ML19322C159

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Cycle 3 Reload Rept
ML19322C159
Person / Time
Site: Oconee 
Issue date: 04/30/1977
From:
BABCOCK & WILCOX CO.
To:
References
BAW-1452, NUDOCS 8001090569
Download: ML19322C159 (50)
v • d • e


Text

,

6' $

B.W-14 52

/

April 1977

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s OCONEE UNIT 2. CYCLE 3

- Reload Report -

NOTICE -

l i

I THE ATf ACHED fit ES AHE Of F1CI AL RECOrtDS Of 1HE l DIVISION OF DOCOVENT CONil40L T HE Y H AVE tie E N l CHARGED 10 WOU FOR A LtutTED TiVE PEHIOD AND

{ MUST DE RE Tu88NE D TO THE RECORDS _ FACJLITY

! HRANCH 016 PL E ASE DO NOT SFND DOCUME*TS g

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PAGE(S) FHOM OOCUMENT FOH HEPHODUC110N MUST l 1 eE eiErEsiRED 10 ritt eEnsoNstt.

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Babcock &Wilcox 8001090 N f t

e BAW-1452

.ipril 1077 OCONEE UNIT 2. CYCLE 1

- Reload Report -

BABCOCK 6 WILCOX Power Generation Group Nuclear Power Generation Division P. O.

Box 1260 Lynchburg, Virginia 24505 Babcock s. Wilcox l

CONTENTS Page 1.

INTRODUCTION AND

SUMMARY

1-1 2.

OPERATING HISTORY.

2-1 3.

CENERAL DESCRIPTION 3-1 4

FUEL SYSTEM DESIGN 4-1 4.1.

Fuel As= _=bly Mechanical Design.

4-1 4.2.

Fuel Rod Design.

4-2 4.2.1.

Cladding Collapse.

4-2 4.2.2.

Cladding Stress.

4-2 4.2.3.

Cladding Strain.

4-2 4.3.

Thermal Design 4-3 4.3.1.

Pcwer Spike Model (Densification) 4-3 4.3.2.

Fuel Te=perature Analysis.

4-3 4.4.

Mat e ri al De si gn....

4-4 4.5.

Operating Experience 4-4 5.

NUCLEAR DESIGN 5-1 5.1.

Physics Characteristics.

5-1 5.2.

Analytical Input 5-2 5.3.

Changes in Nuclear Design.

5-2 6.

THERMAL-KYDRACLIC DESIGN 6-1 6.1.

Evaluation 6-1 0.2.

DNBR Analysis..

6-1 6.3.

Pressure-Tenperature Limit Analysis.

6-2 6.4 Flux / Flow Trip Setpoint...................

6-2 6.5.

Mark C Demonstration Assemblies..

6-3 7.

ACCIDENT AND TRANSIENT ANALYSIS..................

7-1 7.1.

General Safety Analysis.

7-1

7. 2. Rod Withdrawal Accidents 7-1
7. 3.

Moderator Dilution Accident 7-2

7. 4 Cold Water (Pump Startup) Accident 7-3 7.5.

Lots of Coolant Flow 7-3 7.6.

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.

Steam Generator Tube Failure 7-5

................ Babcock & Wilcox 1

~~

I l

l CONTENTS (Cont'd)

Page 7.10.

Fuel Handling Accident 7-5 7.11.

Rod Ejecticn AceIdent 7-5 7.12.

Maximum Hypothetical Accident 7-6 7.11.

'*ste Gas Tank Rupture.

7-6

.a 7.1. LOCA Analysis..

7-6 5

Pr0 POSED MODIFICATIONS TO TECIISICAL SPECIFICATIONS 3-1 1

STAR 1"P PROCRAM,

9-1 10 RE R RENCES l '.-l List of Tables Table 4-1.

Fuel Design Parameters and Dimensions 4-5 4- !.

Input Sur. mary for Cladding Creep Collapse Calculations 4-6 4 - 1.

Fuel Thernal Analysis Parameters 4-7 3-1.

Oconee 2, Cycle 2 and 1 Physica Parameters 5-4 v-2.

Shutd. wn Margin Cceleulat ion for oconce 2, Cycle 3........

5-%

h-l.

Therr.al-ilydraulic Design Conditions...............

%-4 7-1.

Comparisoa of Key Parameters for Accident Analysis 7-8 I. i s t of Figures Figure

,-1.

Leu

..sading Diagram. Oconee 2. Cycle 3...........

3-3 3-2.

Enrichment and Burnup Dist ribut ion Oconec 2 Cycle 3....

3-4 1-3.

Cont rol Rod Locations, Oconee 2. Cycle 3 3-5 Mantrum Cap Si:e Ys Axial Position, Oconee 2. Cycle 3....

4-8.

4-2.

i'ot.er Spike Factor Vs Axial Position, Oconee 2, Cycle 3..

4-9

~. - l. MOC (4 i:-PD), Cycle 3 Two-Dimensional Relative Powe r Dist ribution -- Full Power, Equilibrium Xenon, Normal I

Rod Positions (Group 7 and 8 Inserted) 5-7 3-1.

Core Protection Safety Limits, Oconee 2, Cycle 3 S-2 8-2.

Core Protection Safety Limits, Oconee 2. Cycle 3 S-3 8 - 3.

' Core Protection Safety Limits, Oconee 2, Cycle 3 S-4 8-4 Protective Syste, Maximum Allowable SetPoints.

Oconee 2, Cycle 3.

8-5 ttt.

Babcock & Wilcox

I CONTENTS (Cont'd)

Figure Page 1-5.

Protective System Maxie.us A11cuable Setpoints.

Oconee 2. Cycle 3 8-6 M-*.

Rod Pcsition Limits for Four-Pu=p Operat ten Frec 0 to 100 2 10 EFPD, Oconee 2. Cycle 3 8-7 ri-7.

Red Positica Limits for Four-Pu=p Operation Frca 100 2 10 to 250 : 10 EFPD, Oconee 2 Cycle 3 8-8 3-5.

Rod Position Limits for Four-Pump Operation Af ter 250 1 10 Ef?D, Oconee 2 Cycle 3..

8-9 M '+.

Rod Posit t en Limit s for Two-and Three-Pump Operation From 0 to 100 : 10 EFTD, Oconee 2 Cycle 3....

8-10 5-10.

Rod Position Linits for Two-and Three-Pump Operation From 103 : 10 to 250 e 10 EFPD, oconee 2, Cycle 3 8-11 t!-1 1.

Rod Posit ion Limi ts for Two-and Three-Pump Operation Af ter 250 t la Cr'D, Oconee 2. Cycle 3.........

8-12 3-12.

Operational Pwer != balance Envelope for Operation From 0 to 100 : 10 Ef?D, Oconee 2 Cycle 3...........

8-13 8-13.

Operational Pwer 1thalance Envelope for Operation From 100 t 10 to 250 10 Ef7D, Oconee 2. Cycle 3..........

3-14 8-14.

Oper.ational Pwer 1r. balance Envelope for Operatini Af ter 250 : 10 E}TD, oconee 2 Cycle 3............

8-15 8-15.

APSR Position Linits for Operation From 0 to 100 : 10 EITD, Oconee 2, Cycle 3 8-16 S-16 APSR Position Limits for Operatien Af ter 250 : 13 EFTD, Oconee 2. Cycle 3 8-17 i-l ',

APSR Position Limits for Operation From 100 lo to 250 : 10 ETPD, Oconee 2, Cycle 3..

8-18 I

i i

. - iv -

Babc0Ck & Wilcox

1.

INTRODUCTION AND SL N Y This report justifies the operation of the third cycle of Oconee Nuclear Sta-tion. Unit 2 at the rated core powir of 2568 MWt.

Included are ti.e required analyses outlined in the USNRC document " Guidance for Proposed License Amend-ments Relating to Refueling. dated June 1975. To scpport Cycle 3 operation of Oconee Unit 2. this report employs analytical techniques and design ba.ses established in reports that have been submitted and accepted by the USNRC and its predecessor (see references).

A brief summary of Cycle 2 and 3 reactor parameters related to power capabil-ity is included in section 5.

All the accidents analyzed in the FSAR have been reviewed for Cycle 2 operation.

In cases where Cycle 3 characteristics proved to be conservative with respect to those analyzed for Cycle 2 operation, no new accident analyses were perforced. The Technical Specifications have been reviewed, and the modifications required for Cycle 3 operation are justi-fled in this report.

Based on the analyses performed, which take into account the postulated effects of fuel densification and the " Final Acceptance Criteria for Emergency Core Cooling Systems." i t has been concluded that oconee 2. Cycle 3 can be operated safely at the rated core level of 2568 MWt.

l 1-1 Babcock & Wilcox i

N

2.

OPERATING HISTORY The reference fuel cycle for the nuclear and thereal-hydraulic analyses of the third cycle of Oconee Nuclear Station. Unit 2 is the presently operating Cycle 2.

Cvele 2 schieved initial criticality on July 8, 1976, and power escalat ion began on July 12, 1976. The 100% power level of 2563 MWt was reached on July 18, 1976. No operating anomalies occurred during Cycle 2 operatien that woula adversely affect the fuel performance in Cycle 3 during the design length of 292 EFPD.

No control rod interchanges are planned for Cycle 3.

Control rod group 7 will be withdrawn at 250 (f l0) EFPD of operation.

2-1 Babcock s.Wilcox

3.

CENERAL DESCRIPTION The Oconee 2 reactor core ss described in detail in Chapter 3 of the Oconee Fuelear Station FSAR.I The Cycic 3 core, comprising batches 3. 4, and 5, con-sists of 177 fuel assemblies (FAs) - 173 of which have a 15 by 15 array contain-ing 208 fuel rods. 16 control rod guide tubes, and one incore instrument guide tube.

The cold-worked, Zircaloy-4 fuel rod cladding has an OD of 0.430 inch and a wall thickness of 0.0265 inch.

The fuel consists of diched-end, cylin-drical pellets of uranium dioxide (C0 ), vnich are 0.370 inch in diameter.

2 (See Tables 4-1 and 4-2 for additional data.) The other four FAs in Cycle 3 are demonstration 17 by 17 FAa two Mark C and two Mark CR.

All FAs in Cycle 3 except the 17 by 17 demonst ration assemblies, maintain a constant nominal l

fuel loading of 463.6 kg of uranium. However, the undensified nominal active fuel lengths and theoretical densities vary between batches; these values are given in Tables 4-1 and 4-2.

Figure 3-1 is the core loading diagram for Oconce 2, Cycle 3.

All the batch 2 assemblics will be discharged at the end of Cycle 2.

Five once-burned batch 1 assemblies with an initial enrichment of 2.06 wt % 235U will be reloaded in-to the central portion of the core.

Batches 3 and 4 with initici enrichments of 3.05 and 2.64 wt % 235U, respectively, will be shuffled to new locations.

Eatches 5 and Sa with initial enrichments of 3.03 and 2.53 wt I 2'5U, respec-tively, primarily will occupy the core periphery and four interior locations.

Figure 3-2 is sa eighth-core map ahowing the assembly burnup and enrichment itstribution at the beginning of Cycle 3.

Reactivity control is supplied by 61 full-length, Ag-In-Cd control rods and j

soluble boron shin.

In addition to the full-length control rods, eight partial-langth axial power shaping rods (APSRs) are provided for additional control of axial power distribution. The Cycle 3 locations of the 69 control rods and the group designations are indicated in Figure 3-3.

The core locations of the tctal pattern (69 control rods) for Cycle 3 are identical to those of the initial cycle described in Chapter 3 of the FSAR.1 llowever, the group 3_;

Babcock s.Wilcox

f designations differ between Cycle 3 and the reference cycle to minimize power 2

ea k i n g,. The r.oninal system pressure is 2200 psia, and the core average densi-t h d nominal he it rate is 5. 79 k'el/f t at the rated core power of 2565.wt.

1.

l l

l 3-2 Babcock a Wilcox i

fi.;ure i-1.

Core Lo.adi..; Dia,:ra.. Oconee 2. Cycle 3 Fult InsGsta C444L I

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l'igure 3-2.

Enrichnent and burnup Distribution, f

Oconee 2, c.;1e 3 6

9 10,

11 12 13 14 15 l

2 06 2.06 2 64 3.05 2 53 3.65 3 05 3 03 H

12965 10931 7179 23056 0

19529 23523 0

2.64 2.64 2 64 3.05 2.64 3 05 3 03 l

K 7133 9181 65ts IEDBB 10023 19616 0

l 3 05 2.64 2.64 3.05 3 03 3.03 L

20395 6957 11793 21517 0

0 2 64 3 05 3.05 3 03 4

5940 22153 23964 0

3.05 3.03 3 03 N

23457 0

0 2.64 0

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l it 1.11 initial Enricament XIII BOC Burnup Mea ~stU i

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Babcock & Wilcox i

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Coatral kod Loca'icas, Oconee 2. Cycle 3 I

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8 APSRs TOTAL 63 l

3-i Balcock r. Wilcox i

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U f

4 FUEL SYSTEM DESICX 4.1.

Fuel Assembly stechanical_ Design The Cycle 3 core consists of the norml resident and reload Mark B fuel aswee-blies plus five batch I assemblies. 2 Mark C (17 by 17 array) demonst ratien asse=blies (part of batch 4), and 2 fresh Mark CR (17 by 17) demonstration assechlies. The pertinent f uel design parameters, and dimensions are list ed in Table 4-1.

The five f uel asse=blies f rom hatch 1, which are once-burned as-senblies, are mechanically similar to those in batch 1.

The fresh fuel asw n-blies (batch 5) incorporate minor design modification.. to the spat.*r grid cer-ner cells, which reduce spacer grid interaction daring handling.

In additien, leproved test methods (Dyaamic Impact test inn) 8 hew the spacer grids to have a higher seismic capability and, thus an increased u:ety margin over t he values reported in reference 3.

The two Mark CR ass.emblies included in batch 5 are mechanically identical in f unct ion to the Mark C demonst ration assemblies o: batch 4.

The Mark CR assen-blica are different because they have recanst itu:.able lower end fit t ings; Sev-ever, no reconstitutabilit y tests are scheduled fer the Oconee 2 pool. The mechanical design of the Mark C demonstration asse:.blies and the comparison be-twsen the Mark C and 11 assechlics are given in reference 4.

All fuel assemblics in the Cycle 3 core are mechanically interchangeable, but the Mark C and CK den-onstration assemblies cannot he located in a rodded location. The static and dynamic structural characteristics of the Mark C and CR demonst ration assemblies are compat ible with the Mark B assemblies and haw been designed to m..intain their mechanical integrity throughout the three celes of operation and to successfully withstand all seismic and 10CA loads postulated for the Oconee 2 reacter.

All cther results presented in the FSAR discussion of the fuel. assechly mechani-

~

cal design-are applicable to the reload FAs.

4-1 Babcock & Wilcox m________m

4_. 2.

Fuel Rod Design 4,. _2.1.

Cladding Collapse Creep collapse analyr.es were performed for three-cycle power histories for Oeonee 2.

Table 4-1 is a summary of the batch 3, 4, and 5 fuel designs. The FA power histories were analyzed, and the most limiting histories were deter-mined. Specific assembly power histories were used in the analysis of batch 3.

Batch 4 and 5 fuel was analyzed using a conservative power history envelope.

The =ost Ilmiting assembly is in batch 3 because of the low initial fuel pellet density, thc low initial prepressure, and the previous in-core exposure time.

An input succary for the batch 3 cieep analysis is contained in Table 4-2.

The predicted power history f or the most limiting assembly was used to de-teimane t he mininum collapse

  • ime as described in BAW-100S4P, Kevision 1 (ref-erense 5).

The conservatiscs in the analytical procedure are su=marized belcv:

1.

The CROV, Revision 1 computer code was used to predict the time to col-lapse; CROV, Revision 1 conservatively predicts collapse times.S 2.

So credit is taken for fission gas release; therefore, the net differen-tial pressures used in the analysis are conservatively high.

4 The cladding thickness used was the lower tolerance limit (LTL) of as-Suilt measureecnts; the initial ovality of the cladding used was the upper tolerance limit (UTL) of the as-built measurements. These values were tahen f ro:a a tatistical saepling of the cladding.

6.

A conservative power history envelope was used in the batel 4 and 5 fuet

.inalysis.

The most limitiny, assembly had a c.ollapse time greater than the maximum pro-lected three-cycle design life (Table 4-1).

This analysis used the assumptions on densifIcation described in reference 5.

4.2.2.

Cladding St ress The batch ) fuel is.he most limiting from a cladding stress point of view due t o t he lower prepressurization and low density.- The calculations in BAW-1395, reterence 6 represent the most limiting case for Oconee 2. Cycle 3.

4. 2.1.

Claddin LStrain

]

- The fuel design criteria specify a limit of 1.0* on cladding circunferential plastic strain. The pellet der:Ign results in a plastic-cladding strain of less than 1. 01 a t 55,000 Wt/mtU. The following,conservatisms were used in this analyst.-

4-2 Babcock s.Wilcox g

l.

The maximum sp.cification value for the fuel pellet diameter was used.

2.

The maximus specification value for the feel pellet density was used.

3.

The cladding ID used was the lowest permitted specification tolerance.

4.

The maximus local pellet burnup (expected three-cycle) is less than 55.000 5'd/mtU.

4. 3.

1hermal *)esign

.\\ l l f uel assemblies in this core are thermally similar. The fresh batch 5 tuel in erted f or Cycle 3 operation does not introduce any signif tunt dif-ferences in fuel thermal performance relative to the batch 2 fuel discharged at the end et Cycle 2.

The linear heat generation rate capability of the batch 4 and 5 fuel (Table.-4) is greater than that of batches 1 and 3 (20.15 kW/t t versus 1*.fi kW/tt). These linear heat rate limitations were established using the TMT-3 ' cod ' witn fuel densificat io n penalt ies.

The two !! ark CR assemblie t In batch 5 are thermally identical to the two >! ark C (bat.h ".) de.w nstratten assemblies described in reference 4.

The four demon-

.t rat ion.ns.cmSlies ha.e been placed in nonlimiting core locations.

3.1.

power Sp,ike SIMe! (Densificarlon)

I' 5 e power spikt-nodel uwd for Cycle 3 analysis is the same as that used for Cycle.'. 2 Figures 4-1 and 4-2 show the maximun gap size and power spike fac-tar, respectively, versus axial position. The pcwer spike factor and gap size are based on unirradiateo hatch I t uel (92.51 D) with an assumed enrichment on.l.0 wt

' t'. These values are conservatively high for all batch 1. 4 and 5 fuel.

  • ..t..'.

Fuel Tepperature Analysis j.,

Therru t analysis of the :uel rods assumed in-react or densificat ion to 96.5*

theoretical donalty (TDF). The analytical methods utilized are the same as those doeunented in references 2 and 6 for Cycles 1 and 2 respec t ivel y.

The average f uel t s speratures shown in Table 4-4 are taica from the analyses ut i-li.ed to define the linear heat rate (LHR) capaSility for the fuel (references

,' and 6). This analysis is based on the lower tolerance limit of the speci-fication fuel density and assumes isotropic dia etral shrinkage and anisotropic axial shrinkage (consistent with reference 8) resulting from fuel densiflea-tlon.

4-3 Babcock s. Wilcox

+

t 4.*.

%t erial_ Design 1he batch a it.el assemblies are not new in cencept, and they do not utilize different ee:ponent materials. Therefore, the chemical cocpatibility of all poasible fuel-eladding-coolant-assembly inter.actiens for Lt.e batch 5 fuel as-se:blies are identical to those of the present fuel.

4. 5.

Operating i:xpcLience bus operating experience with the Mark B, 15 by 15 FA design has verified the adequacy et this design. A4 of December 31, 1976, the following operating experience taas been accumulated for the six bu' 177-FA plants using the.'iark B i

tuel assembly:

l Max. ase.emb1y Cuenala t ive Current burnup net electrical Re.netor cycle Wd /mtU output, Wd Oeenee 1 3

23,000 15,232,533 Ocener 2 2

22,200 10.564.123 Oeonce 3 2

19,800 9,931,642 T?t l-1 2

23,700 11.854,960 Arian%2s one 1

19,253 8.957,632 R.in, h.* Seco 1

10,761 3.511.597 44 Babcock & Wilcox

)

Table /.-l.

Fuel Design Parameters and Dimenelons twice-burned Fresh fuel

,pisemblles once-burned assembiies assembiles Ratch 3 Batch 4 Batch 1 Mark C demo Batch 5 Mark CH demo FA type Mark B-3 Mark B-4 Mark B-2 17 by 17 Mark B-4 17 by 17 array array No. of assemblics 60 54 5

2 50/4 2

^

Fuel rod OD. in.

0.430 0.430 0.430 0.379 0.430 0.379 Fuel rod ID, in.

0.377 0.377 0.377 0.332 0.377 0.332 Flexible spacers, type Corrugated Spring Corrugated Spring Spring Spring

-Rigid spacers, type Z rO,-

Zr-4 Z ro,,

Zr-4 Zr-4 Zr-4

-Undensified active fuel length In.

144.0 142.6 144.0 143.0 142.25 143.0 e

Fuel pe11et OD (mean w

D.370 0.370 0.370 0.324 0.3695 0,124 specified), In.

Fuel peIIct initial 92.5 93.5 92.5 94.0 94.0 94.0 density, % TD Init1.

wt%;1,Ufuel enrichment, 3.05 2.64 2.06 2.64 3.03/2.53 3.03 Initial. fill gas pressure Same as Sam. as Mark i

(minimum'specified), psia (a)

(a)

(a)

(a) batch 4 C demo Burnup BOC,.WJ/mtU 21,356 8248 11,338 5952 0

0 g

Cladding collapse time,

.g r.FPil

,30,000-

>30,000 e10,000

>30,000 e30,000 630,000 8

Design life, EFPil 24,912 21,360 17,568 20,088 21,024 21.024 m-98 '

("' Refer to proprietary information in reference 2. Table 4-1.

M

I Table 4-2.

Input Sumnary for Cladding Creep Collapse Calculations Batch 3 Pellet OD (::ean specified), in.

0.3700 Pellet density (mean specifled), % TD 92.5 Densifled pellet OD, in.

0.3650 Cladding ID (mean specifteo), ic 0.377 Cladding ovality (UTL), in.

(a)

Cladding thickness (LT!.), in.

(a)

Prepressure (ninimum specified), psia (a)

Post-densification prepressure (cold), psia (a)

Reactor systen pressure, psia 2200 Stack height (andensiffed), in.

144.0

" Refer to proprietary information in reference 2 Table 4-3.

I N

4-6

. Babcock & Wilcox

Table 4-3.

Fuel Thermal Analysis Parameters Batches 1 ar.d 3 Batch 4 Batch 5 I#

No. of assemblies 5, 60 56 56(b)

Initial density. % TD 92.5 93.5 94.0 Pollet diameter, in.

0.370 0.370 0.3695 Stack height, in.

144 142.6 142.25 pensif ied Fuel Parameters (*)

Pellet diameter, in.

0.3632 0.3645 0.3646 Fuel stack height. in.

141.1 140.5 140.5 Nominal linear heat rate 5.77 5.80 5.80 at 2568 MWt. kW/ft Average fuel temperature at nominal LilR. F 1335 1320(d)

Id}

1320 l

Linear heat rate capability (centerline fuel melt). kW/ft 19.8 20.15 20.15

  1. Includes two tbrk C (17 by 17) demonstration assemblies.

(

Includes two tbrk CR (17 by 17) demonstration assemblies.

(# Densification to 96.5% TD assumed.

I 11 ark C and Mark CR fuel will operate at a lower average heat rate and a corresponding lower average fuel temperature.

4-7 Babcock 3.Wilcox

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-8 Babcock s. Wilcox

Figure J. - 2. Power Spike Tactor Vs Axial Positten, Oconce 2 Cycle 3 1.10 TOF = 96.5%

TOI = 9 2. 5*,

e=

3.005 1.08 3

1.06 M

u 2

Y 5

e 3

1.04 2

1.02 co, acr n

8 1.00 1

I I

I I

I x-e.

0 20 40 60 80 100 120 140

E#

Arial Position, incnes E

f e

a

5.

Nt' CLEAR DESIGN 5.l.

Phys 1:n Characteristics

(

Table 5-1 compares the core physics parameters of Cycles 2 and 3.

The values for both cycles were generated using PDQ07. Since the core has not yet reac hed an equilibrinn cycle, differences in core physics. parametets are expected be-tween the cycles.

.e shorter Cycle 3 will produce a smaller cycle dif ferential burnup than that for Cycle 2.

%e accumulated average core burnup will be higher in Cycle 3 than in Cycle 2 because of the presence t f the burced batch 1. 3. and 4 fuel.

Figure 5-1 illust rates a representative relative power distribution for the beginning of the third cycle at full power with equilibrium xenon and r.ormal rod positions.

The crit ical boron concentrations for the beginning of Cycle 3 are lower in all cases than those for Cycle 2.

End of cycle conditions vary between Cyc1h 2 and 3 cat. Sing higher critical boron concentrations. As indicated in Tabse 5-2. the cont. I rod worths are sufficient to maintain the required shutdown margin. Ilowever, due to changes an isotopics and the radial flux distribution.

the hot, full-power control rod.rorths will be less than those for Cycle 2.

The Cycle 3 ejected rod worths for the same nurber of regulating banks inserted are lower than those in Cycle 2.

It is difficult to coc: pare values between 1

cycles or between rod patterns since neither the rod patterns from which the CRA is assumed to be ejected nor the isotopic distributions are identical.

Calculated ejected rod worths ant' their adherence to criteria are considered at all times in life and at all power levels its the development of the rod in-Hertton limits presented in section 8.

The maxisma stuck rod worths for Cycle 3 are similar to those in Cycle 2.

The adequacy of the shutdown margin with Cycle 3 stuck rod worths is demonstrated in Table 5-2.

The following conserva-Lista were applied for the shutdown calculations:

5-1 Od & EM

l.

Poison naterial depletion allowance.

2.

10 uncertainty on net rod uorth.

3.

Flux redistribution penalty.

Flux redistribution was 7ccounted for since the shutdown analysis was calculated using a two-dimensional nodel. The shutdown calculation at the end (f Cycle 3 is analyzed at approximately 250 EFPD (210). This is the latest time (210 days) an core life at which the transient bank is nearly fully inserted. After 250 EFPD, the transient beni vill be almost fully withdrawn, thus increasing the available shutdown margin. The reference fuel cycle shutdown margin is pre-sented in BAW-1425.2 The Cscle 3 power deficits frou hot zero power to hot full power are similar to but slightly lower than those for Cycle 2.

Doppler coefficients, moderator coef ficients, and xenon worths are similar for the two cycles. The dif ferential baron worths for Cycle 3 are higher than those for ' Cycle 2.

The effective de-layed neutron fractions for both cycles show a decrease with burnup.

5.2.

Analvtical lnput The Cycle 3 incore measurement calculation constants used for computing core power dist ributions were prepared in the same manner as for the reference cycle.

5. 3.

Chances in Nuclear Design There were no relevant changes in core design between the reference and reload cycles. The same calculational methods and design information were used to obtain the leportant nuclear design parameters for Cycles 2 and 3.

The only significant operational procedure changes from the reference cycle are the specification of APSR position limits in addition to the usual regulating con-trol rod and inhalance limits for ECCS. The operational limits (Technical Specif ication changes) for the reload cycle are shown in section 8.

The FLAME code **l2 was used in setting the Technical Specification limits.

The nuclear characteristics of the two fresh, batch 5, Mark CR (17 by 17) fuel ause=blies are nearly identical to the Mark B (15 by 15) assemblies that make up the balance of batch 5.

The similarity in nuclear design between 17 by 17 and 15 by 15 fuel assemblies is described in reference 4 where the batch 4 Furk C demonstration assemblies are compared to batch 4. Mark B asse=blies.

Therefore, the presence of the 17 by 17 demonstration FAs (two Mark C and two Mark CR) will not discernably af fect overall core reactivity coef ficients Babcock s.Wilcox 5-2

or performance. Ilowever, since the two.%rk C and two Mark CF fuel assemblies are demonstration assemblies, standard practice dictates their placement in noniimiting core locations during Cycle 3.

1 5-3 Babcock & Wilcox l

I h

Tab'.* 5-1.

Oconee 2. Cycle 2 and 3 Physics Parameters Cveie 2

  • Cvele 3

(.

Cycle length. EFPD 306 292 Cycle burnup. I'Jd/mtU f.

9582 9142 I

/

rage core burnup - EOC. E'd/=tU 18.606 19.307 Initial core loading, ratL' 62.1 32.0 l

Cri t i ca l boron - BOC. (no Xci, ppa llZP(c)

. group 8 (37.5: wd) ll7,P. groups 7 and 8 inserted

~~

1445 1310 1330 1200 ilFP. grotes 7 and 8 insertoi

!!40 1050 Crit ical boron - I.0C (eq Xe), ppn liZP 0, group 8 (37. M wd. eq Ie)

'434 300 lirPs 87 60 Control rod worths - IIFP(*

BOC

  • 4k/k Croup 6 1.20 1.07

{

Croup 7 0.96 0.90 Croup 8 (37.5% vds 0.54 0.33 rentrol rod worths - IIFP (250 EFPD). I ek/k Croup 7 i

1.33 1.11 l

Crot.p M (37.5% wd) 0.51 0.40 I

?'.ex ejec ted r.2d vort h - IlZP. I Ik/k hoc IdI 250 ETPD O.59 0.37 0.58(d) g,47(d) 2bx stuck rod worth - IIZP. % Lk/k BOC 250 ETPD

?.16 2.13 2.22 2.r:9 Power deficit - Il%P to lirP. % Lk/k BOC 250 EFPD 1.65 1.54 2.49 2.14 Loppler coeff. - BOC. 10-5. Ak/k/*F 100% power (no Xe)

-1.51

-1.48

. Nopler coef f. - EOC. 10~S. Ak/*/*F 100% power (eq. Xe)

-1.55

-1. 51

?bderator coef f. -- IIFP. 10~". ak/k/*F BOC EOC

-1.03

-0.57 g

-2.60

-2.66 I

Boron worth - HFP. ppa /Zak/k BOC (1000 ppm) f ';

109 106

-Eoc (17 ppm) 101

'9g 5 Babcock & Wilcox_

{

6 e

~

l Table 5-1.

(Cont'd)

Cycle 2

  • Cycle 3 Xenon worth - !!FP, % 1*a/k BOC (4 days) 2.60 2.65 EOC (equilibriu:3)'

2.66 2.75 Ef fective delayed neutron f raction - 11FP BOC 0.00577 0.00593 EM 0.00516 0.00531

"'Ba'ned cm Cycle 1 Ict:gth of 440 EFPD.

' Cycle 3 data are for the condition stated; Cycle 2 condition may not be the same.

  1. 'liZP - fot zero power; ilFP - hot full power.
  1. jected rod value fer groups 5, 6, 7, and 8 inserted.

E 5-5 Babcock s Wdcox b

4 e

r; I

Table 5-2.

Shutdown 'fareiq_ Calculation for oconee 2,_Cyc'e 3(#

f Available kod Worth BOC, %2k/k EOL. t',k/k(b)

Total rod worth. ICP(#}

8.37 8.44 Worth reductinn due t f

-0.25

-0.32 burnup of paison raterial Nxtezum stuck rod li2P

-2.18

-2.09 Wet worth 5.94 6.03 Less 10' uncertainty

-0.59

-0.n0

[

Total available worth 5.35 5.43 l

Re, quired Iod Wort _h Power deficit. hip to llZP 1.54 2.14 N x. allowable in3crted rod sorth 3,39 g,39 riux redistribution

_0. 34 O_.pp, Total required worth 3.07 3.99 Shytdowuj argin,,

Iotal available worth ninus 2.28 1.44 total rei; s tred wort h (a) Required shutdown margin is 1.00: 1.k / k.

(5)For shutdown nurgin calculations, this is defined as approxicately l

250 EF?D - the latest time in core 18fe at which the transient bank is nearly fully-in.

(e)!!ZP denotes hot zero pc.rer; HFP det >tes hot full power.

e

}-

I l.

u Babcock & Wilcox

Figure 5-1.

ItoC ('. ETPD). Cycle 3 ivo-Dicensional Relati.*e Tower Distribution-Full Power. Equilibric: Xenon. D rr.al Rod Positions (Group ' and 3 Ir.scrted) 8 9

10 11 12 13 14 15-l' H

1.11 1.24 1.42 1.24 1.39 0.92

0. 48 0.87 K

1.24 1.39 1.28 1.30 1.11 8.99 0.81 0.75 7

8 L

1.42 1.29 0.73 1.13 0.91 0.92 1.23 0.72 M

1.24 1.30 1.13 1.11 0.89 0.90 1.04 8

N 1.39 1.11 0.S0 0.89 0.92 1.18 0.7E 0

0.92 0.99 0.92 0.90 1.17 0.70 7

P 0.47 0.81 1.20 1.04 0.76 R

0.67 0.76 0.72 I

INSERTED R00 GROUP N0.

I. II RELATIVE POWER DENSITY 5_;

Babcock s. Wilcox

i t

c.

TilERMAL-hiDMULIC DESIG 6.1.

Evaluation The thermal-hydraulic design evaluation supporti:2 Cycle 3 operation utilized the same methods and codels previnus!y described in references 1. 2. 4 and 6.

Cycle 3 :=nalyses have been based on 105. 5% of the (first core ) design syste:n flow rate and fully closed internal vent valve configuration. Cycle 2 analy-sss (reference 2) had used 107.'% of design flow based on a measu. d flow a

value of 111.5 (tl.0%). The reduced flow rate has been selected for Cycle 3 cualyses to provide consistenc> with Oconee 1 (reference 11).

The cor e conf igurat ion for Cyc.e 3 dif fers f r<rs tnar of Cycle 2 in that t hs-h-itch 2 f uel removed at the end of Cycle 2 is the brk 8-3 fuel assembly de-si;,n, while the fresh b. itch 5 fuel inserted for Cycle 3 is the Nrk B-4 as-l sembl. des ir,n.

W r k II-4 is c::hlies exhibit a slightly lower resistance to l

flow than do the.tark B-1 assemblies resulting free a revised end fitting de-l

.lgn.

this change lus been considered in the Cycle 3 core flow distribut ion i

.in.e l y. i 4 No credit has been taken f. r the increase in syste s f Iow, which I

resol t s i ro:n t he reduct len in total core pressure drop.

The Mrk C and CR demonstration assemblies have overall resistances greater than M rk B assemblies. Haever, the effect of this additional resistance on the total core pressure drop and system flow is negligible and therefore not included in the analysis.

h.2.

HNHR Anaivsis The 1M-2 CliF correlat ion tus been used for the therml-hydraulic.inalysis of Cycle 3.

This correlat ion, which has been revimd and.spproved f or use with the Nrk-Il fuel assembiv design,12 has been previeusly used ter lic ensin,: of Cycle 2 of the Oconce 2 core.'

The effect of fuel densification on nin;:num DNBR is primarily a result of the reduct ion in act ive tuel length, which increases the average heat f lisx.

The i

e-1 Babcock s. Wilcox

I Cvele 3 DNBR analysis was based on a cold densified active length of 140.3 In., a value which was selected to apply generically to a nu=ber of B&W plants.

This is a conservative method of applying the densification ef fect since all I

the fuel assemblies in Cycle 3 have longer densit'ied lengths and since no credit is taken for axial thernal expansion of the fuel colu=n.

This anat y-sis differs iron that of Cycle 2 in two respects:

(1) the ef f ect of the den-sif icat ion power spike is no longer considered for DNBR analysis on the ba sis s'

inf ormation presented in ref erences 14-16, and (2) the densified active length is incorporated directly into the I NBR analvsis resulting in a calcu-

'at ed nin t=un DNBR of 1.914 at 112'. power (Table e-1).

The C;. ele 2 an.nlysis had been based on a 144-inch active length with the ef f ect or reduced active length.ind the densification power spike calculated separately.

The potenti.nl effect of fuel rod hew on DNBR is considered by incorporating' suitable margins into DNS limited RPS setpoints.

I'.e maximum rod bow magni-11.5 + 0.0t>9 [B3 where t ude was cal. ulat ed f rom t he equat ion e s is the

=

b h

.3 red how magnitude in mils and BU la the burnup in M'd/mtU.

The resulting a

DNBR penal ty based on the naximu:2 predicted assembly burnup at en of Cycle 3 is 6.0Z.

6. 3.

Pressure v_emper.iture Limit Analysis C

Pressure-te=perature limit curves chown in Figure 2.1-3b of the Oconee Nuclear Station Technical Specifications (Figure 8-3 of this report) provide the basis ror the variable low-pressure trip setpoint. The curves shown for four-and three-pump operation represent a locus of points for which the calculated MDNBR is equal to 1.30 (BAW-2) plus a suitable margin to offset the CNBR reduction i

lue to rod bow (discussed in section 6.2).

l Q.

Flux / Flow Trip Setpoint The flux / flow trip setpoint is determined by analysis of an assamed two-pump eesstdown starting from an initial power level (irdicated) of 102*.

A flux /

flow trip setpoint of 1.055 is established for Cycle 3 and is based on a nin-loua DNBR of 1.3 plus a suitable margin to of fset the DNBR reduction due to rod bow.

The DNBR margin provided to offset the reduction in DNBR due to fuel rod how includes credit for 1% excess flow.

i l

6-2 Babcock 8.Wilcox e

6.5. - mrk C Demonstration Ass..mbite.

The DNBR analysis has been based on a core configuration consisting of 177 N rk B (15 by 15) fuel assemblies. Co:nparative analyses have been performed to show that the insertion of the Nrk C and CR (17 by 17) de:nonstration as-actblics will increase the ?IDNBR in the hot assemoly. The demonstration as-sc-h11es have been placed in low power producing core locations to ensure that i

t!.ese assemblies will not be limiting and to provide minimum impact on the

{

h. t asse: ably perforn.ince. Therefore, the presence of the two Wrk C and the L CR deconstration assenblies in Cycle 3 will not tua Ki:

discernibly affect ti.e thernal-hydraulic character of the reactor.

i s

I L

'1 50skenca, e tasa..

F 7

y I

A Table _6-1.

Thernal-Hydraulic Design Conditions _

's II l

Cycle 2 Cycle 3 power level. ?'Wt 2568

- 2568 Syst e-o pressure, psia 2200 2200 Reactor coolant flow, % design flow 107.6 106.5 Vessel inlet coolant t ecipera ture 555.9 555.6 (100% power). F Vessel outlet coolant tanperature, 602.2 602.4 (100% power). F Ref design radial - local pwer -

1.783 1.783 peaking, factor Ref design axial flux shape 1.5 cosine 1.5 cosine A.'tive fuel l en.;t h, In.

(refer to Table 4-3)

! Average heat flux (1000 power),

~17 % '.0 "}

175640(#

I f1t ulle~ t t '

8:tif rorrelat ten RAW-2 BAW-2

  • tininuta DNBH t eux, des ign condi t ions, 1.93 (b) no da n41l Icat Ion penait les)

(112* power)

Ito t c han nel facters Enth.stry rise

~1.011 1.011 ilea t flux 1.014 1.014 flow area 0.98 0.98

!!!ninun DN!!R wit h densif icat ion pen.nl ty 1.892 1.914

(.s)P.and on d n.Ifled length of 140.3 Inches.

(hl See seetion 6.2.

a

'I

. t' t

4 l

Babcock & Wilcox l

,r p

4 m-w

t

.~

7.

ACCIDENT AND TRANSIENT ANALYSIS 7,. l. General Safety Analysis F.ach FSARA a.ccident analysis has been examined, with respect to changes in Cycle 3 par.nete rs, to determine the ef fects of the Cycle 3 reload and to en-sure that tN rnal performance is not degraded during hypothetical transients 1he core therru t parameters used in the FSAR accident analysis were design op-erating values based on calculated values plus uncertainties Cycle I values (FSAR values > of core thermal parameters are compared with those used in the Cyrle 3.inalysis in Table 6-1.

These paraseters are common to all of the ac-eident analpes presented herein. For each accident of the FSAR, a discussion anel t he &cv parameters are provided.

A co=parison of the key parameters (Table 7-l) t ron the FSAR and Cycle 3 is provided with the accident discussion to sh w o

that tSe initial conditions of the transient are bounded by the FSAR analysis.

The en firt 6 et fuel densification on the FSAR accident results have bceu eval-uat ed and are reported in BAW-1395.6 Since cycle 3 reload fuel assemblies

.ontain suel r..ds with theoretteil density higher than those considered in reterenec 6 the conclusions derived in that reference are still valid.

r.il. ul.et len al techniques and methods for Cycle 3 analyses rem.nin s onsi stent with those used ior the FSAR.

Additional DSBR m.orgin is shown for Cycle 3 be-

.. ne.e the LW ' CilF correlat ion was used instead of the W-3.

No new dose calculations were performed for the reload report.

The do.we con-

-lihrations t the FSAR were based on maxicum peaking and burnup ter all core

.v.les; there: ore. t he dose considerat ions are independent of t he reload batch.

i..'. Rod Wit hf rawal Acc ident s This accident is defined as an uncontrolled reactivity addition to the core due to withdr.rwal of control rods during startup conditions or from rated luwer cond i t iccis.

Nith types of incidents were analyzed in the FSAR.I 7-1 Babcock 4. Wilcaw 9

i The i=portant paracet./s durinsi a rod withdrawal accident are the Doppler co-eflieltnt, the moderator te ;crature coef ficient, an.! the rate at which re-art!vity is added to the co:e.

Only high-pressure and high-flux trips are f

ao aunted f or in the FSNt.s.alysis, whicn ignores multiple alarms, i n t e rl oc h.,

and trips that normally preclude this type of incident. For p sitive reac-g tivity additions indicativ,

.f these events, the moat severe results occur fer Hol. condi t ions.

The FSAR values of the key parameters for IML conditions were

-1.17 10 (2k/k/'F) for the Doppler coefficient. 0.3 = 10 ak/k for the

! n.ultrator temper.sture coet fielent, and rod group wrths of up to and includin.;

a 10. ?k/k rod bank worth. Comparable Cycle 3 para =ctric values.are -1.48 10

(*k/k/*F) for the Nppler coef ficient, -0. 57 10 (ak/k/ *F) for the

~

1 r.oder.st..r temperature coef fielent, and a maximus rod bank worth of 8.17%

'k/k.

Iberefore, Cycle 3 parameters are bounded by design values.nssuned for the t%'d.maly is, and th'.as for the rod withdrawal transients, the consequences will be no more severe than thost pr.iented in the FSAR. For the rod with-d r.r. i l tron rated power, the transient consequences are also less severe than tho w presentui in t he densif icat ion report.6

/. 5.

%Ierat or Dilut ton Accident l'.i."i in t he f orm of boric. acid is use<! to control excess reactivity. The boron sontent of the reactor coolant is periodically reduced to compensate

~

ior tuel burnap and transient xenon ef fects with dilution water supplied by t h.

namenp.and pur if icat ien t.W6P) sys t em.

The moderator dilution t ransiect.

.on idered are the pumping of water with zero bores concentr.at ion f rom the r.ak. up t.mk t o the RCS under condit ions of f ull-pc rcr operation, hot shutdown, a nil

r. fueling. T h t-key para =eters in this analysis are the initial boron

. em sat rat lon, boron react ivity worth, and the moderater temperature coef f 1-

.-lent for power cases.

For posit Ive react tvity addition of this type, the cust severe results occur n or WL condit ions.

The FiAA values of the key parameters for BOL conditions were 1;00 ppm for the initial boron concentration, 75 ppa /1.0: (*.k/k) baron react tvity worth, and +0.9a = 10 " Ak/k/*F for the ::oderator temperature co-

~

et icient.

Comparable Cycle 3 values are 1050 ppm for the initial boron con-centrat ion, 81 ppm /l.07. ( k/i) boron reactivity wortn, and -0.57 < 10

(/k/k/*F) for the moder.ator temperature coef ficient. The FSAR shows that the core and RCS are adequately protected during this event.

Sufficient time for i

7-2 Babcock 8.Wilcox

>perator action to terminate this transient is also shown in the FSAR.

even with raxir:um dilution and minimum shutdown rargin. The predicted Cycle 3 parametr!c values of importance to the moderator dilution transient are bounded by the FSAR design values; thus, the analysis in the FSAR is valid.

7. *.. Cold Water (Pump Startup) Accident There are no check or isolation valves In the reactor coolant piping; there-fore, the classic cold water accident is not possible. However, when the re-actor is operated with one or more pu=ps tv>t running and these are turned on.

the increased flow rate will cause the average ecre temperature to decrease.

If the moder.itor temperate.c coefficient is neptive, then reactivity will 2

+

be added to the core, and a power rise will occur.

protective interlocks and procedures prevent the starting of idle pu=ps if the reactor power is above 227.

llowever, these restrictions were ignored, and t.u-pump s;artup f rom 507. power was analyzed as the most severe transient.

lo -.nimi ze reactivity addition, the FSAR analysis assumed the most negative t od4 rator t enperature coef f icient of -3.0

  • 1rf * (ik/k/*r*) and t he least neg.-

tive f*>ppler coefficient of -1.30 - 10 ".%/k.

The corresponding cost nega-t iv. roderator t emperature coef ficient and t he least negative Doppler coeffi-

.ient predicted tor Cycle 'I are -2.66 - 10 ard -1.48 = 10" (ik/k/*F) re-

~*

spec t ive l y.

Since the predict ed Cycle 3 ruderator t emperature coef ficient 1.

Icw us-gat ive and the Doppler coef ficient is more negative than the values used in the FSAR, the t ransient results would be less severe than those re-port ed la the FSAR.

7. 5.

less of Coolant Flow The reactor ciulant flow ra te decreases if one or more of the reactor coolant pimps tail.

A pumping s ailure can be causof 5y techanical failurer. or by a loss et electrical power.

With four independent pumps available, a mechani-cal failure in one pu=p will not affect the o;eration of the others. With t he react or at power t he effect of loss of coolant flow is a rapid increase in coolant temperature due to the reduction of heat removal capability. This increase could result in DNB if corrective actien was not taken irused t.ately.

The hey parameters for four-pump coastdown er a locked-rotor incident are the flow rate, flow coa stdown chcaracterist ics, the reppler coef ficient, the mod-erator teeperature coefficient, and hot channel DSB peaking factors. The

,-3 Babcock a.Wilcox

most conservative initial conditions were assumed for the densification re-port 6 - FSAR values of flow and coastdown

-1.17 a 10" (Ik/k/*F) Doppler coefficlent. 0.5 = 10 (ak/k/*F) moderator temperature coefficient, with densified fuel power spike and peaking. The results showed that the DNBR re-tained ab. ave 1.3 (W-3) for the four-pump coastdown, and the fuel eladding teriperat ure rem.nined below c riteria 1imits for the locked-retor transient.

f The predict ed paramet ric v. slues for Cycle 3 are -1.f.8 = 10 ~ (ak/k/*F) Doppler coefficient, -0.57 10 * (/k/k/*F) modere tor temperature coef fie lent,.and pcshing factors as shown in Table 6-1.

T ince the predicted Cycle 3 values are bounded by those used in the densification report. the resul t s of t hat analysis represent t he cwist severe consequences from a loss of flow incident.

l 7.d. _. _S.t ".c k -ou t,. Stuck-In,_ or Droged Cont rol Red

(

If a control rod was dropped into thu core while it was operating, a rapid decrea e in neutron power would occur accompanied by.: d ec rease in the core average c..olant temperature. The power distribution might be distorted due t o a new. ont rol rod pat tern under which conditions a return to full power l

night lead to localizeil power densit ies and heat fluxes in excess of design Iinitations.

t l'he key p.ramet ers f or this t ransient are moderator temperature coes f icient, dropped rod wort h, and local peaking factors. The FSAR analysis was b.ased on H.I.4 and 0.16% l.k/k rod worths with. moderator temperature coef ficient et - 1. 0 10" (l.k/ k /

  • F). For Cycle 3, the maximum worth rod at power is 0.20
  • k/k with a moderator temperature coefficient of -2.66 - 10 (f.k / k/
  • F ).

Since t he pred ict ed red wort h is less posit ive and the moderator t emper.ature soeffielent ruire posit ive, t he consequences of t his t ransient are less severe than the results presented in the FSAR.

I 7.7.

i.oss of F.lectric power Two types of power losses considered in the FSAR were.: loss of load cond !-

t i an c.ausni by separation of the unit f rom the transmission systs n and.:

hypothet ical condition resulting in a complete loss of' all sy stem and unit power except t lu t. from the unit batteries.

The FSAR analysis evalu.sted the loss of load with and without turbine runhack.

When there is no runhack, a reactor trip occurs 'on high reactor coolant pres-3:ure or t e=perature. This case results in a nonlimiting accident. The 7-4

- Babcock 8. Wilcox

largest of f-site dose occurs for the sceand case, i.e., loss of all electric'al power except unit batteries assuming operation with f ailed fuel and steam gen-erator tune leakage.

These results are independent of core loading; therefore, the results of the FSAR are applicable for any reload.

238 Steam Line Failure A steam line failure is defined as a rupture of any of the steas lines from the steam generators.

Upon initiation of the rupture, both steam generators to blow down causing a sudden decrease in the primary system tempera-start ture, pressure, and pressurizer level. The temperature reduction leads to positive reactivity insertion, and the reactor trips on high flux or low RC pressure.

The "SAR has identified a double ended rupture of the steam line between the steam generator and stean stop valve as the worst-case situation at EOL conditions.

The key parameter for the core response is the moderator te=perature coef fi-

cient, which was assumed in the FSAR to be -3.0 = 10 " (Lk/k/*F).

The Cycle 3 predicted value of moderator temperature coef ficient is -2.66 = 10 " (ak/k/

~

  • F).

This value is bounded by those used in the FSAR analysis; hence, the results in the FSAR represent the worst situation.

J.9 St ean Gen _erator Tube Failure A rupture or leak in a steam generator tube altos, reactor coolant and asso-ciated activity to pass to the secondary system.

The FSAR analysis is based on complete severance of a steam generator tube.

The primary concern for this incident is the potential radiological release, which is independent of c ore loading; hence, the FSAR results are applicable to this reload.

7.10._ Fue1 IL(ndIi,ng Accident The nMunical daruge accident is considered the maximum potent tal source of act ivity release during fuel handling act ivities.

The primary concern is radiological releases that are independent of core loading; therefore, the FSAR result s are applicable to all reloads.

7.11.

Rod Eject ion Accident For reactivity to be added to the core more rapidly than by uncontrolled red withdrawal. physical failure of a pressure barrier comp +>nent in the control rod drive assembly must occur.

Such a failure could cause a pressure 7_g Babcock a Wilca-

I dif ferential to act on a control rod assembly and rapidly eject the nssembly from the core region. This incident represents the ost rapid reactivity in-n.rtion that can he r?asonably postulated. The values used in the FSAR and

~

densificati,n report at BOL conditions, -1.17 a 10 I k/k/*F) Doppler coei-ficient. 0. 5 < 10" (f.k/k/*F) moderator temperature coefficient, and an ejectcJ rod worth of 0.6M ?.k/k. represent the maximum possible transient. The cor-responding Cycle 3 parametric values of -1.48 = 10 (ik/k/*F) Doppler. -0.57 10 ' (*k/k/*F) moderator temperature coef f ic ient (buch more r.egative than those used in referen6 e 6.and a maximum predicted ejected rod sorth of 0.537 "k/k ecst.re that the results will be less severe than those presented in the FSAR3 and the densificat ion report.#'

7.12.

P.atitua HfpothetIcal Aceident There is no postulated mechanism whereby this accident can occur since it

-1 would require a cultitude of failures in the engineered s.afew.ard*.

The hy-poth(tical accident is based solely on a gross release of radioact ivity to the reactor building. The consequences of this accident are independent of iore lu dini,; hence, the results reported in the FSAR are applicable for all reloads.

7.11.

4.st.

t;as Tank Rupt_u_re The wasta gas tank was assumed to contain the ga<.rous activity evolved from d gassing all of the reactor coolant following operation with 1.O' defective fuel.

Kupture o* the tank would result in the release of its radioactive

c..n t en t.4 to the plant vent ilat ion system and to the atmsphere through the unit vent.

The consequonees of this incident are independent of core loading; t here fi. r e, the results reported in the FSAR are applicable to any reload.

7_.li. J _G AnaIys11 A generic 14CA analysis for the BW 177-FA lowered-lo.cp ::SS has been perfornsd u.ing t:n final acceptance criteria ECCS eva!uation ::vdel reported in 1*M-10*03.Ib lb. anc.lys is in BAW-10103~1s generic in nature since the limit-ing va!,es of key parameters for all plants in this category were used.

Fur-thermore, the ccobination of average fuel temperature as a function of linear tu a t rate and the lifetime pin pressure data used in the F.AW-10103 f.OCA limits analysis is concervative compared to those calculated for this reload. Thus.

7-6 Babcock & Wilcox

t he analysis and the I.OCA limits reported in BA*.'-10103 provide conservative results for the operation of Oconee 2. Cycle 3 fuel.

The following table shows the bounding values for allowable LOCA peak linear heat rates for Oconee 2. Cycle 3 fuel.

Allowable peak linear Core elevatlon, fL heat ra t e._ _k'.*/ f t 2

15.5 4

16.6 6

18.0 8

17.0 10 16.0 7_7 Babcock a. Wilcox

7 Table 7-1.

Comparison of Kev Parameters for Accident Analysis

{

FSAR, densif Predicted Para-eter value Cycle 3 value l

EOL Doppler coeff, 10~5, ak/k/*F

-1.17(*}

-1.48 EAL Inppler ccef f,10', Lk/k/*F

-1.33

-1.51 BOL coderatcr coef f,10 ", Lk/k/*F

+0.5(b)

-0.57 f

EOL moderator coeff,10 ", 4k/k/*r

-3.0

-2.66

~

All rod bani vorth (HZP), % Lk/k 10.0 8.37 Initial boron cene (HFP), pps 1400 1050 Boron reactivity sorth, 70F, 75 81 ppm /l: Lk/k Nx. ejected rod worth (l!FP), I Lk/k 0.65 0.53 Dropped rod wrth (IIFP), % Lk/k O.46 0.20 1

-1.2 4 IO[;.'k/k/t was used for stean-line failure analysis.

-1.3 = 10 " '.k/k/F was used for cold water analysis.

+0.94 = 10'- ?.k/k/F was used for the craderator dilution accident.

I 7-8 Babcock a. Wilcox -

P l

t I

8.

PROPOSED MODIFICATIONS To TEciNICAL SPECIFICAIl0SS The Technical Specificatiens have been revised for Cycle 3 operation. Changes ware the results of the follcwing 1.

Specifying APSR position limits in addition to usual regulating control rod and imbalance limits for ECCS. The AISR position limits will provide additional cont rol of power peaking and assurance that LOCA kW/ft limits are not exceeded.

2.

Using 106.5% of design ficw rather than 107.6% as discussed in section 6.1.

3.

The FLAME coe:puter code 9 II used in. setting the Technical Specification limits.

4 The Technical Specificatien limits based on DNBR and LNR criteria include appropriate allowances fer projected fuel rod btw penalties (i.e., poten-tial reduction in DNHR and increase in power peaks).

5.

The penalty on core coolant ilsw due to an assutsed open vent valve has been eliminated based on a vei.t valve surveill. rice program perfor. sed dur-ing each refueling shutd.,vn.

Based on the Technical Specifications derived from the analyses presented in this report, the Final,*.cceptance Criteria ECCS limits will not be exceeded, aus will the thermal design criteria be violated. Figures 8-1 throuch 8-14 t ilust rate revisions to previous Technical Specification safety limits; Fig-8-15 through 8-17 illust rate limits not previously included in the Tech-ures 1

nical Speci fications.

l naheew a, w;s,..

1 Figure 6-1.

C. ire Protection Safety Limits.

l ocenee J. Cycle 3 2400 2200' c,

{}

ACCEPTABLE L

OPERATION E

e at 2000 UNACCEPTABLE OPERATION 5

o 1800 t

l 1600 l

l

[

560 560 600 620 640 Reactor Coolant Outlet Temperature-F i

1 l

l l

32 Babcock 1)Vilcon 4

W rigure 8-2.

Core Protection Safety Limits.

Oconce 2. Cycle 3 fatsmat Pewte Ltytt, s

-- l 2 6

(-28.3.512)

(2s.3.882) 4

-- e e 0 sw/tt Lsuit ACCEPTAILt

,,,,,

  • Pump (33. 02)

OrteATI0s

(.28.3.al.3)

'8 (2s.3,ss.3)

(-50.s0)

(l) go ACCEPfatti g33,yg,33 3 8 e PuuP 70 OPERATION

(-2s.3.5s.2) 60 (21.3.5s.2)

(-50.53.3)

$e 50 ACCEPTAILt 2.3. 4 4 Po wP

-- 40 OPitaiton

(-50.26.2)

-- 30 20 l

\\

\\

10 i

f I

t I

t I

t I

f I

-70

-60 50

-40

-30

-20

-10 0

80 20 30 40 teactor Pe.er estalance. 1 CURVE REACTOR COOLANT FLOW (GPu) 1 374880 2

280035 3

183690 4'

204310

  • THE FLUI/ FLOW 5tiPOINT FOR 2/0 PusP OPERATION NUST BE SET AT 0.949

Figure 8-3.

Core Protection Safety Limits.

Oconee 2. Cycle 3 2400 2200

.u_

E e

3 E

2000 2

3 I

2 2

O 1830 1600 I

i 1

560 580 600 620 640 Reactor Coolant Outlet Temperatute - F REACTOR CCOLANT FLOW PuuPS OPERAilhS natt (cpu)

P0eER (TYPE OF Listi) 1 374880 (1005) 1825 Four Pusp (OhBR Lime ted) 2 280035 (14. f t) 65.14 Thrce Pura (Oh8R L see tea) 3 183690 (495) 59 0",

One Pump in Each Loop 6

(Quality Limitea) l s-a Babcocka.Weico-

l

{

l'i,;u r e n,.

Pro vetl'..

Lysten *2xir.us Allew.able Se t ;u'in t a. i'Jence 2,.'vcic j 2400 T = 619*F P = 2355 usig 2300 2200 g

ACCEPTABLE g

OPERATION.

0 2100 a-b

  • M Ts 43 a

i s

UNACCEPTAS;E o

2000 2

OPERATION

[o.

C

$/

4 1900 P = 1800 psig 1800 T = 584F l

l l

l 540 560 580 600 620 640 Reactor Outlet Ter.ierature. *F o.........:...

_____-____.-m_.

Figure 8-5.

Protect Ive System.%ximum Allo.able.

Setpoints, Oconee 2. Cycle 3 TM(RMAL POWER LEVEL, 1 UNACCEPTA5LE OPERATICS 110 m

(105.5) (88.3.105.5)

(-12.3.105.5) g 100 2

g l

l (23.95)

ACCEPTABLE p

4 PUMP 90 s.

OPERATION i

i

( 7 8. 8) - 80 [

p t

I

(-40.70) 70 l I

(23,65.3)

=

CCEPTABLE l

3 & 4 PUMP

_ 60 l

OPERATION g

I (51.7)g l

$0 (44.43.3) 1 I

_ go l (23.51.2) i ACCEPTABLE 2.3 Al 4 l

30 PUMP O OPERATSON l

l

- 20

(-40.16. 2 )

"l o,

2. )

- 10 u

is,

ta g

a

,e I

f

ll ll

=

.s i

t l

l

-E0

-50

-40

-30

-20

-10 0

10 20 30 40 Pswer imbalance. T.

!E FLU 1' FLOW SETP0thi R 2<0 PullP OPERAll0N i

ST BE SET AT 0.949 i

8-6 Babcock & Wilcox

Fig ar.

H-6.

F od Posit ion Linit6 for Forsr-Pu.p Operation Fro:n O t o 100 10 IP

.l ence 2, r.. l.. i ICO - OPERATith in TMis (110.102)

(174 2.102)

,,t225 8.102)

REGION 15 hof RESTRICTED ALLDIE0 "IEIO" 574 2 92 4225 8 92).

Pf*f ER LE sEL 80 (160 2.80)

QTOFF RESIRfCTED SMUT 00em aiGION ggqgig (146 2.70s

( 253 8 Ill Lluli 50 el32 2 60s a267 B.60s EC e53 3.50s

( 118 2. 50 e PE RWISSIBL E 42I1 8.501 OPERATING t

4G E.~EIC%

.P, 13C0.37s l

2G

'U IS' (10 15:

O t

t t

i t

f g

e t

50 100 150 2C0 2$0 300 R :.: I n.:e s i tmn a.>i 25 5C 15 ICO O

22 50 15 300 I

t f

f f

9 Group 5 G cv 7 0

25 50 15 100 i

f f

f n

G'.a 6 g3 M

h 4 OOW

r

?>

Figure 8-7.

Acc Posit. ion Limits for Toar-Pu=p Operation From h

ICO : 10 to 250 t 10 ETPD, Oconee 2, Cycle 3 (185.It2s (225.8.102) 100 CPERATION IN THl3 4,

PotER LEVEL CUiOFF REGION 11 NOT ALLC9ED I

(174 2,92,

  1. 225 8.92) r RESTRICTED REGION 80 MTMfN III 2 g03 1 39 8.80) s sasGit

'N lI IIS 2 78; 253 8.TCt g

GEstaltYts 60 4tGits sI32 2.60,

'267 8.60)

=

'O l

  • 118 584
  1. 118 2.50)

=

(281 8 50s 40

,g (300 37 P E ksi SSI BL E CPERAll%G 20 155 3.15

  • the 0

0 53 1C0 150 200 250 30C Roo Ir.Je s. < OsteSraen C

15 50 75 100 C

5 50 75 100 t

Craan 5 Greue 7 0

25 50 75

'50 L

f f

1 Grcup 6 8-3 Babcock & Wilcox

+

j Ti.;ure n-8.

Rod Posit i.- : Li=it s for T ar-Pump &ternion After 250 - 10 ETPD, oco::ee 2. Cycle 3 100

'201 1.102 f251 6 1C2, P0eER CPERais;h IN th:5 (251 6 324 m

REGIO % l5 NOT ail 0eID Cul0FF to

'237 6 EDs Skul00es

  • I:3 5 10'

"*"O'"

~

BE1741CTED f.0 LI P!Gics

,2C 3 5 C i

~

.125 50)

(195 6 53e j

40 E

20 -

PE8EIS$tkt OP(RAllhG

- 11 7 15 145 15:

0" Q

f I

I g

g 0

50 100 15C 200 250 300 ROJ IS:ta Sett3raen 0

25 5e 75 tcc 0

25 50 75 100 i

f f

e 9

Gre s 5 Gr:go 7 0

25 13 75 103 t

l GrEup 5 Babcock a. Wilcon s

Figure 8-9.

Rod Posit. ion Limits for Tve-znd Tnree-Purp Operation From 0 to 100 : 13 EJPD, Oconee 2, Cycle 3 (132.2.102)

(110.102)

(160.2.102)

(233 8.182) 100

- CPERail0N IN IMIS REGION IS ggg7gggggg c

3 PJF 8EG104 FOR 3 a

- BOT ALLOWED 8tTH opguggg (146 2.69) 1253 8.83#

py

[

2 UI 3 IlII SIRlCTED DPERA110N A

30 IN THIS b

"E3" wigg,g, (132.2.76)

(267 8.766 SwJTDCsh Llull U

\\

00 7,

~

I

~

iS3 3.50)

RESTRICTED REGION FOR PERulSSIELE iX2 47) 213 PUMP OPERail0N

=

40 CPERATING

=_

s REGION a

6 I

2*

20 (0.15) 0 i

e i

t t

0 50 100 150 2:3 250 XI Roa inses.

Watrisraea 0

25 50 75 500 0

25 50 75 133 e

t i

i t

t Group 5 Group I 0

25 50 15 100 t

I f

f Group 6 I

i 8-10 Babcock & Wilcox

l 1

.Igure 8-10.

Rod Position Limits for Two-and Three-Pu=p operation From 100 1 10 to 250 - 10 EFPD, Oconee 2, Cycle 3 l

(185 102)

(239.8.102) 100 GPERATION IN THl1 REIfflCTED E

REGION 15 h0i Att00E0 ftEGitin FDR O

t 253. 8. 89 i 3 P' EP 2

CPE4&T10h

'k' M

SduTO0eN --

(267 8.76)

E' NARGth 2

~

LINii kl 60 A

E (110.50 y

'369 47)

A 40 PErstiSISLE OPERAllNG f

~

REGIch 20

.53 8.15) 0 i

f 0

50 100 150 200 250 300 500 inass, i Wetration 0

25 50 75 100 0

25 50 75 100 I

t t

9 Groua 5 Group 7 0

25 50 75 100 I

f I

f g

Grcup 6

-l 4

\\

l l

8-11 Babcock & Wilcox w

Figure 8-11.

Rod Position L1=its for Two-and Three-Pump Operation After 250 : 10 EFPD, Oconee 2, Cycle 3

'00 OPERATION IN THl3 REGIC% IS NOT RESTRICTED REGlch 3 p.,

(223 6 871 3

ALLC9ED fliH e

FOR 2 1 3 PUNP ggggggy 2 GR 3 PUePS CPERATION RESUICTED l

g

ED IN THIS MGION (2C3 6. 7Es a

[=

SHUT 00s4 MARGIN Lluti u

" EO a

m 3

(125.50)

PERMIS$1BLE O'ERailhG REGIGt

~g

,2 40

'.o

-% 20

=

471 7.15) 0 i

0 50 100 150 200 250 30 0 Rao inces. i 8stn2raen 0

25 50 75 100 0

25 50 75 ICO Grcup 5 Grsup 7 0

25 50 15 100 I

f f

f g

Group 6 1

I l

Babcock s. Wilcox 8-12

1 I

Figure 8-12.

Operational Pz.er Icba'a::ce Envelope for operation Frc= 0 to 100 : 10 ETPD, oconce 2, cyc:e 3 Poser, 5 of 2568 litt 8

RESTRICTED REGION

(-18.A 102)

(+14.2.102) 100<-

( - 17. 5. 92 )

(+13.4.92)

90..

80-.

70..

60..

50..

40 -.

30 - -

20 --

10 -

0 l

-20 0

0

+10

+20 Core isoalance, ?

3-13 Babcock a. Welcox

.i

1 l

Figure 8-13.

uperational Power Ir. balance Envelope for Operation From 1001 10 to 250 : 10 EFPD, Oconee 2 Cycle 3 Power 5 os 2562 Mft RESTRICTED REGION

(-22.3.102)

( +14. 2.10 2) 100

(-20.6.92)

(+13.4.92) 90 --

80 --

70 SC --

50 --

40 --

30 20 --

10 0

l 30

-20

-0 0

+10

+20

+30 Core imbalance. ",

8-14 Babcock & Wilcox I

l'i gu r e 8-14.

Operational Pov.cr != balance Envelope for Operation Af ter 2: 3 - 10 EFPD, Oconee 2. Cycle 3 Power. 5 of 2568 n t RESTRICTED REGION

( 28 8.102)

_ ( +11.1.102)

(-26.3.92)

. ( +10. 7. 92 )

90 80 -

70.

60 -

50.

40.

30 20 10.

y a

e i

a n

-30

-20

-10 0

+10

+20

+30 l

Core lanalance. 5

\\

Figure 8-15.

APSR Position Limits for Operation From 0 to 100 : 10 EFPD Oeence 2. Cy-le T 100 (7.6.102)

(32.6.102)

(5.2,92)

(35.0,92)

RESTRICTED 90 REGION B0

- ( 2. 8. 80 )

(35.0.80)

RESTRICTED REGION 70 (0.70)

(42.70)

(86.60)

= 60

=

3 (100,60)

_ 50 O

f. 40 2

PERMISSIBLE 30 OPERATING REGION 20 10 0

0 10 20 30 40 50 60 70 80 90 100 APSR. ". Witnaraan 1

n-...........

l l

Fi,;ure d-16.

Al'SR Posit len Litilts f or Operat ioc Af ter 250 - 10 EFPD. Oconee 2. Cycle 3 RESTRICTF0 REGION 100 (6.5.102)

(31.4.102)

RESTRICTED REGick 90 (4.1.92)

(33.8.92) 80 (3.7.80)

(33.8.80) 70 (0.70)

(40.8.70)

E (54 8.60) 60 E

N (100.60) 50 40 PERNISSIBLE 6

E OPERATING 30 REGION 20 10 0

0 10 20 30 4 f'.

50 60 70 80 90 100 APSR, 5 mitnarann 0lPMw

.g.

l

\\

Figure 8-17.

APSR Position Lir.its for Operation Fron 100 10 to 250 : 23 E}7D, oconee 2, Cycle I, RESTRICTED REGION

.f(7.6.102)

(34.1.102) 100 f

RESTRICTED 90 (5.2.92)

(36.5,92)

REGION 80

( 2. 8,80 )

(36.5.80)

(0.10)

(43.5,70) 70

~

(87.5.60)

="

60 3E (100.60) 50 f

40 S

IBLE 30 OPERATING REGION 20 10 l

0 i

3 0

10 20 30 40 50 60 70 80 90 100 APSR, '. Withdraen

'a Rahrncle a. wilen,

it i

9.

STARTt;p PROGlWI The planned startup testing associated with core perforn.ince are provided below.

These te.ts verity that core performance is within the as.,umption. of the afety analysi. and provide the necessary data for continued s.ifig plant operation.

pre-Critical Tests 1.

Control re.d drive trip time testing.

Ze ro power Te.ts 1.

Critical boren concentration.

2.

Temperature react ivi ty coet fic ient.

~). Control rod group worth.

l 4.

Ej ec t ed rod wort h.

l P.wer Tests 1.

Core power distribution verification at approxim.itely 40 75 and 100.: FP normal control rod group configuration.

2.

Incore/out-of-core det ce t er imbalance correlation verification at approxi-

a t e I y 7 57. Fp.

l.

power doppler reactivity coefficient at.epproximately 100; rp.

4.

Temperature reactivity coetficient at approximately 100% FP.

9.g Babcock 8. Wilcox

A.

o.-

10.

REFERENCES I

Oconee Nuc lear Stat ion, l' nits 1, 2, and 3, Final Safety Analys_is_ Report.

Docket Nos. 50-269. -270, and -287.

2 Oconee 2 Cycle 2 Relo.ad Report, _BAW-1425, Rev. 1, Babcock & Wilcox. Lynch-burg, Virginia, April 1976.

R.

V. De mns and R. R. Steinke, Fuel Assembly Stress and Deflection Analy-sis for Loss-of-Coolant Accident and Seismic Excitation. EAW-10035 Bab-cock & Wilcox, I.ynctiburg, Vi rginia,. lune 1970.

R. C.

Childress, J.

.I. Woods, and T. N. Ake, Irradiation of Two 17 a 17 Deraonstrat ion Assemblies in Oconee 2. Cycle 2 Reload Report, BAW-_1424, Bab-cock & Wilrox, I.ync hbu r g, Virginia, 1976.

Progract to Determine in-Reactor Performance of B&W Fuels - Cladding Creep Collapse, P~W-100M4 P, Rev. 1. Babcock & Wilcox. Lynchburg, Virginia Octo-ber, 1976 A.

1. Ecke rt, 11. W. Wilson, and Y. E. Yoon Oconee 2 Tuel Denal fication Report, B.W-1195, Babcock & Wilcox. Lynchburg, Virginia, June 1971.

7 C.

D. Nrgan and 11. S.

Kao, TAFY - Fuel Pin Temperature and Gas Pressure Analysis, I g,10044 B.abcock & Wilcox, Lynchburg, Virginia, m y 1972.

B. J. Bueseher and J. W. Pegram. Babcock & Wilcox Ndel for Pre fict ing In-Reactor Densification, BAW-1003 3P, Rev. 1, Babcock & Wilcox. Lynchburg, Virginia, November 1976.

9 l~l M !E - Thr(e-Ilimensional Noded Code for Calculat ing Reactivi t y and Power nist ribut ions, BAW-lG124A Babcock & Wilcox, 1.ynchburg, Vire,ini.i. August 1976.

55 C. W. Nys, Verification of Three-Dimensional FIAME Code, EAV-10125, Bab-cock & Wilcox, I.ynchbu rg, Virginia August 1976.

to_j Babcock & Wilcox

)

or 11 Oconce 1. Cycle 4 Reload Report, BAW-1447, Babcock & Wilcox, Lynchburg, Virginia, March 1977.

12 Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water, BAV-10000A, Babcock & Wilcox, Lynchburg, Virginia, June 1976.

II K. E. Suhrke to A. Schwencer Letter, "B&W Operating Experience of Reactor Internals Vent Va lve s, " Au gus t 4, 1975.

I' K. V. 1:111 F. E. Mately F. F. Cadek, J. E. Casterline, Ef fects on Criti-cal liest Flux of Local lient Flux Spike or Local Flow Blockage in Pressurized Vater Reactor Rod Sundles, 74-WA/IIT-54, ASME Winter Annual Meeting, New York, November 1974 l

" Critical Heat Flux-Critical Heat Flux Correlation for CE FA With Standard Spacer Grid - Part 2, Non-Chiform Axial Power Distribution," CENPD-207 Combustion Engineering, Jure 1976, IC Core Physics Methods Data l' sed as Input to LOCA Analysis, XN-75-43, August 1975; and letter, D. A. Bixel, Consumers Power to R. A.

Purple, April 5, 1975.

17 A. Schwencer to K. E. Suhrke, NRC 1.etter, November 1975.

t' ECCS Analysis of B&W's 177-FA L& ere ! Loop NSS, BAV-10103, Babcock & Wil-cox, Lynchburg, Virgiaia, June 1975.

W.

O. Parker to B. C. Rusche, Letter, Duke Power Co., Feb r ua ry 27, 1976.

i n_.,

Babcock & Wilcox

END l

llHH0T0GRAPHER_zr,-

l DATEu=n____

g%

fi{ 4 9

'fi MICROFILM SECTION a L su 1s7 2 na was i N NAvv vano WASHING TON. D C, 20374

_ _ _ _ _ _ _ _ _ _

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