ML18030A980

Rev 1 to Browns Ferry Nuclear Plant Reload Licensing Rept, Unit 2,Cycle 6.

ML18030A980
Person / Time
Site:	Browns Ferry
Issue date:	04/30/1985
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18030A977	List: ML18030A976 ML18030A978 ML18030A980
References
TVA-RLR-002, TVA-RLR-002-R01, TVA-RLR-2, TVA-RLR-2-R1, NUDOCS 8601060176
Download: ML18030A980 (33)

Text

TVA-RLR402 Revision 1 BROWNS FERRY NUCLEAR PLANT RELOAD LICENSING REPORT UNIT 2, CYCLE 6

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TVA-RLR-002 Revision 1 April 1985 RELOAD LICENSINQ REPORT FOR BROGANS FERRY UNIT 2, CYCLE 6 TENNESSEE VALLEY AUTHORITY April 1985

I. Introduction This reload lioensing report presents the results of the core redesign and safety analyses performed for Browns Ferry unit 2, oyole 6 operation. The original design is documented in referenoe 13 . The methodology and technical bases employed in the performance of these

\

analyses are discussed in referenoes 1-6.

Items specifioally.,addressed here inolude the nuolear fuel assemblies and core loading to be used in oyole 6, the reload oore nuclear design characteristios, the transient and acoident safety analysis results, and the proposed operating thermal limits.

The oycle 6 reload core vill inolude four Westinghouse QUAD+

demonstration a'ssemblies looated in nonlimiting oore peripheral locations. 'A complete description of th'e demonstration assemblies is contained in Westinghouse Report WCAP-10507 (reference 8).

II, Reload cle Information A. Design Basis Ezposures

l. Actual cycle 5 core average ezposure at end of cyole:

20 8 d ST

2. Minimum cycle 5 core average ezposure at end of cycle 3 . Assumed oyole 6 core average ezposure at depletion of ODOR End of full power capability

2 Revision 1 April 1985 B. Reload Fuel Assemblies Fua 1 ~e Irradiated cle Loaded Number SDRB284L, R2 3 36 PSDRB284L, R3 4 176 PSDRB2658, R4 5 80 PSDRB284L, R4 5 168 Now PSDRB284L, R5 6 300 QUAD+ Demo 6 4 764 Descriptions of the nuclear and mechanical design of the General Electric irradiated and new fuel assemblies to be loaded in cycle 6 are oontainod in reference 7. Tho nuclear, mechanical, and thermal-hydgaulic design descriptions for the Westinghouse demonstration assomblies aro contained in referonoo 8.

C. Reforonce Core Loading Pattern The reference loading pattern is the basis for all reload licensing and operational planning and is comprised of.tho fuel assemblies desi'gnated in item II.B of this, report. It is based on the best available prediction of the core condition, at the end of the previous cycle and on tho dosired coro energy oapability for tho reload cycle. The reference loading pattern is dosigned with tha intent that it will raprosont, as closely as possible, the actual coro loading pattern. Figuro 1 shows the reference core loading pattern for cycle 6.

3, Revision 1 AP ril 1985 Tho xeforence loading pattern includes four Westinghouse QUAD+

demonstration assemblies loaded in peripheral looations.

C Evaluations performed by Westinghouse (referenoe 8) indicate that the results of licensing analyses for tho lead PgrgR fuel assembly'ound those fox the QUAD+ demonstration assemblios.

Cycle speoific analyses performed by TVA confirm this conclusion.

Special Conditions I

The use of increased oore flow (ICF) is piannod for cyole 6 operation. Safety analyses wero performed for both 100 pexcont arid 105 'percent of rated. core ',flow with tho most conservative results used-for determining the operating limits.. The conclusions regarding LOCA analysis, reactor internals pressuro drop, and flow-induced vibration as discussod in refexence 9 are h

applicablo to cycle 6. The flow-biased instrumentation fox the xod block monitor will be signal clipped for a'setpoint of 106 percent since flow rates higher than rated would otherwise result in a ACPR higher than reportod for the rod withdrawal error.

4 Revision 1 April 1985 XXX. Nuclear Desi n Characteristics A. Shutdown Margin The reference core is -analyzed in detail to ensure that adequate shutdown margin exists. This section disousses tho results of core oalculations for 'shutdown margin (inoluding tho liquid poison system).

1. Coro Effootivo Multiplioation and Control Rod Worth Core effective multiplication and oontrol xod worths wore oaloulated using the TVA BWR simulator code (references 2 and 4) in conjunction with tho TVA lattioe phys'ics data genoration code (references 3 and 4) to determine the core reactivity with all rods withdrawn and 'with all rods inserted. A tabulation of the results is provided in tablo 1. These threo oigenvalues (effective multiplication of the core, uncontrolled, fully controlled, and with the strongest rod out) were oalculated at the beginning-of-cycle 6 oore avorage exposure oorresponding to,tho'ctual ond-of-cyolo 5 core average exposure. The core was assumed to be. in a xenon-free oondition.

Cold keff was calculated with tho strongest oontrol rod out at various exposures thxough the oyole. Tho value R is tho difference between tho strongest .rod out koff at BOC and tho maximum oalculatod strongest rod out koff at any

5 Revision 1 April 1985 exposure p'oint. The mazimum strongest rod out keff at any exposure point is equal to or less than:

'SRO Maximum k kSRO (BOC) + R 2.'eaotor Shutdown Margin Technical Speoifications require that the refuelod core must be capable of boing made subcritical with 0.38-percent Ak margin in the most reactive condition throughout the subsequent oporating cyole with the most reactive oontrol rod in its full out position and all other rods fully inser'tod.

The shutdown margin is determined by using the BWR simulator code to oalculate the ooro multiplioation at seleoted exposure points with tho strongest rod fully <<lthdrawn. The shutdown margin for tho reloaded core is obtained by SRO subtraoting the mazimum keff ff from the oritical kefff of 1.0, resulting i'n a calculatod minimum cold shutdown margin of

.1.0-percent Ak for Browns Ferry 2', cyole 6.

Table 1 CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL ROD WORTHS NO VOIDS, NO XBNON, 204C Unoontrolled, KU (BOC) 1.116 eff Fully Controlled, KON (BOC) 0.954 eff Strongest Control Rod Out, KS (BOC) 0.982 eff R, Maximum Increase in Cold Core Reactivity 0.008 With Ezposure Into Cycle, Ak

6 Revision 1 April 198$

8. Standby Liquid Control System Tho standby liquid control system (SLCS) is designed to provide the oapability of bringing the reactor, at any time in a cyole, from full power and a minimum control rod invontory (which is defined to bo at the peak of the xenon transient) to, a suboritioal condition with the reactor in tho most reactive xonon-free state.

The SLCS shutdown margin is determined by using the BWR simulator code to oalculate the oore multiplication f'r the cold, xonon-free, all rods out c'onditions at the exposure point of maximum cold reactivity with tho soluble boron concentration given in the Technical Specifioations. The resulting R-effective is subtractod from the critical k-effootive of 1.0 to obtain the SLCS shutdown margin. Tho results of the SLCS evaluation are given in t'able 2.

Tablo 2 STANDBY LIQUID'CONTROL SYSTEM CAPABILITY Shutdown Margin (Ak)

PPM 204C Xenon Free 600 0;018 B. Reaotivity Coeffioients The reactivity ooef f ioients associated with the nuclear design of Browns Perry, 2, cyole 6 aro implicit in the 1-Doross sections

7 Revision 1 April 198S used for" the safety analyses. As suoh, reaotivity coeffioients are not separately calculated for input to the transient analyses. ,However, a void ooeffioient =is -generated in the 3-D to 1 D cross,se'ction collapsing prooess and is used as a verif ication check-. For Browns Ferry 2, cycle 6 the following results for DOR conditions were obtained:

100% core flow -0.0724 %4k/%void IOS% core flow -0.0731 %hk/%void C. Fuel Performance

."'The Browns Ferry 2, oycle 6 fuel performance is predicted by f

'rojecting the fuel burnup to the end of cyole with the 3-D simulator code. The oalculated peak pellet exposures for the various fuel, types are less than the limits'peoified in references 7 and 8. Furthermore, peak linear heat rates satisfy the assumptions made in the fuel vendors'hermal-mechanical integrity analyses (xeferences 7 and 8). All fuel types loaded* in cycle 6 are predicted to operate within these bounding assumptions. Additionally, the QUAD+ demonstration assemblies are predioted to have substantial margin to the lead PgxSR assembly in steady state bundle power and thermal limits throughout oycle 6 (figures 20-22). The minimum margin for bundle power is 32 peroent; for MCPR the minimum margin is S9 percent and for LHGR'it is 34 percent. Also,'. the QUAD+ assembly is always less limiting than the diagonally adjaoent fresh PgxSR bundle in oore looation 13-48.

8 Revision 1 April 1985 XV. Transient Anal ses A. Pressurization Events The KHKLN computer codo (reference 10) is used to analyze both the reactor system and hot channel xosponses during core-wide pressurization transients. Tho analytic models usod in these analyses are described in reference 5. A description of the CPR oorrelation and its applioation to Browns Ferry is contained in referenoe 11. Analyses aro pexformed for the potentially limiting events at the most advexse initial conditions ezpooted during the cycle. Reload unique initial conditions and transient analyses results are summax'ized in .the following tables.

NSSS Initial Conditions Steam Flow Core Flow Oap Conductance

~Bz osages DOR 105 105 721 Hot Channel Initial Conditions Limitin Event Fuel Bundle Bundle Oap Conductance 1 BTO fts-hx-4F PSXSR= 1.271 6.592 123 .0 1.051 1287

9 Revision 1 April 1985 Pressurization:Event Anal sis Results Peak Power Peak Heat Peak Vessel ACPR~ " System Transient Press sia PSxSR ~Res esse Load 366.8 1232. 9 0.201 Figures Rejection 2-5 w/o Bypass Feedwater 257.4 116.6 1215.1 0.152 Figures Controller 6-9 Failure

~Results presented were calculated for PgzSR fuel and will be conservatively applied to SxSR.

10 Revision 1 April 1985 B. Nonpressurization Events The nonpressurixation events analyzed for reload licensing are either steady state events or relatively slow transients that can be analyxed in a quasi-statio manner usin'g a 3-D:BWR simulator (reference 2). The methods used to.analyxe these events are described in referenoe 1. Results are, summarized below.

Non ressuri ation E ent Anal sis Results Event Loss of 17.6 Peedwater Heating (1004P)

Rod Withdrawal 0.18K Error Rotated Bundle 0.155 Error Mislooated Bundle 14.6 Error Por increased oore flow based on a signal clipped rod block setpoint of 106 percent.

Includes 0.02 penalty required when using the varible water gap method (reference 7).

V Results presented were calculated for the PSxSR fuel type and oonservatively bound the result's calculated for the SxSR fuel type and for the QUAD+ demonstration assemblies".

11 Revision 1 April 198$

C. Overpressure Protecti,on The main steamline isolation valve closure with failure of direot scram is analyzed to demonstrate sufficient overpressure protection (peak vessel pressure must be less than 110 percent of design pressure '-'390 psia). The event is analyzed using the models and methods described in referenoe 5. Results are summarized below.

MSI C osure Flux S ram Resu ts Peak Vessel Peak Steamline System Pressure sia Pressure sia ~Res onse 1276.7 1234.2 Figures 10-13

12 Revision 1 April 1985 V. MCPR"'0 eratin Limit Summa The methods used to determine the required OLMCPR values for each event analyzed are described in references 1 and 5. The applioation of Option A and B limits in determining the, cycle OLMCPR is. dose'ribed in

'he unit Teohnioal Specifications. Results are summarized below and in figure 14.

OLMCPR for Pressurl ation E ants BOC6-EOC6 0 tion Ai 0 tion Bs P8zg Sxg UAD+ Pgzg Sxg UAD+

Load Rejection Without Bypass 1.29 (OLRWOB)

Feedwater Controller Failure (FWCF) 1.24 OLMCPR for Non re'ssurization E ents BOC6-EOC6 Pgxg 8xg UAD+i Loss of Feedwater Heaters (LFWH) 1.25 Rod Withdrawal Error (RWE) 1.25 Rotated Bundle Error (RBE) 1.22 Mislocated Bundle Error (MBE) 1.22 z Results presented were calculated for, PSzSR fuel type 'and conservatively bound the results calculated for the SxSR, fuel type. The QUAD+

demon'stration assemblies will be loaded into nonlimiting core locations and monitored to the same OLMCPR.

13

'Revision 1 Apri'1 198S VI. Accident Anal ses A. Loss of Coolant. Aocident (LOCA)

LOCA analysis xesults for fuel types previously loaded in unit 2 are describe'd in reference 12.'efexence 8 indicates that the MAPLHGR limits for fuel type P8DRB284L 'can be conservatively applied to the QUAD+ demonstration assemblies. These limits are'resented below.

\

LOCA Limits fox UAD+ Demonstration Assembl es

.Average Planar MAPLHGR Ex osuxe MWd t ~tW fa 200 ~ 11.2 1,000 11.3 5,000 11.8 10,000 12.0 15,000 12.0 20,000 11.8 25,000 11.2 30,000 10.8 35,000 10.0 40,000 9.4 B. Rod Drop Accident (RDA)

The methodology used to analyze the rod drop accident is described in appendiz-,A of referenoe 6. Results for BF2, cycle 6 are summarized below.

Results for the Limitin RDA Condition: COLD (684F), EOC Exposure Rod Worth: 1.04% Ak Rod'Position: 10-19 Peak Fuel Enthalpy: 152 cal/gm Core Response: Figures 15-18

14 Rev is ion 1 April 1985 VII. tabi it Anal ses The methodology used to analyzo core a'nd ohannel stability -is described in appendix B of reforence 6~ Tho minimum stability margin ocours at the interseotion, of the natural circulation line and the 105-poroent rod line (the flow biased scram line also passes through this point). Results for BF2, cyclo 6 are summarized below and in figure 19

'tabillt Anal sis Results at Limitln Initial Conditions Maximum

~Anal nin Core Stability 0.85 Channel Stability PgxSR/SxSR/QUAD+ 0 '9~

x Results prosented aro for tho PgxSR fuel type and oonservatively bound the SxSR fuel type and the'UAD+ demonstration assemblies.

References 1.' TVA-EG-047, 'TVA Reload Core Design and Analysis Methodology for the Browns Ferry Nucleax Plant,'ennessee Valley

'uthority, January 1982.

2. TVA-TR78-03A, 'Three'-Dimensional LWR Core Simulation Methods,'ennessee Valley Authority, January 1979.

3. TVA-TR78-02A, 'Methods'or the Lattice Physics Analysis of LWRs, 'ennessee Valley Authoxity, April 1978.

4. TVA-TR79-01A, 'Verification of TVA Steady-'State BWR Physics Methods,'ennessee Valley Authority, January 1979.

5. 'VA-TR81-'01, 'BWR Transient Analysis Model Utilixing .the REIRAN Program,'ennessee Valley Authority, December 1981.

6. =. 'TVA-RLR-001, 'Reload Licensing Report for Browns Ferry Unit 3, Cycle 6, 'ennessee Valley Authority, January 1984.

7 ~ NEDE-24011-P-A-6, 'Genexal Electric Standard Application for Reactor Fuel, 'eneial Electric, April 1983.

8. WCAP-'10507, .'QUAD+ Demonstration Assembly Report,'estinghouse Electric Coxporation, Maxch 1984.

9. NED0-22245, 'Safety Review of Browns Ferry Nuclear Plant Unit No. 2 at Core Flow Conditions Above Rated Core Flow During Cycle 5,'eneral Electric, October 1982.

10. EPRI NP-1850-CCM, 'RLTRAN02 A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Power Reseaxch Institute, May 1981.

Systems,'lectric NEDE-24273, 'GEXL Correlation Application to TVA Browns Ferry Nuclear Power Station,'eneral Electric.

12. 'EDO-24088-1 (as amended), 'Loss-of-Coolant Acoident Analysis fox'rowns Ferry Nuclear Plant Unit 2,'eneral Electric, February 1978.

13. TVA-RLR-002, 'Reload Licensing Report for Browns Ferry Unit 2',

Cycle 6,'ennessee Valley Authority, July 1984.

Revision 1 April 1985 REFERENCE LOADING PATTERN BROWNS FERRY UNIT 2 - CYCLE 6 60 A 8 8 8 BBB 8 8 BBBBA 58 8 E 8 E BEE E E BEBEB 56 ABBEB E 8 EDE E D EBEBE 8 A 64 8 BEDE C E CEC C E C,ECED 8 8 62 BEOED E C ECE E C E CE,DE 0 EBA AAA QDE DE D OEDEO DQBA 60 48 46 44 8 BE DEDED A 8 E D A 8 EDE E.DE DE OE DBC E-D E E

D DE 0 EOE OE 8 B' DE D

E D E

E'DE DE DEOED BDBCB D E DEBB DE DEB DE DE A

8 A 42 8 E 8 E EDECE 8 D 8 D E D E,C C E E D DEDEC D E DEDEB 40 BBECE OE DBD 8CE BDEDB EDECE 8 8 38 E D E C D B.E E,DE DE CEDEO DECEB 36 .8 ECE 8 E D E C E 8 8 E E 8 DE DBD EBE E CBDB EDECE 34 8 8 8 C E 8 D 8 8 E OE DE BEG BECED DE CED 32 E D E C E C E E 8 ECE DEBEC 8 ECB 8 C EBECE EDECE E 8 30 8 E ECE DE BEG E

ECB EBECE EDECE 8 E DE C EOE DE E 8 8 E,C C E 8 C B,ECED DE CEO E 8 28 26 8 E 8 8 ECE OEDBD C E 8 C EBE 'E 8 E CBDB D EDECE 8 8 EDEDE CE8 CEDEO DECEB 22 24 8 E 8 E D E 8 E ECE OE DBD 8CE BDEDB EDECE 8 8 8 8 8 E 8 E D EDEC,E E D E C D DEDEB 20 18 A 8 E D.E 0 E D B.C 0

8 D E DE 8 DE C C E E D OE DE BOB CB C

D E DEOE E 8 8 A 16 8 E D E DE DE D E DEB 8 E DE DED DE DEB A 14 8 BE DEDED E D EOE E D E DE DE D E DE 8 8 12 A 8 A QDE DE D E DED D E D E.D E'D D0 8 AA 10 ., BE DED ECE ECED'E D EBA 8 8'BE OE C E CEC C E CECED 8.8 6 ABBEB E 8 EOE E D EBE BE 8 -A ABE 8 E 8'E E BEBEB A 2 A A 8 8 8 8 8 8 8 BBBAA

'l I .I 3 7 9

11 13 16 17 19 21 23 26 27 29 31 33 36 37 39 41 43 46 47 '1 63'7 49 56 69 FUEL TYPES A= SDRB284L,R2 8= PSDRB284L,R3 C=.P8DRB265H,R4 D= PSDR8284L,R4 E= PSDRB284L,R6 Q= QUAD+DEMO,R6

Revision 1 April 1985 FIGURE 2 GENERATOR LOAD REJECTION W/0 BYPASS 400 300 Legend TOTAL POWER (%)

AVE SURFACE HEAT FLUX (%)

CORE INLET FLOW (%)

INLET SUBCOOLING ()I')

'ORE 200 100 0

0 3 4 6 TIME (SEC)

FIGURE 3 GENERATOR LOAD REJECTION W/0 BYPASS 200 Legend VESSEL PRESS RISE IPSI)

TOTAL S/A VALVE FLOW (8) 160 BYPASS VALVE FLOW (5) 100 I

I I

I 50 I I

I I

I I

I I

0 3 4 6 TIME (SEC)

Revision 1 I

April 1985 FlGURE 4 GENERATOR LOAD REJECTlON I/O BYPASS*

160, 100 I

( I $l

( IY I ( I ir

( I I I I 60 I I I I I

I I

I I

0 '.q ~ ~ ~ I~ '~ ~ I~ 4

~ ~

I I I I I I

I II Il Legend'EVEL I (INCH.REF SEP SI(IRT)

I I

-60 III VESSEL STEAM FLOW ()()

I I TURBINE STEAM FLOW (%)

II II FEEDWATER FLOW I

()(I'100-0 3 4 6 TIME (SEC)

FIGURE 5 GENERATOR LOAD REJECl;ION WlO BYPASS

.2 Lege'nd TOTAL REACTIVITY (S)'CRAM REACTIVITY ($ )

-2 0 06 1.6 2 TIME (SEC)

Revision 1 April 1985 FIGURE 6 FEEDWATER CONTROLLER FA'ILURE 300 Legend TOTAL POWER (96)

AVE SURFACE HEAT FLUX (%)

CORE INLET FLOW (5)

CORE INLET SUBCOOLING (%)

200

~g 100 ~

~

0 0 . 10 16 20 26 TIME (SEC)

FIGURE 7 FEEDWATER CONTROLLER FAILURE 160 Legend,-

VESSEL PRESS RISE IPSI) 100 TOTAL S/R VALVE FLOW (%)

BYPASS VALVE FLOW (%)

I I I

I I

60 I I I I I

~ ~ 'I ~ ~ ~ ~ ~ ~ ~ I ~

I I I \,

I I

0

-60 0 10 16 20 26 TIME (SEC)

0 Revision 1 April 1985 FIGURE 8 FEEDWATER 'GONTRQLLER'AILURE 160 100 l II

)I,

'III I

I I

I 60 I)(I I Legend LEVEL (INCH-REF-SEP SKIRT) 0 VESSEL STEAM FLOW (%)

~ ~

TURBINE STEAM FLOW (%)

FEEDWATER FLOW (%)

-60 0 10 16 20 26 TIME (SEC)

RGURE 9 FEEDWATER CONTRQLLER FAILURE 0

-10 Legend TOTAL REACTIVITY (S)

SCRAM REACTIVITY (8)

-16 16 17 19 TIME (SEC).

Revision 1 April 1985 FIGURE 10 MSIV CLOSURE (FLUX SCRAM) 600 400 Legend TOTAL POWER (%)

AVE SURFACE HEAT FLUX (5)

CORE INLET FLOW (%)

300 CORE'INLET SUBCOOLING (%)

200 100 >tee ~ ~ V ly 4l 0

TIME (SEC)

FIGURE 11 MSIV CLOSURE (FLUX SCRAM) 260 Legend VESSEL PRESS RISE (PSI) 200 TOTAL S/R VALVE FLOW (%)

BYPASS VALVE FLOW (%)

"'60 100 I

I I

60 I I

I I

I I

0

-60 I 0 4 TIME (SEC)

4 Revision 1 April 1985 4

FIGURE 12 MSIV CLOSURE (FLUX.SCRAM) 160 Legend LEVEL IINCH-REF-SEP SKIRT)

VESSEL STEAM FLOW I%)

100 TURBINE STEAM FLOW"I9L) 11

( FEEDWATER FLOW (%)

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I I I I I 60 I I

(

( II, (I'I I I'(

I(l(1 I (

'I (

I

~ ~

0 I( II 1(

~ ~ 0 II ll

-60 0 4 TIME '(SEC)

FIGURE 13 MSIV 'CLOSURE (FLUX SCRAM)

Legend TOTAL REACTIVITY ($ )

SCRAM REACTIVITY ($ )

0

-2 I

1 1

\

-4

'0 2 3

'IME (SEC)

~

Revision 1 April 1985 FIGURE 14 OLMCPR FOR PSXSRlSXSRlQUAD+

1.38 1.34 1.32 1.32 1.30 GLRWOB 1.28 FWCF 1.26 1.26 AWE 5 LFWH C~

~~~ r '.24 r

/ K r

1.23 =-'BE 5 MBE 1:22 1.20 0 0.2 0.4 TAU" 0.6'.8 "SCRAM SPEED INTERPOLATION PARAMETER AS DEFINED IN THE TECHNICAL SPECIFICATIONS

Figure 15 BFRGYS Rod Drop Aooldent 1000 e000

&000 4000 a

3000 Reactor PoMer 2000 1000 0

0 0.5 1.$ 2 25 S.5 4.5 Time (See)

Fllgure 38 BFRCYS Rod Drop Aoelldettt 40 CS aO 80 S

D CL E

+ 20 Core Avera e Tee erat,use Ri.ee 10 0.5 1.5 2 24 $ .5

. Time (Sec)

Hgure 17 SFRCYS Rod Drop Accident 200

-200 40

-400

.. O

-100 gore Reacti.vi.t

-'I00

'-1000

-1200 0.5 1.5 2 2d 4.5 Time (Sec)

160 140 120 100 E

CO CP IO CL eo g

Maxi.mum PLn Ent,hoL 40 20 0

0 0.d 1.5 2 24 4.d Time (Soo)

FIGURE 19 DECAY RATIO VERSUS REACTOR POWER

~ WITH CONSERVATIVE ADDER 0.8 NATURAL CIRCULATiON 106% ROD LINE 0 0.6 I~

0-0 0.4 Cl 0.2 0,,

0 20 40 60 80 100 120

% POWER

IFigure 20 Bundlle Power Cornparlson: QUAD+ vs Lead Bundle 1.6 1.6-0 ~ . ~

1.4 W~W ~-

1.3

-o 12

~a 5

0.9 0.8 0 DEMO BUNDLE 0.7

~ LEAD BUNDLE 8 ADJACENT BUNDLE 0.6 0 .2 3 6 6 Cycle Exposure (GWD/MT}

Figure 21 MCPR Cornperison: QUAD+ vs Lead Bundle 3

2.8 0 =DEMO BUNDLE 4 LEAD BUNDLE 2.6 4 ADJACENT BUNDLE 2.4 2.2 4

1.8 g~ 4

~4 4~,.4 ~ ~4~ ~4 1.6 4 4

~~

~=e o~e. I~E:srl:N qpL I e I ~ ~

OPERATING LII1IT 1.2 0 3 4 6 - 6 Cycle Exposure {GWD/MT)

Figure 22 MLHGR Comparison: QUAD+ vs Lead Bundle 14 LIMIT I~I OPERRTING I ~0= ~ = ~ OESIGN GOAL

~ ~ ~' ~ ~ ~ =~

~

10 lK C9 x

8 0 DEMO BUNDLE

~ LEAD BUNDLE

'a ADJACENT BUNDLE 0 ,3 6, Cycle Exposure (GWO/MT)

B