Percent of Roted I

22 ?

1m o T. y eo3 m A PJ CD FIGURE 2.4 FLOW DEPENDENT MCPR LIMITS FOR GRAND GULF UNIT 1 CYCLE 6

6 i

t t

l I

k 1.1 1

'j i,

4 L

4

~.

1.0

[

.. L

?

+

~

/>

O.9 q+

n i

v u

S gS h<-

r

.o.e j

.r ea t

I t

i t

0.7 i

t o.s t

I.

0.5

^

^

1 0

10' 20' 30 40 50

-60 70 80

' 90 100 110 12 0 -

m Total Core Flow, Percent of Rated 3r 1.=

$ < *$3 E*

eo*

i 3 e

'r MMW.

j FIGURE 2.5 FLOW DEPENDENT LHGRFAC VALUE FOR GRAND GULF UNIT 1 CYCLE 6 3

e

~. -

,L,.

m e

.E,

..,m'1

-,,,-n-m.

-_-.,-.m.--

~

m.4...,

...+_.. _, _ - -.-..

EMF 91,168 Revision 2 Page 16 3.0 THERMAL LIMITS ANALYSIS 3,1 IntroductioD The scope of the thermal limits analysis includes system transients, localized core events, and safetylimit analysis. Results of these analyses are used to establish power, 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 analystss 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)s 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 steady state thermal hydraulic code used in the analysis of slow flow excursion events (MCPR,). The ANFB Critical 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 Thermallimits have bean appropriately 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 analyzed for Cycle C using COTRANSA2. The Load Reject No Bypass (LRNBL 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 pressurization analyses were performsd for Cycles 2,3,4, and 5 and were 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

f e

EMF 91 168 Revision 2 Page 17

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 Grand Gulf Unit 1 Cycle 6. These analyses show less restrictive results or little change from the Cycle 5 analyses due to Cycle 6 changes, thus justifying that the less limiting transients not analyzed for Cycle 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.

3.2.1 Deslan Bai!3 The LRNB and FWCF transients have been determined to be mott limiting at end of full power capability when control rods are fully withdrawn from the core. Between BOC and EOC 30 EF.

the CRWE transient is most limiting. From nominal EOC 30 EFPD to EOC + 30 EFPD, the LRNB and FWCF transients are limiting. The delta CPR calculated fr ;OC 30 EFPD, EOC, and EOC+30 EFPD is conservative for cases where control rods are partially inserted. The analysis for Grand Gulf Unit 1 with MEOD was performed using conservative rnalyticallimits 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 transfor disabled. Recirculation pump trip on high dome pressure was enabled for events initiated at core powers below 40%

rated.

3.2.2 Anticiented Transients SNP's transient methodology report for jet pump BWRs (Reference 19) considered eight categories of anticipated transients. The most lirniting 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:

i l

l

~

~

EMF 91168 Revision 2 Page 18 Loss of Feedwater Heating Load Rejection No Bypass Feodwater Controller Failu'e No Bypass Other transients are inher'"tly non limiting or bounded by one of the above as shown in the NSSS vendor MEOD m..r as 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.

3.2.2.1 kg3s Of Feedwater Heatina Analysis of the loss of feedwater heating event was performed to reflect reactor operation over the MEOD operating power versus flow map and conditions anticipated during actual Grand Gulf reactor operation.

Calculations performed for Cycles 1 through 5 essumed a conservative reduction of 100*F in the feedwater temperature. Results for Cycles 1 through 5 are provided in Table 3.1 of Reference 22. Table 3.1 provides the conditions for the cases analyzed in Cycle 6 in terms of cycle exposure, core power, and core flow. The initial and final MCPR values are presented for each case.

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

MCPR(initial) = -0.04974 + 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) u 0.02386 + 1.1021 ' SLMCPR i

1 l

EMF 91 168 Revision 2 Page 19 This bounding relationship and th~ Cycle 6 data are presented in Figure 3.1. Substituting the SLMCPR of 1.06 Se MCPR operating limit for the LFWH event for all operating conditions analyzed is 1.15.

3.2.2.2 Load Relacilon No Bvoass The Load Rejection No Bypass (LRNB) event is the most limiting of the class o' transients characterized by rapid vessel pressurization for Grand Gulf Unit 1.

The load rejection causes a fast closure of the turbine coritrol' Hves. The resulting compression wave travels through the steam lines into the vessel and creates the rapid pressurization condition.

A reactor scram and a recirculation pump transfer from high to low v,eed are initiated by fast closure of the control valves. Condenser bypass flow, which car. mitigate the pressurization effect, is not credited. The excursion of the core power due to void collapse is primarily terminated by reactor scram and void growth due to the recirculation p

• 2 high to low speed transfer.

Figures 3.2,3.3, and 3.4 present the response of various reactor and plant parameters to the LRNB event initiated at the Re!oad Licensing Analysis condition (104.2% power /108%

core flow). The MCPR operating limit of 1.20 is bounding for all exposures up to EOC 30 EFPD. The MCPR operating limit of 1.25 is bounding for all exposures between EOC 30 and EOC + 30 EFPD conditions. Table 2.1 lists the delta CPRs for this transient at the power / flow conditions and exposure conditions considered.

3.2.2.3 Feedwater Controller Failure The f allure of the feedwater controlier to maximum demand (FWCF) is the most limiting _

of the vessel inventory increase transients. Failure of the feedwater control system to maximum comand would result in an increase in the coolant level in the reactor vessel, increase ' feedwater flow results in lower temperatures at the core inlet, which in tum cause an !ncrease in core power level. If the feeriwater 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

EMF 91 168 Revision 2 Page 20 stop valve is closed to avoid damage to the turbine from excessive liquid inventory in the steamline. The high water level trip also initiates reactor scram, and subsequent turbine trip leads to recirculation pump high to low speed transfer. The core power excursion is terminated by the same mechanisms that end the LRNB transient.

Figures 3.5,3.6, and 3.7 present the response of various reactor and plant parameters to the FWCF without bypass event initiated at the Reload Licensed Analysis condition I (104.2% power /108% core flow) at an exposure of EOC 30 EFPD. These responses are I typical of respor'ses at other exposures. The delta CPRs calculated for this event at EOC 30 l and EOC + 30 EFPD are equal to the LRNB results and support the ettablished MCPR operating i limits. The cases of FWCF with bypass and with feedwater heaters out of service ( 100 'F) were analyzed in previous cycles and shown to be bounded by FWCF without bypass case.

3.2.2.4 Control Rod Withdrawal Error Reference 4 documents SNP's gencric CRWE analysis for Grand Gulf Unit 1 operation within the MEOD. The applicability of these analyses was confirmed by performing CRWE analyses with MICROBURN B using SNP's A.JFB critical power correlation.

Based on Reference 4 operating conditions and analytical procedures,1 and 2 foot CRWE events were simulated. De agns 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. Figuru 2.2 shows the 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 Limit Transient analyses have been performed to define appropriate multipliera on the fuel design limit LHGR for part power operation. The purpose of these multipliers is to protect fuel l

from failure due to centerline melt and exceeding the 1% plastic strain mechanical performance design criteria during off rated conJition transient events. Analyses were performed for the Load Rejection No Bypass (LRNB) and Feedwater Contrciler Failure (FWCF) l

EMF-91 168 Revision 2 Page 21 p essuriration event transients and the Control Rod Withdrawal Error event which is a localized event. Analyses performed for Cycle 2 shc.<ed the LRNB and FWI tansients to be limiting relativo to MCPR and LHGR increases. CRWE mnalyses 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 LhNB and FWCF transients were evaluated fer Cycle 6 c%iderina e variety of exposure and operating conditions. The result'

' ese analyse, are provided in Figure 2.3 and demonstrate odequate margin to the ope

,, Smit. Separate limits are established for SNP 8x8 and 9x9-5 fuel types ba ad upon the appropriate transient LHGR limit for each fuel type, l

3.3 Flow Excursion Analysis The flow excursion transient is analyzed 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 operstion analyzed for Grand Gulf Unit 1 Cycle 6 is " loop manual" only. This mode of operation corresponds to a single recirculation loop flow excursion event.

The results # the flow excursion transient analyses were used to establish new flow dependent thermal limit

• f MCaR,. For these analyses the change in critical power along the flow ascension path wm 4.alated 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 niaximum achievaL3 now. Figure 2.4 presents the MCPR, limit for rnaximum acnievable core flow, conservatively assur.Jng that the recirculation system equipment is capabl of 110% of rated flow on the limiting rod line. For f!ow rates less than 30% rated flow, the recirculation system operates at low speed restricting the maximum possible flow.

Because of this restrictien,. ' MCPR, curve conservatively remains fixed between 20% flow and 30% flow.

i EMF 91 168 Revision 2 Page 22 The Cycle 6 LHGRFAC, analysis was performed with the CASMO 3G/MICROBURN B neutrc*uc codes assuming a singic pump runup flow excursion. The analysia assumes that the recirculation flow incrossas slowly along the limiting rod line (Reference 2) with a maximum core flow capacity of 110% of rated. A series of flow excursion analyses were perfo ined starting from different initial power / flow conditions. Variations in the cycle exposure and control rod patterns were also considered.

The final conditions are conservatively dmermined based on the maximum attainable core flow rate. Xenon is conservatively L.sumed to remain constant during the event. The operating limits were established to bound the limiting results and are shown in Figure 2.5. Separate limits are established for SNP 8x8 and 9x9 5 fuel types based upon the appropriate transient LHGR limit for each fusi type. Because of restrictions in flow rates attainable for operation with core flows less than 30% of rated, the LHGRFAC, conservatively remains constant for core flow rates between 20% and 30%.

Safetv Limit The nfety limit MCPR is defined as the minimum value of the critical power ratio at u ch the fuel could be operated, with the expected number of rods in boiling transition not

.tceeding 0.1% of the fuel rods in the core. The safety limit is the minimum critical power l

ratio which would be permitted to occur during the limiting anticipated operawnal occurrence.

l The safety limit MCPR for all fuel types in Grand Gulf Unit 1 Cycle 6 operation was calculated

[

l to be 1.06 using the methodology presented in References 9 and 11. The determination of the safety limit explicitly includes the effects of channel bow and relies on the following assumptions:

1.

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

1 1

2.

The channel exposure at discharge will not exceed 40,000 mwd /MTU i

l based on the fuel bundle average exposure.

l 3.

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

l 4.

The limiting module contains a conservative exposure configuration.

l

1

)

EMF 91 168

+

Revision 2 Page 23 The input parameter values for uncertainties used in the refety limit MCPR analysis are unchanged from the Cycle 2 analysis presented in Reference 2 except for the uncertainties -

associated with the ANFB correlation, its implementation in the safety limit evaluation, channel bow,1nd 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 mode:ed in the safety limit evaluation.

l:

3.5 Summarv of Results 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 1.25 at rated conditions is shown -

from EOC 30 to EOC +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 operating limit of interest ir he larger of the three values for a given reactor operating condition.

3.5.1 Power Denendent Thermal Limits and Values The power dependent MCPR limh (MCPR,) protects against exceeding the safety limit MCPR during anticipated operational 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 and 1% clad strain during anticipated t y stem 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 f

~v

EMF 91 108 Revision 2 Page 24 establish the MCPR, operating limit for Cycle 6. The Cycle 6 LHGRFAC, values establish the applicable operating limits for SNP 8x8 and 9x9 5 fuel.

3.5.2 Flow Deoendent Thermal Limits and Values The flow dependent MCPR limit (MCPR,) protects apinst 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 Egagsure Deoendent Thermal Limits The exposure Cependent MCPR limit (MCPR ) protects against exceeding the safety limit MCPR ducing 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.

EMF 91 168 '

- Revision 2 Page 25 TABLE 3.1 GRAND GULF UNIT 1 CYCLE 6 LFWH DATA

SUMMARY

j Initial State' Final State Cycle Total Core Total Core Core Total Core Total Core Core Exposure Power Flow Minimum Power Flow Miniinum

.(SWd/MTl (MWt)

(M1b/hr)

CPR

_ (MWt)

(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 l=.427 4332 101.25:

'l.346 5.00 3833 104.63 1,430 4333 104.63

1.342 6.00 3833 99.00 1.397-4324 9940 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-j 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 1.3141 12.45-3833 112.50 1.409 4266

-112.50 1;338-13.27 3833 112.50 1.423-4261

112.50 1.353

3.5 Cycle 6 LFWH Analysis 4

3.0 9

/

D N\\

[.

2.5 b

U 7

0 1

y 4

c 2.0 i'

- g c

1.5

,,. i 1.0

1. 0 -

' 1.5 -

2.0 2.5 -

3.0 3.'i m-Finol' MCPR

=5 2 I. 6 a e. "*

eo3 u

m i

m N m --

FIGURE 3.1 ' ANALYSIS OF LFWH INITIAL M' CPR VERSUS FINAL MCPR e

=

s

.q

,my

's,gp' A

7'vFFw-

- +

e ir"

-4

"'---Ah--------

-""'-d--

550

+= Relative Core Power 50 0 -

x=

Relative Heat Flux o=

Relative Recirc Flow v=

Relative Steam Flow 250-a= Relative Feed Flow-200-15 0 -

10 05 -b lv-

, r 50-O-

' m5 2

10 0 oy ?

a O

0.5 1

' 1.5 2

2.5 3

3.5 4

4.5 5

7 g*

Time, seconds

.,8 g.

nue FIGURE 3.2 LOAD REJECTION WITHOUT BYPASS (POWER AND FLOWS)

, _ _ - - _ = - - - _ _ - - _ _ - _ _ ~

1500 1

1400-

.9 m

O. 1300-u3 in m

e.5

-e m 1200-m 1100-

,2 yg,,7 1000 0

f,3 j

25 i

is 4

4.5 3

Time, seconds i

RGURE 3.3 LOAD REJECTION WITHOUT syPASS (VESSEL PRESSURE)'

f' 4.50 4.25-s 4-w".

Y

>0 3.75-u2

- O5 3.50-

-e en

<n 3.25-3-

r m

fh 2.75 O

O.5 1

1.5 2

2.5 3

3.5 4

4.5 5

7 5. io. -

e w-Time, seconds e 5' 2.

3 p

on to M EMD-FIGURE 3.4 LOAD REJECTION WITHOUT BYPASS (WATER LEVEL ABOVE SEPARATOR SKIRT)

250 Relative Core Power 200-

_ _Relat_ive _He_at F_ lux

....R..e..la..t. iv e...R.e..c..i.rc...F..lo.. w......

Relative _S_ team Flow 150_

Relative Feed Figw 2

v-g iOO-.______________________--_------ ------------' 2.:y, g

-o

's.......

~~

I

~~~

ceo 50-j i

w*

g ii l

V E

O-

, -100 m

0 2.5 5

7.5 10 12.5 15 17.5 20

,g E E. 3 Time, seconds oS eo u o O h3 00 FIGURE 3.5 FEEDWATER CONTROLLER FAILURE (POWER AND FLOWS)

1275 i

1250-1225-1200-i Or, a m5-s fn g 1150-tt 112 5 -

110 0-1075-1050 0

2.5 5

I.5 1O

'12.5 1S d3 23 m

Time, reconds E'6 FWCF.

.Iy-$_

Steam Dome Pressure eo*

3 w

m FIGURE 3.6 FEEDWATER CONTROLLER FAILURE (DOME PRESSURE)

L

.+

e-6.00 5.75 -

550-

~

• +.=

.,. 5.25 -

_..e i

e 5.00 -

-.2 L

o 4.75 -

~

ON 4.50-

__o t

' M m

4.25-o>

4.00 -

3.75 -

4

.3.50' i

i i

i.

i i

i

.0 1

2 4

5 7-8 10

- 11 12 14 15 17 18 20 m

Time, seconds y5 y s. 6 ao E. y e o _.

(n) U On

. M PJ CD FIGURE 3.7 FEEDWATER CONTROLLER FAli.URE (WATER LEVEL ABOVE SEPARATOR SKIRT)

EMF.91-168 Revision 2 Page 33 C0NTR0L 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 0.907 1.047 0.989 0.814 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 1.050 0.970 0.990 1.031 0

D 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

EMF-91-16s Revision 2 Page 34 4.0 MAXIMUM OVERPRESSURl2ATION Maximum system pressure has been calculated for the containment isolation event (rapid closure of all main steam isolation valves) with an adverse scenario ss specified in the ASME Pressure Vessel Code for Cycles 2 through 5 (References 2,22,24, and 25). These analyses demonstrate that the Grand Gulf Unit 1 safety velves have sufficient capacity and performance to provent pressure from reaching the establist.ed transient pressure limit of 110% of design pressure (1.1 x 1250 = 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 13 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.

l 4.2 Maximum Pressurization Transients t

l Secping analyses described 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 all transients are avaluated on the same basis (without direct scram) because of the smaller steam line volume associated with MSIV closure. Though the closure rate of the MSIVs is substantially slower than turbine stop or control valves, the compressibility of the additional fluid in the steam lines associated with a turbine isolation causes these faster closures to be less severe. Once the containment is isolated, the subsequent core power production must be absorbed in a smaller volume compared to that of a turbine isolation resulting in higher vessel pressures.

l

i 2

EMF 91 168 Revision 2 Page 35 4.3'.

Resulte

- The maximum vossol pressures at the most limiting power / flow point for the previous cycles analyses demonstrate that the maximum vessel prcssure varies over ~a very narrow range (1271 psig to 1298 psig) independent of fuel and core design-and that sufficient margin, more than 75 paid,is available to accommodate the minor changes represented by the Cycle 6 reload.

't 1

1 1

Y

EMF 91-168 Revision 2 Page 36

5.0 REFERENCES

1.

Lester L. Kinmer, USNRO, Letter to O. D. Kingsley, Jr., MP&L, " Technical Specification Changes to Allow Operation with One Recirculation Loop and Extended Operating Domain," August 15,1986.

2.

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

" Grand Gulf Unit 1 Cycle 6 Reload Analvsis," EMF 91-16.2, Siemens Nuclear Power 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, Exxon Nuclear Company, Inc., Richland, WA, October 1986.

5.

"COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analysis,"

ANF-913(P)(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-84-105(P)fA), Volume 1, Exxon Nuclear Company, Inc., Richland, WA, February 1987.

7.

" Exxon Neclear Methodology for Boiling Water Reactors: Neutronics Methods for Design and Analysis," XN-NF-80-19(A), Volume 1 and Supplement 3, Exxon Nuclear Company, Inc., Richland, WA, March 1983.

8.

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

9.

" Advanced Nuclear Fuels Critical Power Methodology for Be".ing Water Reactor "

ANF-524(P)(A), Revision 2, and Supplements, Advanced Nuclear Fuels Corporation, Richland, WA, April 1989.

i 10.

" Grand Gulf Unit 1 Reload XN-1.3 Cycle 4 Mechanical Design Report," ANF-88-183(P),

Supplement 1, Advanced Nuclear Fuels Corporation, Richland, WA, August 1991.

l 11.

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

12.

" Grand Gulf Nuclear Station Unit 1 Revised Flow Dependent Thermal Limits,"

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

l

EMF 91-168 Revision 2 Page 37 13.

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

ANF-88152(P)(A) with Amendment 1 and Supplement 1, Advanced Nuclear Fuels Corporation, Rich;.'d, WA, November 1990.

14.

"ANFB Critical Power Correlation," ANF-1125(P)(A) and Supplements 1 and 2, April 1990.

15.

" Grand Gulf 1 ANF-1.5 Design Report, Mechanical, Thermal Hydraulic, and Neutronic Design for Advanced Nuclear Fuels 9x9 5 - Fuel Assemblies," ANF-91-080fP), _

July 1991.

16.

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

17.

Not used.

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, November 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 Fu.s Corporation, Richland, WA, August 1990.

23.

" Grand Gulf Unit 1 Cycle 6 Single Loop Operation LOCA Analysis Report," EMF 91-122, Siemens Nuclear Power Corporation, Richland, WA, August 1991.

24.

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

l 25.

" Grand Gulf Unit 1 Cycle 4 Plant Transient Analysis Report," ANF 88150, Advanced Nuclear Fuels Corporation, Richland, WA. November 1988.

l l

I

4 EMF 91 168-Revision 2 ~

.Page A 1 APPENDIX A SINGLE LOOP OPERATION Analyses have been provided by the NSSS vendor that demenstrate 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 bundle power levels and nodal power levels that are possible when both recirculation systems are in ooeration. The physical interdependence between core power and recirculation flow rate inherently limits the core toless 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 sing!e-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 pumo suddenly stops rotating This causes a rapid decrease in core flow, a decrease in the ra'.e at -

which 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 stoppeo 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 flow. Figures A.1, A.2, and A.3 show the graphical representation of important system parameters during the accident.

.m.

-e,.,,..

i EMF 91 168 Revision 2 Page A-2 Thermal hydraulic analysis using SNP safety limit methodology has shown that the two-loop MCPR l'mit provides the required protection below 70% of rated core power such that any postulawu fuel failures would not result in exceeding a small fraction (< 10%) of the 10CFR100 requirements.

A.2 MCPR SAFETY LIMIT For single-loop 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. SNP has evaluated the effects of these uncertainties using SNP safety limit methodology and determined that augmenting the two-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 It is appropriate to use the reduced flow and power two-loop operating MCPR and LHGRFAC limits for single-loop operations. The reduced flow MCPA limit is to protect against boiling transition during flow excursions to maximum flow. The reduced flow LHGRFAC is based on the heat flux increase associated with an Neursion 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 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 durirJ 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.

A.4 MAPLHGR 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 the peak clad temperature from a single-loop operation LOCA is bounded by the two-loop LOCA analysis. The application of

EMF 91 168 Revision 2 Page A 3 these limits is valid for average planar burnups of up to 50000 mwd /MTU and 55000 mwd /MTU for SNP 8x8 and 9x9 5 fuels, respectively (Reference 23).

80

+= Relatise Core Power 70J x = Relative Heat Flux o=

Reic tive Recirc Flow V=

Relo'tive Steam Flow a = Relative Feed Flow 60-D

< t

)

v 50-1 P'

u 40-u 5

30-7 20-m ygh 10 0

1 2

3 4

5 6

7 8

9 10 11 3 [5. to

{

Time, seconds

>ww FIGURE A.1 PUMP SEl2UPE EVENT SLO (POWER AND FLOWS) i

1015 1010-1005-

.9 E.1000-d 5

my 995-

-eom 990-N

'e>

985-980-m ygh 975 0

1 2

3 4

5 6

7-8

'9 10 11

$ <.,e g

Time, seconds

• 5' f.

D3 on mu=,

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

6 5.75-5.50-O

^15

>c 5.25-J L2 03 5-7 mm o>

4.75-

'50-

.n5 4.25 33 3 g. e?

=

'O

• O 1

2 3-4-

5 6

7 8

9 10 11 Time, seconds

• pfg FIGURE A.3 PUMP SEIZURE EVENT SLO (WATER LEVEL ABOVE SEPARATOR SKIRT)

EMF 91-168

. Revision 2 issue Date:

9/23/92 GRAND GULF UNIT 1 CYCLE 6 PLANT TRANSIENT ANALYSIS Distribution O. C. Brown R. A. Copeland L. J. Federico D. L. Garber N. L. Garner D. E. Hershberger M. J. Hibbard J. N. Morgan.

C. C. Roberts l

C. J. Volmer G. N. Ward Entergy Operations /S. L. Leonard (21)

Document Control (5)

L l'

l