ML19308B519

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Cycle 2 Reload Rept.
ML19308B519
Person / Time
Site: Oconee Duke energy icon.png
Issue date: 02/28/1976
From:
BABCOCK & WILCOX CO.
To:
References
BAW-1425, NUDOCS 8001090525
Download: ML19308B519 (50)
v • d • e


Text

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3AW-1425 S?

February 1976

?000 OCONEE UNIT 2. CYCLE 2

- Reload Report -

THE ATTACHED FILES ARE OFFICIAL RECORDS OF THE OFFICE OF REGULATION. THEY HAVE BEEN CHARGED TO YOU FOR A LIMITED TIME PERIOD ANS MUST BE RETURNED TO THE CENTRAL RECORDS STATION 008. ANY PAGE(S).

REMOVED FOR REPRODUCTION MUST BE RETURNED TO ITS/THEIR ORIGINAL ORDER.

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DEADLINE RETURN DATE -

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_m l M% MARY JINKS CHIEF CENTRAL RECORDS STATION

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BAW-1425 February 1976 OCON C CEIT 2. CYCLE 2

- Reload Rep 5t -

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i BABCOCK & WILCOX

( Power Generation Group Nuclear Power Generation Division P. O. Box 1260 t

Lymchburg. Virginia 14505 l

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CONTENTS Page

1. INTRODUCTION AND

SUMMARY

. . . . . . ... . .. . . . . ..... 1-1

2. OPERATING HICTORY . . . . . . . . ., .. . .. .. . . ..... 2-1
3. GENERAL DESCRIPTION

. . . . . . . ... . .. . . . . . ..... 3-1 4 FUEL SYSTEM DESIGN . . . . . . . . . . . . . . . . . . . .....

4-1 4.1. Fuel Assembly Mechanical Design ... . .. . . . ..... 4-1 4.2. Fuel Rod Design . . . . . . .. .. . ... . . . .....

4.2.1. Cladding Collapse 4-1 4.2.2. Cladding Stress

. . ... . . .. . . . ..... 4-1 i . . . .. . . . . . . . . ..... 4-2 4.2.3. Fuel Pellet Irradiation Swelling .

{ 4.3. Thermal Design . . . . . . .. ... . . .. ..

4-2 4-3

'4 4.3.1. Power Spike Model

'I 4.3.2. Fuel Tempera:ure Analysis

. . ... . . ... . . . .... 4-3 1

4.4.

. . . . . . . . . .... 4-3 Material Design 4.5. Operating Experience .

. . . . . . ... . . . . . . . . . .... 4-4

. . . .. . . . . . . . . . .... 4-4

5. NUCLEAR DESIGN . ..........

... . . . .. . . . .... 5-1 5.1. Physics Characteristics 5.2. Analytical Input

. . . ... . . .. . . . . .... 5-1

5. 3. . . . . .. . .. . . . . .. . . . .... 5-2 Changes in Nuclear Design . .. .. . . . . . . . . .... 5-2
6. THERMAL-l!YDRAULIC DESIGN . . . .. . ... . . . . . . . . .... 6-1 6.1. Thermal-Hydraulle Cesign Calculations . . . .. . . .... 6-1 6.1.1. Introduction of Mark B-4 Assemblies ... . ....

6.1. 2. Increased FC System Flow . 6-1 6.1.3. B&W-2 DNB Correlation

. .. . . . . . . .... 6-1 6.2. DNBR Analysis .... .. . . .. . .... 6-2

. . . . . . . . . ... . .. . . . . .... 6-2

7. ACCIDENT AND TRANSIE'.T ANALYSIS . . ... . . .. . . . . .... 7-1 7.1. General Safety Analysis . . . . ... . . . . .. . ....

7.2. 7-1 Rod Withdrawal Accidents . . . ... . . . .. . . . .... 7-1 7.3. Moderator Dilutfon Accident . .. . . . . . . .. . .... 7-2 7.4 Cold Water (Pump Jeartup) Accident . . . . . .. . . .... 7-3 7.5. Loss of Coolant Flew .

7.6 . . . . . .. . . . . . . . . .... 7-3 Stuck-Out, Stuck-In. or Dropped Control Rod . . . . .... 7-4 7.7. Loss of Electric Power . .. . . . . .. . .. .. . .... 7-4 7.8. Steam Line Failure . 7-5 7.9. Steas Generator Tube.Failure . .. . . .. . . . . .. . . ....

7.10. Fuel ibndling Accident . . . . ... . . . . . . . . .... 7-5 7.11. Rod Ejection Accident . . .. . . . . . . . .... 7-5

. . . . ... . . . . . . . . .... 7-5

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LoNTENTS ( Cen t_' d._a

?.s a e 7 . 1.' . Siax!=un liyPothet ical Acc ident . . . . .. . ... .....

7. ; 5 'Ja t e G.i.4 7:.nk Rupture . . . .. . . . .. ... .....  ?-*

7.1i. Li tA An.. l y.,1 6 ... . . . . . . .. . . . ..... .... J-PROPOSF.D MODIFICaT!oNS To TECIINICAL SPECIFICATIONS . .. ..... s-:

's . SIARTUP Ph0GRA.'l . . .. . . . . . ... . . . ..... ..... -1 REFERENCES . . . . . .. . . . . . . ... .. . ... . ..... A-!

s L_ist of Tables Lihte

'. - ! . fuel Design Paraceters .

. . . . ... . . .... .. ..... J.- 5

'..'. fuel Rod Dimensions- . . . . . ... . .. . .. . .. ..... ". - 5

'. - 6 . Input Sumary for Cladding Creep Collapse Calculations ..... ~. - 6

. . . Fur! Thernal An.ilysis Parameters . ... . . . .... ..... '

.-6

>- l . Oconer 2. Cyc i s- 2 Phys.ics Parancters . . . . ... .. ..... 5- 3

's - J . Shandown .'Lirgio Calculation - Ocenec 2. Cycle 2 ... ..... 5- 5

'i- l . Cn !c I and 2 tiximus Design Conditions . .. ... . ..... e>- 4 7-1. Conp.ir l son of th y Petrancters for Accident Analysis ... .... 7-8 List of Figures Fi gui'c l- l . oconee 2. Cycle 2 -- Core Loading Diagram . . .... ...... 3-3 1-2. Oconee 2 Enricitment and Burnup Distribution for Cycle 2 .... 3-4 1- 3. Oconee 2. Cvele J - Control Rod Locations

-l.

... ... . .... 3-5

>Lixinun Cap Size Vs Axial Position - Oconee 2. Cycle 2 . .... J.- 7

'. - 2 . Power Spike Factor Vs Axial Position - Oconee 2. Cycle 2.. ..

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'i- 1. P.0C (4 Er5'n), C>c:c 2 Two-Dimensional Relative Power I Dist r ibut is.a - Ful1 Power, Equilibrium Xenon, Normal Rod Positions (Croups 7 and 8 loserted) . . . ...... ... 5-6 -

8-1. Oconee 2, Cycle 2 - . Core Protection Safety Limits ... ....

S-2. Oconee 2. Cycle 2 - Core Protection Safety Li=its .. . ....

8-2 8-3 f

S- 1. Oconee 2. Cycle 2 - Core Protection fafet y Limits . ... ... 6-l.

s-4 Oconee 2, Cycle 2 ~ Prott et ive System itiximus Allowable SetPoints . . . . . . . . . . . . ... . .. ... ... ... 8-5 '

8-5 Os onee 2. Cycle 2 - Protect ive System .%iximun Allowable

SetPoints . . . . . . . . . . . . ... . .. ... .... .. 8-6 i

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11sures (Cont *di F-a.:e Y i .: sre 3-6. Oconee 2, Cycle 2 .- Rod Position Limits for Feur-Pump 0;,cration Fron G to 150 ( 10) EFPD . . . . . . . . . . ..... 3~

M- 7. (h:once 2. Cycle 2 - Rod Posit ton Limits. for Feur-Pu:np Operation From 150 (?10) to 261 (:10) ETP3 . . . . . . ..... 4 's 4-6 oconec 2. Cycle 2 - Rod Position Limits for Fo.ar-Pump Oper.itian After 261 (*10) EFPD . . . ... . . . .. . . ..... 5*

M-9. Oconee 2. Cycle 2 - Rod Position Limits for Two .ind TLrte-Panp Oper.ation Fron 0 to 150 (:30; EFF3 .. . . ..... 4-10 74-14 Oconec 2. Cycle 2 - Rod Position Limits for Two- and Three-Pump Oper.ation From 150 (:10) to 261 i:10) EFPD ..... 3-11 6-11. oconee 2. Cycle 2 -- Rod Posit ion L! sits for Tww. and Three-Pir.p operation After 261 (:10) EFFD .. .. . . ..... 3-12 M-12 skonee 2. Cycle 2 - Operational Power Imbalasce Envelope for Operatinn Frio O to 150 (:10) EFP3 ... .. . . ..... 8-13 h- ! 3 neenee 2. Cycle 2 - Operat fonal Pcwer Imbalance Envelope ior oper.itfen From 150 (:10) to 261 ( 10) EFPD . . . . ..... 3- 1.'.

3 - : *. .

Oconce 2. Cycle 2 - Operat ional Power Inhalance Envelope for oper.ition After 201 (?10) EFPD . . . .. ... . . .....

B-15 4

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1. INTRODUCTION AND SLW.ARY This report justifies the operation of the second cycle ef Oconee Nuclear Sta-tion, Unit 2, a t the rated core power of 2568 MWt. Included are the required analysca out!!ned in the USNRC document " Guidance for Proposed License A=end-ncnts Relating to Refueling," dated June 1975. To support cycle 2 operation of Oconee Unit 2, this report employs ana1>tical techniques and design bases established in reports that have been submitted and accepted by the USNRC
(see references).

Cycle I and 2 reactor parameters related to power capability are summarized briefly in section 5. All the accidents ar.alyzed in the FSAR have been re-vicued for cycle 2 operation. In cases where cycle 2 characteristics proved f to k conservative with respect to those analyzed for cycle 1 operation, no new ecioent analyses were performed. -

The Technicaa Spc-ificat ions have been reviewed, and the modifications re-quired for cyr,!c 2 operation are justified in this report.

Rased on the analyses perforned, which take into account the postulated ef-fects of fuea densification and the Final Acceptance Criteria for Emergency Core Cooling Syste=s, it has been concluded that Oconee 2, cycle 2, can be operated safely at the rated core level of 2568 MWt.

1-1 Babcock s.Wilcox

2. OPEitATING li1 STORY

't' nit ' 2 of the Oconee Nuclear Station achieved initial criticality on Neve=ber ll, 197 5, and power escalation co=:.enced on Dece:nber 1,1973. The 100; pewer level of 2568 L't was' reached on June 19, '97'.. A contrnl ro.1 interc h nee I

w:n. performed at 248 effective full-power days (EFPD). The design feel evele I

is scheduled for completion in April 1976 af ter 460 EFPD. De first c cle involv el no operat ing anorma t ies th.st would adverselv affect fuel periornance

.intinst the second cycle.

O... r:it inn af evele 2 is acheduled to bec,in in early June 1976 The desien cycle length . is 2'32 EFPD, and no control rod interchanges are planned.

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2-1 Babcoch & Wilco'x

" O Y P T T

I f 3. CENERAL DESCRIPTION The Geonee Unit 2 reactor core is described in detail- in Cnapter 3 of t he Unit 2 FSAR.I The cycle 2 core consists of 177 fuel assemblies, 173 of which are I

15 by 15 array containing 208 fuel rods,16 control rod guide tubes, and one incore instrument guide tube. The fuel rod cladding is cold-worked Zircaloy-4 with an OD of 0.430 inch and a wall thickness of 0.0265 inch. The fuel consists of dished-end cylindrical pellets of uranius dioxide which are 0.700 inch in length and 0.379 inch in diameter. (See Tables 4-1 and 4-2 for add t-tional data.1 The other tw.* f uel assemblies in cycle 2 are demonst rat ion 17 b, 17 :brk C fuel assembifes.2 All fuel assemblies in cycle 2 except the 17 hv 17 demonstration assenblies maintain a constant nominal fuel loading of 461.6 kg of uranium. The undensified nominal active fuel lengths ard theoretical den-( sities vary between hitches, however, and these values are given in Tables 4-1 and 4-2.

Figure 3-1 is the core loading diagram for Oconee Unit 2, cycle 2. The ini-t ial enrichments of batches 2 and 3 were 2.75 and 3.05 vt : uranium-2 3 5. re-spectively. Batch 4 is enriched to 2.64 wt I uranium-235. All the hatch I assemblies w111 be discharged at the end of cycle 1. The batch 2 and 3 as-semblics will be shuffled to new locations. The batch 4 assemblics will oc-cupy primarily the periphery of the core and eight locations in its interior.

Figure 3-2 is on eighth-core map showing the assembly burnup and enrich =ent distribution at the beginning of cycle 2.

Reactivity contrcl is supplied by 61 full-length Ag-In-Cd control rods and solubic boron shtm. In addition to the full-length control rods, eight par-tial-length axial power shaping rods (APSRs) are provided for additional con-trol of axial power distribution. The cycle 2 locations of the 69 control rods and the group designations are indicated in Figure 3-3. The core loca-tions of the total pattern (69 control rods) for cycle 2 are identical to those of the reference cycle indicated in Chapter 3 of the FSAR.I However, 3-1 Babcock s Wilcox

t !.e t;roup desi.; nations dif fer between cycle 2 and the ref erence cycle t. .ini-nize p.iwer pe.sking.  ;.ither control rod interchance n 'r r.urnable pois.n re2s

. ire n( t essary f or cyc le 2.

Tne n.>ni na l sy, tem pressure is 2200 psia , and t he core .r. er..cc densi f it d n oi-n.it hsat rate is 5.78 kW"ft at t he rated core power of 2 %o L't . This he.it rate i s sli.:ht ly hi;;her t han tisat of cycle I because of the sh>rter st.n k height of batch 4.

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4 F1'EL SYSTEM DESIGN 4.1. Fuel Assenbly Mechanical Design pertinent fuel desi.;n parameters are listed in Table ".- l . All fuel a-senblies are mechanically interchangeable. but the hrk. C denonst rat ica assen5 t ies can-not he located in a rodded location. The 5'. tew m rk s-a f_cl asseeSlie- Incor-porate ciinor modif icat iena to the end fittincs. primarily O reduce fuel assen-bly pressure drop and inercase holddown cargin. Two cf tlw % Mark S-4 reload fuel assemblies were especially constructed te exaulne the et fect on rod how

- of pow it i..n ing the fuel rods O.fi inch above the fuel asseen51v bot toe trillage.

'this repositioning of the fuel with respect to the lover ced iltting ai sets no nei hanical operating para =vt ers with the pcssible exce; t ion of rod how, which may be decre.ased. In addition, reposittening tee fuel within the two f uel asse-blles wilI not decrease fuel assenS!*r intecrity during accident con-ditlon%.

Standard Lrk n fuel rods .ere used and the fera and function of these two t'oel assemblies rermin identical to the standard brk B-4 fuel as-senhlies. The two deronstration 17 by 17 Nrk C fuel assesblics are described in reference 2.

All other results presented in the FSAR fuel a sembly necha.mical discussien are applicable to the reload fuel assenblies.

to . 2. Fuel Rod Design Pertinent f uel rod di=ensions for residual a-d =cw fuel are listed in Table 1-2.

0 The nechanical evaluatien of the fuel rM is discussed below.

1 2.1. C! adding Colla;*se e

collapse analyses were performed for three-cycle assse51y power histories teonee .l. 'able 4- 1 is a sumnary of the batch 2. 3. and a f uel red de-signs. The fuel asser.51v power histories were ar.alyzed a- d t he most limiting histeries were determined. Specific assembly rever histerics were used in the analyses of batches .' and 3.

Satch 4 was analy:ed usin.t a conservative power history envelope. Actual operatine history was used when available. This 4 Babcock s.Wik:ox e

F f ecludeal the initial pawsr operation at 49 and 60' core re.er. The predicted a<,sebiv power hist'rv fer the most lini t 11i: assenbly was u<.2 to deter-ine t'e P.ns t limitin: ellar <e time, as described in MW-1000!.i'-A.

  • iw 2000-hour dens!*^tca: Lon assumption described in reference 3 was used in the analysis since it rspresents the most severe conditien. T?.. co*.<ervatisns in the analytical procedt.re are surriarized belev.
1. fhe r.i<cV conputer code was used to predict the tire te collap*.:

CMOV cornerv.itively prcOle.s collapse times.3

. ' . No c red i t in tai.n for f.4sion n.as rc l e.t se. Therciere. the act dif fer. nt ial ;'ressures used in the a alysis are conservat ivelv hich.

1. The claddine t 5!ckness used was the lower tolerance limit t1.71. ' of the as-built reasurements. The initial ovality of the claddir:: used w.i s the upper tolerance limit (UTI.) .cf the as-built reasurenents.

The* - values were taken frem a statist!ral sampline of the (!. adding.

.. A censer.*itise power histary envelepe was used in the hatch 4 uel onalvsi .

.5. no . limit Ian .:s .nhlv was found to have a collapse t ine greater tha : the

.si: um projes t ed t hrer-evele lifetime of 25.059 effective full-pewer h.urs

  1. .. T..S I . 4-17. Tb !~ .m it ysis was perf orned usine the assuret ion i en d. nsi-at ien described la eterance 3.

.. ... Cl.iddinA Kt r. 4 s i ,c. the batett .' a nd I t'uel is the most limitinc from a cladding stress point ei si.w due to the 1.wer prepressurizat ion and Iw densit y, t he calculat tens pert erned in the Or..nce 2 Fuel Densification Reyrt, 1:AW-13%," represent ed t ue nc ,t limiting c.swe fer Oconee 2, cycle 2.

..J.1 Fuel pellet ' Irradiation Swelling The fuel design criteria speelfy a limit ef 1.0 on claddin,: circuiferential pla4tle strain. The pel!ct design is such t'ut the plastic cladding str. min is

.. .4 t han 10 at i .004 WJ /nt!'. The follawinc censervat ism, were used ir this

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1. The r:axinem recificat ion value fer the fuel pellet diameter w.as used.

J. The maxirum specificatien value for the fuel pellet den

  • l ty wa- used.
1. The cladding ID used was the lowest permitted specification tolerance.

.. The maxitum exrected three-cycle Iceal pellet burnup is less than 5 5,0M *:' 'J /m t t' .

I Babcock s,Wilcox  : .

4_2 1

4.3. Thereal _ pe>:Ign The core loadina for cycle 2 operation is shown in Figure }-:. There are 36 f resh (batch a) fuel assechlios and 121 once-burned (batches 2 and 3) assen-blies.

Two of the fresh assemblies are Mark 0,17 by 17 cesa stration fuel; these were not considered in the thermal design analysis sie:c the core con-figuration would be = ore liniting if they were not presect. he therma! de-sign of the two Kirk C deconstration assemblies is described in a separate report.2 The two Ibrk B-4 lif ted rod demonstration fuel assemblies are ther-nally identical to the cther 52 Mark B-4 assemblies and do n3t require separate analyses. The hatch a fuel has a higher initial theoretica; density (TD) and a correspondingly higher linear heat rate capability (20.15 Ys 19.8 kW/f t) than the batch 2 and 3 fuel. These linear heat rate limitations were established using the TAFY-3 code! .ith full fuel densification penaltica.

'. 3.1. Power Spike Fedel the power spike model used in this analysis is identical to that presented in PdW-10055 except for two modificatiens; the modificatiens P. ave been applied to Fgand Fk as desertbed in referen[e 7. These probabilitia have been

.: hanged to reflect addi t ional data f rom operat ing reacters t%t support a woneshat d i f t'eren t approach and yield less severe penaltie- cue to power spikes.

Fg was changed f rom 1.0 to 0.5. Fk was changed f rom a Gau.<ian to a linear dist ribut ion, yhich reflects a decreasing f requency with increasing gap size.

The maximun gap size versus axial position is shewn in Figure 4-1, and the power spike factor versus axial position is shown in Figure ;-2. The calculated power spike and gap sire were based on an initial theoretica; density of 92.52 and an enrichnent of 3.0 wt 2 uranite-235. The correspondine values for the batch 4 Mark B-4 and Mark C demonstration fuel would be smaller because of the increased density and lower enrichment of this fuel.

6.3.2. Fuel Tenperature Analvsis Thermal analysis of the fuel rods assumed in-reactor fuel de sification to 06.52 TD.

The basis for the analysis Ir. given in references ; and 6, with the following modifications:

1.

The code option for no restructuring of fuel has besm used in this analysis in accordance with the NRC's interim evaluation of TAFY.

~

4-3 Babcock & Wilcox

2. The calculated gap cenductance was reduced by 25% in accordance with the :;RC's interis evaluation of TAFY.

During cycle 2 operation the highest relative assembly power levels occur in batch 3 fuel (see Figures 3-1 and 5-1). The fuel temperature analysis for this f uel docueented in ebe Oconee 2 Fuel Densification Report" is applicabic for  ;

cycle 2 and is based on limiting BOC conditions (zero burnup), as shown in  ;

Table 4-4 Although batch a feel has a reduced active fuel length and a cor- .

respondingly higher average linear heat rate, the maxt=nm predicted centerline temperature of this fuel is lower than that of batch 3 fuel, even with the ,

same peaking factors applied. This is due to the higher initial density of j the batch 4 fuel. .

4.4 Material Design }

The chemical compatibility of all possible fuel-cladding-coolant-assembly interactions for the bstch 4 fuel assemblies is identical to that of the pres-f ent fuel.  ;

1 4.5. Onerating Experience t The Mark 3-4 assemblies and Mark C demonstration assemblies do not constitute a departure f rom past design philosophy, the adequacy of which has been demon- {

strated La the operation of six 177-fuel assembly plants.

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Table 4-1. Fuel Desirn Pa raecters Residual gey fuel assemblies fuel assemblies Batch 2 Batch 3 batch 4 Fuel assenbly type N rk B " Mark E-3 b rk 3-4 Mark-C No. of as% chlies 61 60  % 2 init!al fuel enrich, wt 7 2 3 *,g. 2.75 3.05 2.*4 2.6%

initf..I fuel densitv. 7 TD 92.5 92.5 93.3 94.0 Init tal fill gas pressure, psig (a) (a) (1) (a) fiat ch hurnup. BOC. .Wd/mt U 16.743.8 10.904.4 0 0 Ctadding collapse time. 1:FPH >25.000 $28.000(h) 23.000 125.000 liesian 1Ife. EFpli 18.048 25.0 % 21.024 2I.02a I" Proprietary.

(b) Ratch 1 is most limitint.

Table 4-2. Fuel Rod Dimensions Residual fuel. " "" #'"

batches 2, 3 ffark B

  • P. ark C demo ruel rod OD. In. O.430 0.430 0.379 ruel rod ID. in. O.377 0. 177 0.332 Fuel pellet OD. in. O.370 0.370 '0.324 ruel pellet density. 2 'D 92.5 93.5 94.0 Undensified active fuel length, in. 144.0 142.6 143.0 Type of flexible spacers Corrugated Spring Spring Solid spacer material Zrc y Zr-4 Zr-4 4-5 Babcock & Wilcox

F f'

,Tah,le 4 3. Iny,u t Su mary for Cladding Creep C311 apse Calculat1;,n_s P. ark C demo f x1 Batches 2, 3 _Satch 4 asse=M 1+*

Prlict OD (nean s.pecified), in. 0.3790 0. 3.~00 0.3240 Pellet density (nean specified). -% TD 92.5 93.5 94.0 Densified' pellet OD, in. 0.3650 0.3663 0.321' Cladding !!* (cean spec if f sd), in. 0.377 0.377 0.312 CI. adding. ovali t,s (L'TI.) , in. * *

  • t:laddins; thickness (I.TI.), in. * * *

"re; reasure (minir:um specified), psia * *

  • l'..st-Jen si fIcat inn pr, pressure (cold), psia * *
  • F:e... tor system preetsure, psia 2200 2200 2200 Sta. k height (undensified), in. 144.0 142.6 143.0 ePreprietary.

T.ible 4 '.. Fuel Thert.a t Analvsis Parameters ILit.-hes 2, 3 Batch 4 L minal ' linear heat - rate, kW/ft

~

5.77 5.80 Densifled .si;tive fuel length, in. 141.1 140.5 1.Incar heat rest e (I.llR ) to central fuel meIt, k'.'/ f t 19.9 20.15 1 hit ehdnnel factor en Ll!R 1.014 1.014 Init tal TD 7 92.5 93.5 Infti.nl fuel pellet 00, in. 0.370 0.370 avg (uel temp 9 nominal LHR, F 1335 1311 Fuel center!!ne melting temp 6 EOL Id) ,F 5060 5080 (a)At ecro burnup.

. . sm. ......e.c . _

. . . . .. . . . ee ~ .. ~.

Figure 4-1.

Maximum Cap Size Vs Axial Pc'11:on - Oconee 2 Cycle 2 4.0 TDF = 96.51

.TDI = 92.5%

e - 1.001

, 3.0 -

~2 u

3 8

n-

e. 2.0 -

O S u M' e

B M

i

, 1.0 -

g g. i a a i i 1 0 20 40 60 80 100 120 140 F

. P Axial Positlon, Inches I

=

8 M

i i

C 4

Pd

.s =4 x

s c

. m u

c.

S k.

o

.W C

. - c

=

.=.

o z r_

* .

- c  %

- x .=.

.M:

e

y. E l -

u .

O r a

u  :

E ~

3 -

s 4

.4 .e

~_~_ *:

r.

b 3

C. - C 4

c.

e unn

+ w. o s e. c.

o.

t oee n . . .

m m_.s - e a

O 6-

' f e 3 o

m a , ,, c e C C

. o. -

- c

~ ~ d J J J Jossed anidr; Jamd i

i I

I 4-8 Babcock & Wilcox l

5. NUCLEAR DESIGN 5.1. Physics Characteristics Table 5-1 compares the core physics parameters of cycles 1 aad 2. The values for both cycles were generated using PDQ07. Since the core has not yet reached an re,uilibrium cycle, dif ferences in core physics parameters are to be expected betweca the cycles.

The shorter cycle 2 will produce a smaller cycle differential burnup than that for cycle 1. The accunulated average core burnup will be higher in cycle 2 LP_an in cycle 1 because of the presence of the once-burned batch 2 and 3 fuel.

Figure 3-1 illustrates a representative relative power distribution for the L%ginnlag of the second cycle at full power with equilibriun xenon and nor=al red positions.

Ine critical boron concentrations for cycle 2 are lower in all cases than for e,s e le 1. As indicated in Table 5-2, the control rod worths are sufficient to maintata the required shutdr.n margin. However, due to changes in isotopics and the radial flux distribution, the BOC hot, full power control rod worths are ge=arally less than those for cycle 1. The cycle 2 ejected rod worths for tne same number of regulating banks inserted are lower than those in cycle 1.

It is difficult to compara values between cycles or between rod patterns since nelt ber the rod patterns from which the CRA is assumed to be ejected nor the isotopic distributions are identical. Calculated ejected rod worths and their adnerence to criteria are considered at all times in life and at all power ie.els in the develop =ent of the rod insertion limits presented in section 8.

The =ax1=as stuck rod worths for cycle 2 are less than those in cycle 1. The adequacy of thc. shutdown cargin with cycle 2 stuck rod worths is demonstrated in Table-5-2. The following conservatisms were applied for the shutdown cal-culaticas:

1. Poison material depletion allowance.
2. 107. uncertainty on net rod worth.
1. Flux redistribution penalty.

5-1 Babcock & Wilcox e

r W

t iux redist ribution was ac:cun6cd fcr since the shutdown analvsis v.' , 8 calcu-lated uwing a two-dicension al codel. The shutdown calculation at the snd of ycle 2 is anal: ze; at approxt=ately 261 EFPD. This is t.e latest tire l

( *. 10 days) in core lifu at which cl.e transient bank is nearly fully inserted.

After 261 EFPD, the transient bank vill be alcast fully witLJrawn, thus increas-ing tie availt.51e ahutdown =argin. The reference fuel cycle 4nutdown r_argin is presented in the Oconec !. 2, and 3 FSAR, Table 3-5.

The cycle 2 power Jeficits'fron hot :ero power to hot full pcwer are si=itar to but s t i;;nt i.' higher than those for cycle 1. Coppler coefficients, moderator coefficients,'and xenon worths are sinilar for the two cyclen. The differen-tial boron .orths for cycle 2 are lower than those for cycle 1 due to depletion si the fuel and the associated buildup of fission products. The effective

.iclayed neutron fractions for both cycles show a decrease with burnup.

1.2. Analvtical Innut

Le cycle 2 incore'~neasurement t alculation constants used for co=puting core tvwer illwt ributions were prepared in the sare manner as for the reference

,vele.

2.l. -Channes in Nuclg r__ Design ihere are.four deconstration fuel assemblics in b tch 4. The two 17 by 17 tark C t!cmonstration fuel assemblies have been shown to have no significant difeet on_the nuclear design of cycle 2.2 The rotential impact of raised fuel rods (0.6 inch above the lower end fitting drillage) on tae nuclear design and safety analysis of Oconee 2, cycle 2 has been reviewed. Since only two fuel assemblies with raised fuel rods are being inserted into the core as part of batch 4, the impact on the overall nuclear para eters is negligible, and no additional analysis is required. Further-more, no additional restrictions on the placement of these assemblies in the-

- core are necessary.

The same calculational methods and design information were used to oitain the inportant nuclear design parameters for cycles 1 and 2. In addition, there are no significant operational procedure changes frem the reference cycle with regard _ to axial or radial power shape control. xenon control, or tilt control.

fhe operational 11 nits (Technical Specifications changes) for the reload cycle are shown in section 8.

}

%J

-- ~. _ . A

Table 5-1. ' :enee 2 Cycle 1 and 2 Physics Paranaters Cycle 1 Cycle 2 Cycle length, EF?3 460 292

~

Cycle burnup, X.'d/rtU 14.396 9144 Average core burzup -- EOC, r.;d/=tc 14.396 18,612

=

Initial core 1cading, atU 82.1 82.1 E

Critical boron -- SOC, ppm (no Ie)

HZP(a), all rois out 1634 1350 HZP, groups 7 and 8 inserted 1494 l 1187 IIFP, groups 7 a 4 8 inserted 1382 1004 Critical boron - F,3C, ppm (eq Ie) l  !!ZP, all rods out 480 390 IIFP, group a (37.5% wd, eq Ie) 180 46 Control rod wortis -- HFP(a), BOC, tak/k

? Croup 6

' 1.58 1.17 Group 7 0.99 0.97

~ croup 8 (37.3% ud) 0.44 0.54 Control rod worth.s - EFP, EOC ilk /k Group 7 1.37 Croup 8 (37.5 1.32 l 4) 0.26 0.50 i

Max ejected rod v3rth - HZP, 24k/k BOC 0.48(C) 0.57(b)

EOC 0.72(C) 0.54(b)

Max stuck rod wor:h - HZP, Itk/k BOC 4.27 2.15 EOC 2.69 2.21 Power deficit, EZ? to HFP, Zik/k BOC i

1.32 1.64 EOC 2.10 2.48 Doppler coeff - L3C,10-5 (ik/k/*F) 100 power (O I4) -1.51 -1.49 Doppler coeff -- EDC,10-5 (ik/k/*F) 100% power (eq Ie) -1.67 -1.53 Moderator coeff - hTP,10 (ik/k/*F)

BOC (0 Xe,1000 ppa, groups 7, 8 ins) -0.23 -1.06 EOC (eq Xe, 17 7ps, group 8 ins) -2.70 -2.63 l Boron worth - HFP, pps/%Lk/k BOC (1000 ppm) 98 408

. EOC (17 pps) 95 100 p q m , . ..-.

Tabl= 5.1-1. (Cect'd)

Cycle 1 Cycle 2 Xenon worth - HFP Zak/k BOC (4 days) 2.71 2.61 e EOC (equilibrium) 2.65 .2.67 .

Effective delayed neutron fraction - H7P BOC O."069C 0.00577 EOC O.00514 0.00516 (a)HZP: hot zero power; HFP: cst full power.

I Ejected rod value for groups 5, 6, 7 and 8 inserted.

(* Ejected rod value for groups 6. 7 and 8 inserted. -

  • ?

s l

l l l

i

.l l

9 1

e l

l l

I l

9 I

l~ '

l I

m. - -.
At 1

. -1 I

Table 5-2.

l Shutdown Marif s Calculatica - Oconee 2. Cicle 2 Ava i l .i b l e M.id O rth 300, % *.k/k EOC(3). i ..k/k Total rod worth, F.2P(b) 9.70 9.75 g Wrth reduction d.ee to burnup cf -0.19 -0.30 I poison nr.aterial

!!aximum stuck rod,liZP -2.15 -2.21

, Wt worth 7.36 7.24

1. css 10% uncertainty -0.74 -0.72 Total availabic worth 6.62 6.52

' 1:ert u i red Rod W r t ..

Power deficit, !!FP to liZP 1.64 2.48

!!.ix allowable inserted rod wort!. 1.07 1.27 flux redistributica 0.40 1.00 Tot.il required worth 3.11 4.75 gliu,t,down !!arnin Total avail. worth - total req. vorth 3.51 1.77 Note: Required shutdown margin is 1.00: *k/k.

(a)For shutdown margin calculations, this is defined as approximately -'

261 EFPD, the latest time in core life at which the transient bank ,

is nearly full-in. '

(b)llZP: hot zero povor; liFP: hot full power.

mD J

l l

l I

l

[ 5-5 - Bagk & Wilcox 1

i

Figure 3-1. EOC (4 EFP3), Cycle 2 Two-;i ensionsi P.clati = Pr. e r Distri'a ution - Full Power. I:;;ilibri.n lLncr.. ::er _si Zod Po sitions (Croups 7 and i inserted) 8 9 10 11 12 13 14 15 H 1.J1 1.20 1.17 1.13 1.19 0.84 0.78 0.63 7

K I.20- 1.34 1.31 1.19 1.23 0.57 0.71 3.62 a

L

1. L 7 1.31 1.14 1.10 1.11 0.96 1.00 0.55 M

1.13 1.19 1.10 1.33 1,29 1.26 0.96 l

l s

N 1.19 1.23 1.11 1.29 1 10 1.12 0.69 7

0 0.ta 0.57 0.96 1.26 1.12 0.75 l

P O.7a 0.71 1.00 0.96 0.69 p

0.o3 0.62 0.55 l

l l

l I x Inserted Rod Group No.

, x.xx Kelative Power Density 1

5-6 Babcock & Wilcox

6. TliERMAL-liTDRAULIC DESlaN 6.1. Thermal-Hydraulic Design Calculations Thermal-hydraulic design calculations for support ef cycle 2 operation utili:ed the analytical cethods documented in references 1 And 4. These calculations were made to account f or the introduction of the Mark B-4 asse=blies in batch 4 to consider the minimum actual reactor coolant system flew rate as ceasured during first cycle operation, and to incorporate the B&W-2 CHF correlatien in place of the previously used W-3 correlation. The 54V-2 CHF carrelation was used in licensing oconee 1 cycle 2.

6.1.1. Introduct ion of Mark B-4 Assemblies As discussed in section 4.1. the Mark B-4 asse=blics differ from the Mark B-1 assemblies ,primarily in the end fittings. This difference causes a slight re-duction in the flev resistance of the B-4 assemblics. Since the B-4 assemblies are loaded primarily on the periphery of the core, the hottest (highest power) assembly is a E-3 (see Figures 3-1 and 5-1). In erder to conservatively account for the introduction of the two Mark C assemblies. the thermal-hydraulic model utilized a 56 B-4 121 B-3 configuration and retaieed the B-3 assemblies in the hottest core locations. This assumption increases the conservatism of the cycle 2 design by reducing the calculated hot assembly flow rate. For the two Mark B-4 lif ted-rod demonstration assemblies, the Assembly forn loss will be less enan the Mark B-3 fuel. Hence, the B-3 assesbly will still have the highest hydraulic losses and will be DSB-flow-limited.

6.1.2. Increased RC System Flow

~ Reactor coolant flow data obtained during cycle 1 eperation verified that the system flow w.is creater than the destr.a flow.

The nessured value was 111.5% l of the design flow. For the evele 2 thernal-hydra lic design analysis, the increase in systen flow was conservatively choAen to be 107.6% of design.

6-1 Babcock s.Wilcox 1

1

l l

6.1.3. B&W-2 DNB Correlation The B&W-2 DNB correlation, a realistic prediction of the burnout phenonena,Y+9 has been reviewed and approved for use with the Mark 3 fuel assembly design.

In the application of this correlation to the Oconee 2 crcle 2 core, two modi-fications which have also been applied to the TMI-1, cycle 2 and Oconee 1, cycle 3 cores, have been Instituted:

1. The limiting design DNBR of 1.30 was used, representi2g a 95% confidence level for 951 population protection. The limiting DN5R of 1.32, which had been used for this correlation in previous design analyses, repre-sented a 997. cenfidence level for a 95 protection. This change is con-sistent with industry practice and the statistical standards associated with liniting design DNBR values accepted by the NRC Staff and the ACRS.
2. The pressure range applicable to the correlation has been extended down-ward from 2000 to 1750 psia. This revision is based on a review of rod bundle CilF data taken at pressures below 2000 psia, which shows that the RAW-2 correlat ion conservatively predicts the data in this range.

The use of this correlation in conjunction with increased system flow for the cycle 2 analysis indicates that the margin to DNS is higher than had been predicted f or first core operation, as shown in Table 6-1.

6.2. DNBR Analysis In addition to the items discussed above, the maximum design conditions considered in the FSARI and generic fuel assembly geo etry beed on Mark 3 as-built data were taken into account. This resulted in a minimum DNBR of 1.98 at 112% power for undensified fuel.

The effects of densification can be divided into two categories: (1) the re-duced stack height and (2) the power spike resulting from Jensification-induced gaps in the fuel column . The active length was calculated to be 141.1 inches, a reduction from the undensified length of 114.0 inches. These densified and undensified lengths are based on fuel batches 2 and 3, the lim-iting batchei in cycle 2, as discussed in section 4.3.1.

The axial flux shape that produced the maximum change in DN3R from the original design value was an outlet peak with a core offset of 411.8%. The spike magni-tude and the maximum gap size are discussed in section 4.3, and the values .

6-2 Babcock & Wilcox l

l I

  • me

W used.in the' analysis are 1.087 and 3.10 inches, respectively. The results of the two ef f ects are -1.88 and -1.061 change

  • in the mini =u= het channel DNBR

.rnd peaking sargin, respectively. The changes in these cargins are su==arized

-in Table 6-1. vnich includes comparisons of other pertinent cycle 1 and 2 data.

the DNBR analysis has been based on a core configuration consisting of 17; Mark B fuel assemblies. The incorporation of two Mark C demonstration fuel '

assemblies in hatch 4 in place of two .'brk B assemblies. recuits in .an increase in the overall targin to DNB. as discussed in reftrence 2.

d a

Change t rom undensified values.

cycle I densified conditions. These are actually an innro ement over

,, n.o.......-

(n

Table 6-1. Cvele I and 2 Maxinum Design Conditions C yele il Cycle 2 Design power level. W: 2563 2568 Systen pressure, psia 2200 2200 Reactor coolant flow. I design flow- 100.0 107.6 Tessel inlet coolant temperaturg 2002 power. F 554.0 555.9 Tcssel outlet coote: :emperature.

100 power. F 603.8 602.~4 E.cf. design radial-local power peaking factor 1.78 1.78 Ecf. de sign axial flux shape 1.5 ce ine 1.5 cosine Active fuel length. in.

144 (undens.) 141.1 (dens.)

Avg heat flux. 1002 powr, Etu/h-fe-171.474 174.995 w

ax heat flux. 100% powr, I.t u /h- f t - (fo. DNER cale) 457.S25 467.236

-OlF correlt ,lon W-3 B&W-2 winimum DNBR (Enx design conditions. 1.55 1.98 no densif penalties) (114* power) (112,', power) hit channel factors Enthalpy rise 1.011 Ilea t flux 1.011 1.014 1.014 Flow area 0.98 0.98 Densification effects Change in DNBR nargin. 2 -6.03- -1.88 Change'in power peaking margin. 2 -2.82* -1.06 g Babcock s.Wilcox -

7 ACCIDENT AND TRANSIENT ANA1.YSIS 7.1. General Safety Analvsis Each FSARI accident analysis has been exanised with respect to changes in cycle

.!. parameters to deter =ine the ef fects of the Ocle 2 reload and to ensure that tuerr.11 performance is not degraded during hypothetical transients.

The core thermal para =eters used in the FSAR accident atlysis were design operating values based or. calculated vilues plus uncertainties. Cycle 1 values (FSAR values) of core taermal parameters are compared with those used in the cycle 2 analysis in Table 6-1.

These parameters are comsm>n to all of the acci-dent analyses presented herein. For each accident of the FSAR, a discussion and the key pareneters are provided. A comparison of the key p 'ameters (s-e Table 7-1) from the FSAR and the presi t cycle 2 is provided with the accident discussion to show that the initial conuitio=s of the transient are bounded by- the FSAR analysis.

The ef fects of fuel densification on the FSAR accident results have been eval-uated and are reported in RAV-1395." Since cycle

  • reload fuel asse blics 1 contain fuel rods whose taeoretical density is higher than those considered in ref erence 4, the conclusiou derived in that reference are still valid.

Calculational tecnniques and methods for cycle 2 analyses remain consistent with those used for the FSAR. Additional *:NIR margin is shown for cycle 2 because' the B&W-2 CitF correlation was used instead of the W-3.

No new dose calculations were performed for the reload report. The dose con-siderations in the FSAR were based on maxiar.m peaking and burnup for all core cycles; .therefore, the dose considerations are independent of the reload batch.

7.2. Rod Withdrawal Accidents This accident is defined as an uncontrolled reactivity addition to the core i

due to withdrawal of control rods during startup conditicas or from rated I

-power conditions.

Both types of incidents were analyzed in the FSAR.I 1

l 1

7-1 Babcock 4.Wilcox -

l

The !=portant par neters during a rod withdrawal a:cident are Doppler coeffi-cicat. moderator.te=perature coefficient, and the rate at whica reactivity is added to :: c core. Only high pressure and high-flu trips are accounted for in t: . F -Je .inalysis, Qic:a ignores =ultiple alar: 3 I nt e r lock.s . and trips t ..at a~rn lly preclude this type of incident. - For positive react ivity addi-tiaas indicative of these events, the nest severe results occur for BOL cen-litions. The FSAP. values of the key parameters for BOL conditions were -1.17

- 10~L (Ik/k/*F) for the Doppler coefficient. 0.5 4 2k/k for the codera-tse temperature coefficient and rod grcup worths up to and including a 100

?k/k rod bank worth. Co= parable cycle 2 parametric values are -1.49 - !O'S l (Jk/k/*F) for the Doppler coefficient, -1.06 a 10~~ (ik/k/* F1 f or the modera-sor temperature coefficient, and a max

  • rm rod bank worth of 9.7% Ak/k.

Increfore, cycle 2 parameters are bounded by desig= values assumed for ,^he FFAK analysis. 'Thu4, for the rod withdrawal transients, the censequences will be no nore severe than those presented in the LM. For the rod withdrawal from rated power, the transient consequences are also less severe than those presenteil in the duasification report. '

I J.J. Moderator Dilution Accident doroa in the form of boric acid is utiliced to control excess reactivity.

Ine boron content of the reactor coolant is periodically reduced to compensate tor-Tuel burnup and transient xenon effects with dilution water supplied by the nakeup and purification system. The moderator e'J.ution transients consi-Jered are the pucping of water with zero boron concentration from the makeup tank to the RCS under conditions of full power operation, hot shutdown, and refueling. The key parameters in this a=alysis are the init!al boron concen-tration, boron reactivity worth, and coderator te= erature coef ficient for power cases.

For pasitive reactivity addition of this type, the most severe results occur for DOL conditions. The FSAR values of the key para eters for SOL conditions were 1400 pps for the init ial boron concestration, 75 ppm /l: (Ak/k) boron reactivity worth and +0.94 10~" Ak/k/ *F for the coderator teeperature coef-ficient.

Ceaparable cycle 2 values are 1004 pps for the initial boron concentration, s3 pps/12 (ak/k) boron reactivity worth and -1.06 10~" (ak/k)/*F for the i

7-2. Babcock s.Wilcox

.,]

=uderator te=pera ure coefficient. The FSAR shows that the ccre and RC5 are adequately protected during this event. Sufficient ti=e fer crerator action to turninate this transient is also shown in the FSA2. e.e= wita naxina= dilu-tion and minimun shutdown margin. The predicted cycle 2 pararetric values cf importance to the =oder tor dilution transient are housded br the FSAR design values; thus, the analysis in the FSAR is valid.

7.4 Cold Water iPurp Startup) Accident There are no check cr isolation valves in the reactor coolant piping; there-fore, the classic cold water accident is not possible. Er.ever, when the reactor is operated with one or more pu=ps not running, and then these are turned en. the increased flow rate will cause the average cere terperature to decrease. If the =oderator te=perature coef ficient is nerative, then reacti-vity will be added to the core and a power rise will occur.

Protective interlocks and procedures prevent starting idle pu=ps if the reactor power is above 22!.

However, these restrictions were ignored, and two punp startup f rom 50; power was analyzed as the most severe transient.

To maxinize reactivity addition, the FSAR analysis assumed the most negative nederator temperature coefficient of -3.0 - 10~" (*k/i)/*F and least necative Deppier coefficient of -1.30 > 10-1 ak/k. The correspondine most negative moderator temperature coefficient and least negative Doppler coef ficient pre-dis ted f or cycle 2 are -2.63 10~" and -1.49 = 10-5 (li/i)/*F, respectively.

Sluce the predicted cycle 2 moderator te=perature coef ficie:t is less negative and the Doppler coefficient is mor e negative than the valt.es used in t he FSAR, t he t ransient results would be less severe than those reperced in the F<Ah.

i.J . Lon of Coolant Flow The reacter coolant flew rate decreases if one or core of the reactor coolant pumps fail.

A pumping f ailure can be caused by mechanical f ailures or a loss of electrical power. With four independent pumps available, a rechanital fail-ure in one pump will not affect the operation of others. Vith the reactor at power, the effect of loss of coolant flow is a rapid increase in coolant ten-perature due to the reduction of heat removal capability. I'nis increase enuld result in DSB if corrective action were not taken i==ediately. The key para-meters for four-pu=p coastdown or a locked-rotor incident are the flow rate, flow coastdown characteristics, Doppler coefficient. =oderater teeperature 7 Babcock s. Wilcox

coefficient, and hot casanel DNs peaking f actors. The most censervative initial condit ions were assumed f or the de=sif ication report:- FSAR values of flew and coastdown. -1.17 10~i (ik/k)/*F Ooppler coefficient. +0.5 19 (lk/k)/*F -

=oderator te=perature coefficient, with densified fuel power spike and peaking.

The results showed that tne DNBR re=ained above 1.3 ('4-3) f or t re f cur p u=p ceastdown. and the fuel cladding tc=perature re=ained below criteria li=its for the locked-rotor transient.

Iae predicted parcsetric values for cycle 2 are -1.49 10~5 (li/k)/'F Deppler euefficient. -1.06 10~' (;k/k)/*F noderator te=perature coefficient, and peaking f actors as shown in Table 6-1. Since the predicted cycle 2 values are bounded by those used in ene densification report. the rescits of that analysis represent the most sever = consequences from a loss of flow incident.

7.n. Stuck-cat , Stuck-In. or Drop <d Control Rod it a control rod were dropped into the core while it was operating, a rapid decrease in neutron power would cccar, accompanied by a decrease in the core average coolant temperature. The power distribution sight be distorted due

- to a new control rod pattern, under which conditions a return to f ull power night lead to localiced power densities and heat fluxes in excess of desica 4

IInitations.

i !.e key para =eters for ti.is transient are moderator te=perature coefficient, erepped rod worth, and local peaki=g factors. The FSAR analysis was based en u.40 a:id u.Jo?. ik/k rod worths with a noderator tenperature coef ficient of

- 3. 0 * (;L/k)/'F. For cycle 2. the maximus worth rod at pewer is 0.20'.

h/k and a cederator temperature coefficient of -2.63 - 10~- (Li/k)/*F. Since

:.e predicted red worth is less positive and the soderator tenperature coef f i-e ie.s t is more positive, the consequences of this transient are less severe than the results presented in the FSAR.

i

7. 7. I.oss of Electric Pewer '

T-o types of power losses were considered in the FSAR: (1) a less of load condition caused by separation of the unit f rom the transmissics systen and (2) a hypothetical condition resulting in a co=plete loss of all systes and l unit power except that from the unit bat terie s.

Inc FSAR analysin eVJ1uated the loss of load with and without turbine runback.

L~asn there is no runbeck, a reactor trip occucs on high reactor coolant pressure y_4 Babcock s.Wilcox i

l l

l or tenperature. This case results in a non-li=iting accident. The largest oifsite dose occurs for the seccad case, i.e., loss of all electrical ; cver except unit batteries, and assu=ing operation with failed fuel and stes= gene-rator tube leakage. These results are independent of core Icading; therefore, l the results of the FSAR are applicable for any reload.

7.8. Steam Line Failure A steas line f ailure is defined as a rupture of any of the steam lir.es f rom t.'io steam generators. L*pon initiation of the rupture, both steam generators ll start to blow down, causing a sudden decrease in the pri=ary syste= tecpera-ture, pressure, and pressurizer level. The temperature reduction Icads ta positive reactivity insertion, and the reactor trips on high flux or lov AC pressure.

The FSAR has identified a double-ended rupture of the steam line between the steam generator and steam stop valve as the worst-case situation, at end of-life conditions.

The key parameter for ther core responce is the moderator t mperature coeffi-cient, which was assumed in the FSAR to be -3.0 = 10~'* (.*k/k)/*F. The cycle 2 predicted value of moderator te=perature coefficient is -2.63 10~' (2k/k)/

  • F. This value is bounded by those used in the FSAR analysis; nonce, the re-sults in the FSAR represent the varst situation.

7.9. Steam Generator Tube Failure A rupture or Icak in a steam generator tube allows reactor coolant and asso-ciated activity to pass to the secondary system. The FSAR analysis is based on conplete severance of a steam generator tube. The pri :ary concern f or this incident is the potential radiological release, which is independent of cere loading. Ifence, the FSAR results are applicable to tnis reload.

7.10. Fuel llandling Accident The mechanical damage accident is considered the =aximum potential source of activity release during fuel t adling activities. The primary concern is radiological releases that are independent of core loading; therefore, the FSAR results are applicable to all reloads.

7.11. Rod Ejection Accident For reactivity to br added to the core more rapidly than by uncontrolled rod I

withdrawal, physicai tailure of a pressure barrier conponent in the centrol 7-5 Babcock 8.Wilcox

rod drive assembly =ust occur. Sach a failure could cause a pressure differen-tial to act on a control rod asse=bly c:.J rapidly eject the assembly fro = the core region. This incident represents the most rapid reactivity insertion that can be reasonably postulated. The values used in the FSAR and densification report at BOL <saditicas. -1.17 4 10-5 (Ak/k)/*F Doppler coef ficient, +0.5 =

10-" (7.k/k)/*F coderator tecperature coefficient, and an ejected rod worth of 0.657. J.k/k, represent the =aximus poss; ale transient. The corresponding cycle 2 parametric values of -1.49 - 10~5 (ik/k)/*F Doppler, -1.06 10-4 (l.k/k)/*F coderator te=perature coefficient (both more negative than those used in refer-ence 4), and a =axinus predicted ejected rod worth of 0.13: ik/k ensure that the results will be less severe than those presented in the FSARI and the den-l sificatica report.'

J_.J_2 . "axi=um 1:ypothetical Accident The re is no postulated =echanism whereby this accident can occur since it would ecquire a multitude of failures in the engineered safeguards. The hypothetical accident is based solely on a gross release of radioactivity to the reactor bu i ld ing. The consequences of this accident are independent of core loading; hence, the results reported in the FSAR are applicable for all reloads.

7.13. Waste Cas Tank Rupture The vaste gas tank was assu=ed to contain the gaseous activity evolved from Jeg.nsin.; all of the rc.setor coolant following operation with II defective fuel.

Rupture of the tank would result in the release of its radioactive contents to the plant ventilation system and to the atmosphere through the unit vent.

The consequences of this incident are independent of core loading; therefore, the results reported in the FSAR are applicable to any reload. ~

7.14. LCC Analvsis A generic LOCA analysis for B&W's 177-FA, lowered-loop NSS (category I plant)  !

has been perfor=ed using the Final Acceptance Criteria ECCS Evaluation Model. IO l That analysis is generic since limiting values of key parameters for all plants i An this category were used. The average fuel temperature as a function of linear heat rate and lifetime pin pressure data used in the RAW-1010310 LOCA limits analysis are conservative compared to those calculated for this reload. 1 i

Therefore, the analysis and the LOCA limit.s reported in EAV-10103 provide conservative results for the operation of the Oconec 2, cycle 2 fuel.

7_g Babcock & VVilcox

{

l a

i l

f

~he tabulatica below shows the bounding values for allowable LOCA peak linear heat rates for Oconee 2, cycle 2 fuel.

Core Allowable peak elevation, linear heat ft rate. kW/ft 2 15.5 4 16.6 6 18.0 8 17.0 10 16.0 The Mark C 17 by 17 deconstra ion assembly wil" be located on the periphery of the core. Because of its location, the maxinun linear heat rate within the assembly will be approximately 10 kW/f t. Oper. tion at this low linear heat rate will prevent cladding rupture during blowdown should a LOCA occur. Since rupture during blowdown causes t'.e highest peak cladding temperature, the con-sequences of the LOCA should be less severe for the Mark C 17 by 17 demonstra-tion fuel. In addition, the low linear heat rate provides substantial margin relative to the LOCA 11aits calculated in EAW-10103.ID Therefore, compliance with the acceptance criteria of 10 CFR 50.46 is ensured.

i l

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7-7 Babcock & Wilcox i

w

4 8

Table 7-1. Comparison of Key Parameters for Acciden: Analysis

, FSAR, densif Predicted Para eter value cycle 2 value BOL Dorpler coeff,10-5 (ak/k)/*F -1.ly(a) _g,49 E01. Doppler coef f,10-h (ik/k)/*F -1.33 -1.53

' BOL mcderator coeff,10-6 (ak/k)/*F +0.5(b) -1.06 EOL' coderator coef f ,10-' (ak/k)/*F -3.0 -2.63 All rod bank worth (EZP). I ak/k 10.0 9.7 Initial boron conc (EFP), ppm 1400 1004 Boron reactivity worth (70F), 75 83 pps/1; 2k/k "ax ejected rod worth (EFP), % ak/k O.65 0.18 Dropped rod worth (ETP), I ak/k O.46 0.20 I" (-1.2 10-5 Ek/k/F) was used for steam line failure analysis.

(-1.3 105 f.k/k/F) was used for cold water analysis.

(h)(+0.94 ' ak/k/F) was used for the acderator dilution accident.

i I

7-8 Babcock 4.Wilcox

f ..

A P-J 8. PROPoiED MODIFICATION' TO TECHNICAL SPECIFICATIONS ,,.

k-The Technical Specifications have been revised for cycle 2 operation. Changes

. were the results of :be following:

.l

1. Us'ag the EL'i-2 CMF correlacion rather than 'n'-3, as discussed in section 6.1.

I- 2. Using a 95/95 confidence level rather than 99/95, as discussed in section 6.1.

3. Using 107.6% of design flow rather than 100%, as discussed in section 6.1. i
4. Using the Final .tecoptance Criteria LOCA Analysis for restricting pea'.us during operation, as discussed in section 7.14.  !
5. Rev isins; the a.,sc=:pt ions on which the flux flow RPS setpoint is based. -This setroint now acccmitr for signal noise on the basis of data accu _ulated fron operating EE. . reactors.

!!ased on the Technical ";pecifications derived f ren the analyn.. presented in this repcrt, the Final .Weeptance Criteria ECCS licits vill n, t he exca dsd, nor will the ther=.nl design criteria be violated. Figures h-1 through S-10

~

J illustrate ruvisions to previous Technical Specification safety limits.

t m

., M . L . . 0. e t e.**

_____.__i___.___.___._ _ _ _ -

i- - - - -

Tigure 8-1. Oconee 2. Cycle 2 -- Core Protectics Safety Linits 2400 _

l i

l t

1 2300 .

l ACCEPTABLE

~

OPERATION 2200 .

E J

=

i G r 2l00 -

2 y UNACCEPTf3LE

= OPERATION 2000 -

3 1900 -

I 1800 . l 1

s I t e 560 580 600 620 S40 E60 l'

Reactor Outlet Teugerature, F

. _ a.w6 . was...

s" Fi.;u re 8-2.

Oconee 2. Cycle 2 - Core Protection S.afet, Licits Thermal Pneer Level, s 128 I****'#I

(-35.a.es:1 LTNACCEPTABLE OPERATIOIf 333 KW TT I

EN TT LisIT ACCEPTABLE Litli

. leg 4 Puer OPERATION

(*so.so )

(-3s.s.es.s) '

90 g.3,,,,,,,,,

ACCEPTABLE 80 3 & A Puer OPERATl04 t-*8 55 'I l ( - t s . e . s s . s _I . gg g .3 , , , , , , , ,y l ACCEPTABLE 2,3, & 4 PUNP OPERAT104 g.so,3e.,3 , & (8 35 S) 30 20 10 50 A0 .n . 20 .a0 40 Reactor Paeer labalance, s CURVE REACTCE COOLANT FLOW (LB/HR)

I 141.3 a 108 2 It5.E : 106 3 69.3 106 4 64.7 x 106 8-3 Babcock a,Wilcox

~.

Fir,ure 6-3. Oconee 2, Cycle 2 - Core Protection Cafety L1=its 2400 -

1 2 3

[

l 2300 -

m 2200 -

ACCEPTABLE

[ 0,tRATION 2

5

$ lit' e m

I 5

5 2000 _

O UNACCEPTABLE OPERATION i

1900 -

1800 -

560 580 600 620 640 660 Reacter Outlet Tempersture. F REACTOR COOLANT FLOT PUMPS OPERATING Ct:RVE (LBS/HR) POWER (TTPE OF LlulT)

I 141.3 x 106 (1004)* 112*, FOUR PUNP (DNBR LlulTED)

) 2 105.6 x 106 (74.7%) 86.45 THREE PUuP (OhBR LIEITED) 3 69.3 x 106 (495) 58.9% ONE PUMP IN EACH LOOP (OVALITT LlulTED)

  • 107.6*, OF CTCLE 1 DESIGN FLCf s_ f, Babcock a.Wilcox

I l

l i

1 l

Figure 8 '+ . Oconee 2. L9 1e 2 - Prettetive Systen 1 i

MaximumAllshableSetpoint-  !

2400  ; '

s 4 l P = 2355 psig 3 T = 619'F N,

2300 - l 4

5 l

?.

. 2200 -

l

-2 S*  :

5 4

$ 2l00 -

ACCEPTABLE l

I OPERATION <

~

o i 0 .'

g 2000 - # @@

g .

E 8

{

D UNACCEPTABLE M00 -

T OPERATION P = 1800 psig 1800 (587.5)

I E 9 e 540 560 580 600 620 640 i Reactor Outlet Temperature. F d

1 33 Babcock a.Wilcox

7 rigure 6-5. Oconee 2, cycle 2 - Fracec:ive Syste=

M.aximum Allevable Setpo nts Poser Level. S

" 120 UMACCEPTABLE OPERATION

[

110 ,,o73

/*

A' " 100 ##

ACCEPTABLE

+

p 4 PUNP

(

+

" 90 OPERATION i

ED I79 3) o 70 ACCEPTABLE 3 & 4 PUNP CPERATION

,, E0 l (s2.a1 50 l

ACCEF T ABLE

- 40 2.3 & 4 PUNP OPERATIDN 30 20

. M. _. in .

3 Le se il 19 R

E E R E I E "

-60 -4D -20 0 20 40 60 Poser Isoalance. ',

  • THE FLUI FLCW TRIP SETPOINT FOR 2 0 PUNP CPERATION P:JST BE SET AT 0.961 g_5 Babcock & Wilcox

Figure 6-6.

Oconee 2 Cycle 2 - Rod Position Limits for Four-Pump operation From 0 to 150 (t10) EFPD

  • 8se.9.e02 .e 200.S.sa2 tQQ .

I so. 9.92 200.5. er

- POWER L(VEL S3 RESTRICTED CUTOFF RE G10eg 20s. s. es On -

30 - Ri$fRICfE0 '

s2s.s.es ,

alGIDs

  • 5 60 -

PERulS$18tE 3 50 _

OPERATING REGION MO ***'

, 40 .

nr. s. ss. s 5

2 30 _

10 - .  !

10 0 ! t i e , , , , . , . ,- ,

0 20 40 60 80 100 120 f40 150 180 200 220 240 260 280 300 Roa snees. t setneraen .

~

0 25 50 75 100 )

0 25 50 " 75 t00

. . , , . I Croup 5 G*sup 7

  • O 25 50 I 75 108

. 1 f R I Groua 6

_ 1 l l l

37 Babcock s.Wilcox

! l i

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l Figure 8-7.

Octsnee 2, Cycle 2 - Rod Position Limits for Four-Oper.ition Fron 150 (t10) to 26: (110) EF?~)

ess.sor see.s.so 2aa.7.iet 103

- OPERat10N 1% TMis 1 REGION is 43r ALL0eED ses.n.92 90 - 2 e . r. ,r O t8 tittL C O CF5 63

- SkUID0eN wanGli

,,,,,,,g yg L illli Ri$IRlCIfD RESl04 pt1TRICIf3

  • 81G104 EO -

3 C

SC s r . *,

  • PERWI 18 E

~ Golan11mG 2 2 2 n. ~'

t!GION 30 -

10 e.e, 10 -

0 ' ', e C 10 40 ED 80 100 120 140 160 183 2G2 220 **I 250 223 300 Res fr:es. EstMrson 0 25 50 IS

. . 120 8 25

. 58 75 100

.. f Gesap $ ,

pmy l

0 25 58 9

75 123 e f f f Gross 5

.. ..e

i

/

l i

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Figure 8-8. Oconee 2, Cycle 2 - Rod Position Limits for Four-Pump operation After 261 (210) EFPD 1

ees.se2 2es.2.se2 189 -OPERATICN IN INIS REGl001 15 4T Att0st0 90 E Ul -

2 n.s.92 CUT 0FF 80 tl SMuiO9eN 70 -

gagggy tiesi RI5it CIED RIGlost 5 SC SC 222.s.es so...

  1. 4C h

a le PERel5510tt 20 CPEasiness c.s REGIGN 10 _

0 - ,

0 20 40 60 80 100 120 140 180 180 200 220 240 260 280 333 Ros ingen. 1 Wetadfase 0 25 50 15 100 0 25 50 75 ICS I a e I a a a a 9 .

Grc6p 5 Group 2 0 25 50 75 100

, . I I E Creup E g_9 Babcock s.Wilcox

I 1

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l Figure 8-9. Oconee 2, Cycle 2 - Rod Position 1.imits for Two- and Three-Pump Operation From 0 to 150 D 10) EFPD asi.so2 lus.co2 l

_. RESTRit?ED PESICg 2ec.sc2 2:s. 02 g 100 - FOR 2 a%? 3 o;,p  ! RESTRICTED l  % OPERA' '% REGION FOR 2 & 3 E 90 *0 90 PUWP OPERATION

-Q  %

a 80 **' **

p

$ 222. e. soa.s i 10 p RESTRICTED REGION FOR 3 u  %@g. j a q DUWP OPERATION <

60 Y
a. % 222.% #00.St 0

fd q PERNISSIBLE

.i 4g CPERATING

., REGION 30

., 20 _

e

$' 10 -

U  ! -

0  ?? 40 60 90 100 120 140 160 180 200 220 240 260 280 33D Red index. K Betnaraen 0 25 50 75 103 0 25 $3 75 .00 m i a f a a l 9 l 3 d' Opp % Gr:np 7 0 25 50 75 100 Grcup 6 l

6

! l 8-10 Babcock s.Wilcox [

l

7.

l

}.

F i'.;u r t- R-10.

Ocena e 2 Cycle 2 - Rod Position Limit s for Two- and Thre.--Pump Operation Fro:n 150 (:10) to 261 (-10) EFPD

  • OPERATION IN Tett$ it?.s02 es0.102 280.002 289.102 ICO REGION IS %DT i g at(Ce!D 27.ss RESTRICTED REGION g F3R 2 & 3 PUNP Q 90 -

CPERATION N. SC S 222.e7 v

[ 80 ~ 5"U10084 NARGIN 9ESTRICTED RECrcs 300. .*

f I' I

[- FOR 3 PutP 0*ER4Tl0N y s '- -

q

~

g$

.y p-

'i- -  %.'2

  • O 300.6

~

. PERMISSIBLE O OPERATING 40 s

(( REGION i 9

, 30 "

20

. . e-

. 't e i 1 . , . , , , ,

t 20 40 60 EJ 100 120 140 160 180 200 220 240 260 200 300 Ro3 Indes. ) Retndraen 0 25 50 75 I00 0 25 50 75 100 Gooup 3 Grcup 7 0 25 50 75 100 G cup 6 a.ii Babcock s.Wilcox

Figure 8-11. Oconee 2, Cycle 2 - Rod Position Li=it.< for Two-and Three-Pu p Operation After 261 i-10) EFPD le3.102 22a.802 232.s02 100 _ CFfRATION 1% THIS PEGION f

[

h 15 ADT ALLCoiD RESTRICTED FCR 2 & 3

_ 90 - PUwP OPERATIC 8 j a 11. s) 222.ss F B0 G RESTRICIED REGION FOR

.- ;G 3 PUMP OPERA!'ON e.'

= bC .: , 2:2. 0 sgi vau;is 1:v:1 o.

PERul55 TELE

.

  • SPERAltNG 3e REGION

,, 20

  • e .RE5!RICliD FOR h

Il 2 & 3 Fl#

a

.' a if r~

80 103 120 !40 160 :80 200 270 ::# 76C 190 230 99n teace, y sitnaraon t

25 5 'J h a ice 0 25 29 15 'C0

. a a e a E a 4fOuJ b Gi t t: ?

C 25 50 75 100 k R R a R G0i.p E i

}

a _ , ., Babcock a.Wilcox I

Figure 8-12. oconee 2, Cycle 2 - Operational Power I: balance Envelope for Operation From 0 to 150 (:10) EFPD Power. ', of 2568 MWt g RESTRICTED ,

REGION

-tI. 2.102 . 14.28,102 18.04.92 12.88.92 90 4

-16.58.65 15.73.e5

-11.25.77 80 19.25.77 70 Pf9415$18LE 0*fe4 Ting 60

  1. f Glost 50

-20.sa.us.5 40 L 30 f

f 20 1

IO

$w I e

50 40 20 -10 0 10 20 30 20 50 Art.H Pceer '.rDalance. .

\

e Eggh

Figure 8-13. Oconee 2, Cycle 2 -- Operational Power I= balance Envelope for Operation From 150 (210) to 261 (210) EFFD Foser.

  • of 2568 NWt RESTRICTED g REGION

-16.32.10.* , g gg 14.79.802

-15.64.92 90 I 13. 3 s. s2

-85.75.87.5 80 18.25.73 70 60 PERM l$$1bLE OPEPAilNG

-28.05.47 50

  • REGICm 40 30 20 10 i R a a R m e

-50 40 30 20 -10 0 10 20 30 40 3r l Axial Power leoalance. , -

. . . Rahreir a Wilen, _ _ _

figure S- l ' . Oconee 2, Cycle 2 - Opera:fonal Power Imbalance Envelope for Operation Af ter 261 (:10) EFPD Power. A of 2568 unt RESTRICTED l REGION

-se.sr.so2 f 100 1 83.2s.sc2

-1s.12.92 .

90 15.25.92 80 70 Pf#wi$$3$Lf 86.06.Eb ounariss , gg stssc=

50

.a.c..-

40 30

! . 20 l l t 10 f

t_ . . . ___,. u .  ! . . , ,

50 40 20 10 0 10 20 33 40 53 asia! Poses letalance. 5

9. STARTUP PROGRAM The planned startup ter,ts associated with core perfocmance are described below.

, These. tests verify that core performance is within the assumptions of the safety analysis and provide the necessary data for continued plant o;< ration.

Precritical Tests

1. Control Rod Drive Trip Time Testing -

Zero Power Tests

1. Cirtical Boron Concentrat. 4
2. l Temperature Reactivity Coefficient
3. Control Rod Group Worth 4 Ejected Rod Worth Power Tests 1.

Core Power Distribution verification at Approximately 40. 75, a=d 100% FP.

. Normal Control Rod Group Configuration

2. Core Power Distribution Verification at Approximately 40% FP With k'orst-Case Dropped Rod Fully Inserted 3.

Incorn/Out-of-Core Detector Imbalance Correlation Verification at Approxi-cacely 75% FP 4

Power Doppler Reactivity Coefficient at Approximately 100 FP

5. Temperature Reactivity Coefficient at Approximately 100% FP

B-9 V_

i I

l REFEUl;CES Oconea Nuclear Station, Units 1, 2 and 3 Final Safety Analysis Repert.

Is$cketNos. 50-269, -270, and -287.

treadiation e of Two 17 - 17 beconstration Assenblies in Oconee 2. Cycle 2 -

Rdload Report, BAW-1424. Babcock & Wilcox, Lynchburg, Virginia, January 1975.

A. F. J. Eckert, 11. W. Wilson, and K. E. Yoon. Progran to Determine Inteac-L'? r rerformance of 35W Fuels - Cladding Creep Collapse. BAW-100 sap-A, Bab-c.pk & Wilcox, Lynchburg, Virginia (1975).

Uiyaee2FuelDensiilcationReport. BAW-1395, Babcock & Wilcox I.ynchburg, V{rglaia, June 1973.

s s.

C$

D. Morgan and II. S. Kao, TAFY - Fr'l Pin Temperature and Gas Pressure Analy-ts, liAW-10044, Babcock & Wilcox, Lynchburg, Virginia ".ay 1972.

Fuel .5cnsification Report, CAW-10055, Rev 1, Babcock & Wilcox, Lynchburg, Virginia, June 1973.

"penstilcation i Kinetics and Power Spike Model," Meeting with USAEC, July 3, l'174; J. F. liarrison (B&W) to R. Lobel (USAEC) Telecon, " Power Spike Factor,"

Jyly 13. 1974. -

Correlation of Critihal }{ eat Flux in Bundle Cooled by Prossurized Water, InW-10000, Babcock & Wilcox, Lynchburg, Virginia, March 1970.

~~

Cerrelation of Critical lleat Flux in Bundle Cooled by Pressurized Water, liAW-100 3o . Babcock & Wilcox, Lynchburg, Wriginia, February 1972.

I; ECCS Analysis of B&W's 177-FA Lowered *.oop NSS, BAW-10103, Babcock & Wilcox.

Lynehburg, Virginia, July 1975.

o A-1 Babcock e ililcox

_ _ _ _ - _ _ _ _ _ - - - _ _ -

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