ML19308B544

Cycle 5 Reload Rept.

ML19308B544
Person / Time
Site:	Oconee
Issue date:	03/31/1979
From:	BABCOCK & WILCOX CO.
To:
References
BAW-1502, NUDOCS 8001090539
Download: ML19308B544 (50)

Text

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P.AU-1522 Mare,h 1979 {

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5 0CosmE UNIT 3. CTCLE 5

- Raioad Report -

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0 N * & VILCOX Power Generation Group Muclear Power Ceneration Diviston P. O. Box 1260 Lynchburg, Virginia 24505 B&bCock & Wikcx  !

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'I CotrrENTS '

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Page

1. DrrBoDUCtION . ........................... 1-1

2. OPERATING HISTORT ......................... 2-1

3. CENERAL DESCRIPTION

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

ITEL S T:T*Di DES ICN . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.1. Puel Assembly Markante*1 Design ............... 4-1 s 1

4.2. Fuel Rod Design ....................... 4-1 4.2.1. Cladding Collapse 4.2.2. Cladding Stress

.................. 4-1 ,

................... 4-1 6 4.2.3. Cindding Strain ............... 4-2 4.3. Thermal Design . . . . . . . . . . . . . . . . . . . . .... ... . 4-2 l

4.4. Material Design ....................... 4-2 l 4.5. Operating Experience . . . . . . . . . . . . . ........ 4-2

5. NUCLEAR DESIGN . . . . . . . .................... 5-1 5.1. Physics Characteristics '

5.2. Analytical Input .

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

,j 5.3. Changes in Nuclear De. sign

.................... .. 5-2

.................. 5-2

)) 6.

THERMAL-HTDRAULIC DESIGN . . . ................. .. 6-1 3

d 7. ACCIDDrf AND TRANSIENT ANALYSIS .................. 7-1

) 7.1. General Safety Analysis ............. ...... 7-1 g 7.2. Accident Evaluation ................... .. 7-1 i S.

r PROPOSED N001FICATIONS TO TZCENICAL SPECIFICATIONS . . . . . . . .. 8-1

9. STARTUP PROCRAM - PHYSICS TESTING ................. 9-1 9.1. Precritical Tests .................... .. 9-1 9.1.1. Control Rod Trip Test 9.2.

................ 9-1 Zero Power Physics Tests . . . . . . . . . . ......... 9-1 9.2.1. Critical Boron concentration . . . . .

........ 9-1 9.2.2. Temperature Reactivity Coefficient

.......... 9-2 9.2.3. Control Rod Group Reactivity Ucrth . ......... 9-2 9.2.4 Ejected Control Rod Reactivity E rth . . . . . . . .. 9-3 -

9.3. Power Escalation Tests . . . . . . . . . . . . . . . . . . . . 9-3 9.3.1. Core Power Distribution verification at 4 0, 75, and 100% TP Vith Nominal Control Rod Position .... 9-3

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Page 9.3.2. Incore Vs Excore Detector Imbalance Correlation Verification at $40% FP ............... 9-5 9.3.3. Temperature Reactivity Coefficient at $1002 FP . . . . 9-5 9.3.4. Power Do pler Reactivity Coefficient at S1001 FP . .. 9-5 .

9.4. Procedure for Use When uceptance Criteria Are Not Met .... 9-5 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1 '

i Lisc of Tables Table 4-1.

I 4-2.

5-1.

Fuel Design Parameters and Dimensions Feel Ihermal Analysis Farameters . . . .

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

............ 4-5 Oconee 3 Physics Parametsrs - Cyclas 4 and 5 . . . . . . . . . . 5-3 5-2. Shutdown Margin Calculation for Oconea 3 Cycle 5 . . . . . . . . 5-5 I 6-1.

7-1.

7-2.

Maximum Design Conditions for Cycles 4 and 5 . . . . . . . . . . 6-2 Comparison of Key Parameters for Accident Analysis . . ..... 7-3 Bounding Values for Allowable IDCA Peak Linear Heat Eates ... 7-3 I

List of Figures Figure 3-1. Core Loading Diagram for Oconee 3. Cycle 5 . . . . . . . . . . .

3-2. 3-3 Enrichment and Burnup Distribution for Oconee 3. Cycle 5 . . . . 3-4 3i 3-3. Control Rod Locations for Oconee 3. Cycle 5 .......... 3-5 5-1. BOC (4 EFFD) Cycle 5 Two-Dimensional Relative Power Distribu-tion - Full. Power. Equilibrium Xenon. APSRs Inserted ..... 5-6 B 8-1. Core Protection Safety Limits. Unit 3 ............. 8-2 8-2. Protective System Maximum Allowable Setpoints. Unit 3 .....

8-3. 8-3 Rod Positions Limits for Four-Pump Operation From 0 to 100 2 I 8-4.

10 ETPD - Oconee 3. Cycle. 5 ..................

Rod Position Limits for Four-Pump Operation From 100 2 10 to 8-4 200 1 10 EFPD - Oconee 3. Cycle 5 ........... .... 8-5 I 8-5.

8-6.

Rod Position Limits for Four-Pump Operation After 200 2 10 ETPD - Ocone e 3. Cyc l e 5 . . . . . . . . . . . . . . . . . . . .

Rod Position Limits for Two- and Three-Pump Operation From 8-6 I O to 100 1 10 EFFD - Oconee 3. Cycle 5 . . . . . . . . . . . . . i 8-7. 8-7 Rod Positten Limits ter Wo- and Three-Pump Operation From 100 2 10 to 200 2 10 ETPD - Oconee 3. Cycle 5 ......... 8-8 4

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Figure Page 8-8. Rod Position Limits for Tnso- and Three-Fuay Operation After ,

200 1 10 ETFD - Oconee 3. Cycle 5 . . . . . . . . . . . . . . . 3-9 8-9. Operational Power Imb41anca Envelope for Operation From 0 to l 100 1 10 EFFD - Oconee 3. Cycle 5 . . . . . . . . . . . . . . . 8-10 - ll 8-10. Operational Power Inhalance Envelope for Operation From 100 1 10 to 200 1 10 IFFD - Oconee 3. Cycla 5 . . . . . . . . . . . 8-11

'c,3 8-11. Operational Power Inhalance Ecvelope for Operation Af ter 200 1 10 EFFD - Oconee 3. Cycle 5 . . . . . . . . . . . . . . . . . 8-12 8-12. AFSR Fomition Limits for Operation From 0 to 100 2 10 EFFD - '~

Oconee 3. Cycle 5 . . . . . . . . . . . . . . . . . . . . . . . 8-13 l 8-13. AFSR Fosition Limite for Operation From 100 1 10 to 200 1 10 L EF7D - Ocenea 3. Cycle 5 ............. . .. .. .

8-14. APSE Fosition Limits for Operation After 200 1 10 EFFD -

Pe-14 j Oconee 3. Cycle 5 . . . . . . . . . . . . . . . . . . . . . . . 8-15

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1. DrTEODUCTION AND St30Wtf 4 5

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This report justifies the operation of the fif th cycle of Oconee Nuclear Sta-tion Unit 3. at the rated core power of 2568 W t. Included are the required '

analyses as outlined in the USNRC document " Guidance for Proposed License Amenoments Relating to Refueling." June 1975. -

To suppare cycle 5 operation of Oconee Unit 3 this report employs analytical h techniques and design bases established in reports that were previously sub- ,

mitted and accepted by the USNBC and its predecessor (see references).

A brief summary of cycle 4 and 5 reactor parameters related to power capability is included in section 5 of this report. All of the accidents analyzed in the FSARI have been reviewed for cycle 5 operation. In those cases where cycle 5 characteristics were conservative compared to those analyzed for previous cy-cles, no new accident analyses were perfo-med.

The Technical Specifications have been reviewed, and the modifications required for cycle 5 operation are justified 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 Cool-ing Systema, it has been concluded that Oconee Unit 3 can be operated safely for cycle 5 at the rated power level of 2568 wt.

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N reference fuel cycle fcr the nuclear and thermal-hydraulic analyses of .

Oconee Dmit 3 is the currently operating cycle 4. Cycle 1 was terufnated ~f 'i

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after 159 EFFD of operation. Cycle 4 achieved initial criticality on July

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13, 1978, and power escalation e-ed on July 15, 1978. m 100% power if level of 2568 MWt was reached on July 24, 1978. N fuel cycle des 18m lan8th

.\[y-for cycle 5 - 292 EFFD - is based on a planned cycle 4 length of 277 EFFD. ' l'Q t, ,r...

No operating anomalias cccurred during cycle 4 operation that would adversely <1 %

affect fasi performance in cycle 5. ~

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3. CENEgAL DESC3IFTION I

The Oconee Unit 3 reactor core is described 14 detail in Chapter 3 of the FSA1.1 The cycle 5 core consists of 177 fuel assembV s. each of which is a 15 by 15 array containing 208 fuel rods.16 control :od guide tubes, and one incere instrument guide tube. The fuel rod cladding is cold-worked Ziresloy-4 with an OD of 0.430 inch and a vall thickness of 0.0265 inch. The fuel con-sista of dished end, cylindrical pellets of uranium dioxide which are 0.3695 loch in diameter. (See Table 4-1 for additional data.) All fuel assemblies in cycle 5 maintain a constant nominal fuel loading of 463.6 kg of uranium.

The undensified nominal active fuel lengths and theoretical densities vary be-tween batches, however; these values are given in Table 4-1.

Figure 3-1 is the core loadin*, diagram for Oconee 3. cycle 5. Thirty-six of the batch 4 assemblies will be discharged at the end of cycle 4. Nine once-burned batch 1 assemblies. with an initial enrichment of 2.01 wt Z 23N,vggi be reloaded into the central portion of the core. The remaining 16 batch 4 assemblies, designated "4C." and batches 45. 5. and 6 with initial enrich-ments of 2.53, 2.64. 3.02, and 2.97 wt I 23 %. respectively will be shuffled to new locations. Batch 7, with en initial enrichment of 2.80 wt % 23%, ,gtg cccupy primarily the core periphery. Figure 3-2 is an eighth-core map showing the assembly burnup and enrichment distribution at the beginning of cycle 5.

Cycle 5 will operate in a rods-out, feed-and-bleed mode. Core reactivity con-trol is supplied mainly by soluble boron and supplemented by 61 full-length Ag-In-Cd control rod assemblies. In addition to the full-length control rods, i eight partial-length axial power shapi'as rods (AFSRs) are provided for addi-tional control of axial power distribution. The cycle 5 locations of the 69 -

control rods and the group designations are indicated in Figure 3-3. The core locations of the 69 control rods for cycle S are identical to those of the reference cycle indicated in Chapter 4. However, the group designationa l dif fer between cycle 5 and the reference cycle to minimize power peaking.

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,"4-The nominal system pressure is 2200 pais, aos the densified nominal bear. rate '

is 5.60 W/ft at the rated core power of 2568 Mut. . I

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• 7 7 y S 6 S 6 5 Ey 7 7 j B7 L2 CIS C13 F10 C3 C1- L14 C16 D 7 7 4C S 6 5 45 5 6' S 4C 7 7 CFl 313 El M2 ES 03 K11 K14 AS 56 C71 E Cll 7

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"- C2 E12 F6 013 FS K10 071 E6 L11 C3 L10 N4 K14 7 48 4C S 6 4C S N 5 4C 6 4a 5 4C 7

5) 34 07 M9 R10 G9 78 G7 34 M7 C1 512 M1) 7 5 6 S S 6 S S S 6 S S 6 5 7 M4 47 til L5 F1 M6 r13 nao 75 49 M12 L 7 7 5 6 6 4C 6 4C 6 4C 6 6 S - 7 7 FIO Ele 7

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1 2 3 4 5 6 7 8 9 to 11 12 13 16 15 location of regenerative neutroo aource assemblies and retainer assemblies '

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2.01 3.02 2.53 2.97 2.64 3.02 2.53 2.80 M

14.215 13.306 21,222 7,47a 19,724 18,362 19,698 0 3.02 2.97 3.02 3.02 2.97 3.02 2.80 E

15,811 6.649 15.590 16,461 6,747 13,866 0 2.53 2.97 2.97 3.02 2.80 2.50 g L g i

21,255 9,165 7,472 16.344 0 0 3.02 3.02 2.01 2.80 M s 10,210 14.217 17.647 0 -

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4 FUEL $157d( DESIGN 4.1. Puel Assembly Mechanical > fen N types of fuel assemblies and pertinent fusi design parameters for Oconee

3. cycle 5 are listed in Table 4-1. Two retainer assemblies will be used on two fuel assemblies containing rege6,ees:1ve neutron sources. The justification .

for the design and use of the retainers described in reference 3 is applicable to the RMS retainers used in Oconee 3. cycle 5. All fuel assemblies are iden-tical in concept and are mechantes11y laterchangeable.

4.2. Fuel Red Design The mechanical evaluation of the fuel rod is discussed below.

4.2.1. Claddina collapse The fuel of batches 4 and 5 is more limitina than other batches due to its longer _ revious incore exposure time. The batch 4 and 5 assembly power his-g tories were analyzed, and the most limiting assembly was used to perform the creep collapse aaalysis using the CROV computer code and proceduras described in topical report RAW-10084P-A. Rev. 2.2 The collapse time for the most lim-iting assembly was censervatively determined to be more than 30.000 EFFH. which is greater than the maximum projected residence time of cycle 5 fuel (Table 4-1).

4.2.2. Claddinz Stress The oconee 3. cycle 5 stress parmeters are enveloped by a conservative fu=1 rod stress analysis. For design evaluation, the primary membrane stress must be less the 2/3 of the minimum specified unirradiated yield strength, and all stresses must be less cl.an the minimum specified unirradiates, yield strength.

In all cases, the margin is in excess of 301. The following conservatisms with re:pect to Oconee fuel were used in the analysis:

1. A lower post-densification ti.tsrnal pressure. '

2. A lower initial pellet density.

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4. 2. 3. Cladding Strain

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>N % fuel design critaria specify a 1f ait of 1.0% ce cladding circumferential 3

e plast'ic strain. n e pellet design is such that the plastic cladding strain l

is'less than 1% at 55,000 mwd /mtU. N following cladding strain conservatisas

.Jare applicable with respect to the Oconee 3 fuel:

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m h assimo Specificacica value for the fuel pe13et density was used.

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- . !p 3. .The ' cladding ID used was the lowest permitted Specification tolerance. ,

l.bh' 4.. The =artmum expected three-cycle local pellet burnup is less than j 4, [$jfc4 ~(55,000 mwd /etU. l

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' All fuel sasemblies in this core are thermany similme. h fresh batch 7 1

~.., fuel inserted for cycis 5 operation introduces no significant differences in K

fw'. thermal performance relative'to the other fuel remaining in the core. }

((, h design minimum linear heat rate (IER) capability and the average fuel ten-f Q parature for each batch in cycle 5 are shown in Table 4-2. LUR capabilities

@ are based on centerline fuel melt and were established using the TATY-34 code h with fuci densification to 96.5% of theoretical density.

Y 4.4 Material Design '

D O h batch 7 fuel assembliss are not new in concept. nor do they utilize dif- 4 h ,.

.a -s ferent component materials. Therefore, the chemical compatibility of all I j' possible fuel-cladding-coolant-assembly interactices for the batch 7 fuel as-1 3

-o semblies is identical to those of the present fuel.  ;

rf 4.5. Operating Experience i

e' A Babcock & Wilcox operating experience with the Mark B 15 = 15 fuel assembly M

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has verified the adequacy of its design. As of November 30, 1978, the follow- }

y 2ng experience has been accumulated for the nine operating E4W 177-fuel assen-(; bly plants using the Mark 8 fuel assembly: j

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l burnup. mwd /atU a) Cumulative net m3 J. Current electrical put. J p Reactor cycle Incore Discharged NWh b ,.

a Oconee 1 5 31.900 31.103 23.108.841 ./," 9 Oconee 2 3 lo.800 33.700 19.165.343 -

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oconee 3 4 20.900 29.400 '20.224.866  ; ~' 4, r THI-1 -4 29.430 32.200 21.856.837 T'

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'h ANO-1 3 30.835 28.300 18.145.982

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,No. of FAs ' '

.9 16 . 4 56 36 56

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' Fue1 rod OD. in.

0.430 0.430 0.430 0.430 0.430 0.430 1 g

Fuel rod ID. in. 0.377 0.377 0.377 0.377 0.377 0.377 g Flex spacers, type

• Spring Spring Spring. Spring fipring Spring i Rigid spacers, type Zr-4 Zr-4 Zr-4 Zr-4 Zr-4 Ir-4 Undensif active fuel 141.0 142.2 142.2 142.2 142.2 142.2 1 :

length, in.

g g Fuel pellet 00 (mean 0.3682 0.3695 0.3695 0.3695 0.3695 0.3695 g

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Fuel pellet initial 95.3 94.0 94.0 94.0 94.0 94.0 density (mean spec).

1TD 1 Initial fuel enrich- 2.01 2.53 2.64 3.01 2.97 2.80 ment, wt. % 23 g Avg 50C burnup. 17.266 20.468 13.724 15.406 7.504 0 3 F".Jd /atU g Est residence 25.488 24.403 24.808 24.405 27.672 28.032 time. EFFIt Cladding collapse >30.000 >30.000 >30.000 >30.000 >30.000 >30.000 time. EFFil . l I

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- . .,r Batch No. h. .I 1D &c/At 5 6 7

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. v. g No. of assemblies 9 16/4 56 36 -56 ', , ' '. ';'

L Initial density. 1 TD 95.3 94.0 94.0 94.0 94.0 ' zk

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Fellet diameter, in. 0.3682 0.3695 0.3695 0.3695 0.3695 ' M, Stack height, in. 141.0 142.2 142.2 142.2 142.2 -

i a Densified Fuel Farameters " .5 Fellet diameter, in. 0.3649 0.3646 0.3646 0.3646 0.3646 -g Feel stack height in. 140.2 140.5 140.5 140.5 140.5 _1)

Nominal linear heat 5.80 5.80 5.80 5.80 ~ 5.80 .

rate 9 2568 MWt kW/ft ..

-' l Average fuel temp 9 -

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Pa=iaal LHR F 1310 1320 1320 1320 1320 . >

I t.famar heat rate cape-bility (centerline fuel melt). kW/ft 20.15 20.15 20.15 20.15 20.15 d

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p 1 -

i U

5.

NUCLEAR DESIGN  !

b f ,

_5.1. Physics Characteristics t j Table 5-1 compares the core physics parameters of design cycles 4 and 5; the values for both cycles were generated using FDQ07.6.7.s since the core has not yet reached an equilibrium cycle differences in core physics parameters are to be expected between the cycles. The longer cycle 5 vill produce a larger cycle differential burnup than that for the design cycle 4. The average EOC i

core burnur will be higher at EOC 5 than at EOC 4 because of the presence of the twice-burned batch I and the fourth burn of batches 45 and 4C fuel. Fig- -

ure 5-1 illustrates a representative relative power distribution for the be-I ginning of the fif th cycle at full power with equilibrium zenon and APSRs in- ~

serted.

The critical boron concentrations for cycle 5 are higher than for cycle 4 be- --

cause of the larger feed batch and a different radial power distribution.

The group 7 tod worths are significantly higher in cycle 5 since group 7 con-tains 12 CRAs rather than 8 as in cycle 4. As indicated in Table 5-2. the .

j control rod worths are sufficient to maintain the required shutdown margin. ,

The ejected rod worths in Table 5-1 are the maximum calculated values. Cal-colated ejected rod worths and their adherence to criteria are considered at _!

all times in life and at all power levels in the development of the rod in-sertion limits presented in section 8. The sef w= stuck rod worth at EOC 5 I is less than that et EOC 4. The adequacy of the shutdown margin with cycle 5 stuck rod worths is demonstrated in Table 5.2. The following conservatisas were applied for the shutdown calculations:

1. Poison material depletion allowance.

1 2. 10% uncertainty on net rod vor th. -

3. Fluz redistribution penalty. -"

Flux redistribution was accounted for since the shutdown analysis was calcu-lated using a two-dimensional model. The shutdown calculation at the end of  ;

5-1 MM&NK j

.. .. .. --l  ;

N

I i -

cycle 5 was analysed at 292 EFPD. h reference fuel cycle shutdown margin.

is presented in reference 5. Table 5-2. -

i h cycle 5 power deficits from hot zero power to hot full power are similar to thuse for cycle 4. Doppler coefficients, moderator coefficients, and xenon i vorths are also similar for the twn cycles. h differential baron worths for cycle 5 are approximately equal to those for cycle 4. h effective delayed j neutron f ractions for both cycles show a decrease with burnup.

5.2. Analytical Input 1 The constants used to compute core power distributions from incere detector measurements were obtained in the same manner for cycle 5 as for the reference '

cycle 4.

5.3. Changes in_ helear Design I Cycle 5 is destgoed to operate in a feed-and-bleed mode in contrast to the rod-ded operation of cycles 1, 2, 3 and 4. The major difference in operational modes during equilibrium steady-state conditions is that no full-length control rods are inserted into the core. (A small bite insertion - approximately 101 -

of one regulating bank is maintained to allow discrete changes in soluble boron and to accomodate small temperature and load d m nd changes.) During load-fol-low operation, the regulating bank is inserted into the core only to offset power Doppler reactivity changes. Transient sanon reactivity effects are com-pensated by changing the solubla boron concentration.

The same calculational methods and design information were used to obtain the important nuclear design parameters for cycles 4 and 5. As in cycle 4, both APSRA and CIA position limits, as well as power imbalance limits, will be specified based on 14CA analyses. These operational limits and the IPS limits (Technical Specification changes) for cycle 5 are presented in section 8. 3 A fuel melt limit cf 20.15 kW/f t has been employed in calculating the rasetor protection system setpoints and is the same as in previous cycles. Two batch 5 assemblies have been assigned a maximum linear power rating of 19.74 kW/f t based on as-built data. These assemblies will be placed in non-limiting loca-

.tions during their entire core residence.

I I

bM&Ng 5-2 E

.. ~ _m . m ...,.__._m.._........_m.. _ . . . _ . _._._.m__ . . . . ._..__W

'd , I Table 5-1. Oconee 3 Physics Parameters I ") - Cycles 4 and 5 Cycle & Cycle 5 Cycle length. EFFD 270 292 ,

Cycle burnup. mwd /stU 8450 9137 Average core burnup. EOC, mwd /stU 17.233 18.711 laitial core loading, atU 82.1 82.1 Critical boron - BOC (no xenon), ppe NZP. group 8 wd 37.51 1250 1351 BZP. groups 7 and 8 inserted 1188 EPP. group 8 wd 37.5% 1170(C) 978 1161 Critical boron - EOC (equil xenon).ppe HIP group 8 37.5% wd 304 339 HFF. group 8 37.5% wd 41 61 Control rod worths - HFP. BOC. % Ak/k Croup 6 1.00 1.00 Croup 7 0.87 1.70 Croup 8 37.5% wd 0.35 0.49 Control rod worths - HFP. EOC(d) % Ak/k Croup 7 1.12 1.64 Croup 8 37.5I vd 0.40 0.51 Max ejected rod worth - HZP. I Ak/k BOC. grou;s 541 inserted 0.44 0.46 EOC. groups 5-8 inserted 0.50 0.50 Max stuck rod worth - 1:ZP. I Ak/k BOC 1.8G 1.81 EOC 1.85 1.75 Power deficit. HZP to HFP. % Ak/k BOC t.30 1.34 EOC 2.12 2.11 t

Doppler coeff - BOC. 40-5(Ak/k *P) 100% power (no menon) -1.48 -1.51 Doppler coeff - EOC,10-5(Ak/k *F) 100% power (equil menon) -1.61 -1.57

~

5-3 Eabcock & Wilcox

~

ar .- . . - ..

1. , ..s<.c..  ; v.

l s , _ , .

,.v. '

T -

.,t ., , , , , >

' ~ ~ - ~ *

, q : -q .

Table 5-1. (Cont'd) - .

Cycle AI ") Cycle 5' Moderator coeff - ITF. 104 (Ak/k *F)

BOC (0 sence.1161 ppe, group 8 ins.) -0.70 -0.66 5 20C (equil menos 17 ppe, group 8 ins.) -2.57 -2.69 ,3 Soron wort.*t - EFF. pya/1 Ak/k BOC (1150 ppe) -106 109 ./

EOC (17 pys) .95 95 )

Xenon worth - EFF. I Ak/k 50C (4 days) 2.64 2.65 20C (equilibrium) ~2.75 2.75  :

^

Eff delayed neutros fraction - RFP ^

'N 30C , 0.00583 0.00585 -

BOC 0.00522 0.00519 I"' Cycle 5 data are for the conditions stated in this report. The cple 4 core conditicos are identified in reference 5. I, (b)aased on 169-EFFD cycle 3.

(c)Ecros concentratica for cycle 4 HFP conditioes, banks 7 and 8 inserted.

(d)235 EFFD in cycle 4, 272 ZYPD in cycle 5.

I I

I I

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. I

~; , ~,~ .

~:_. _ _ a .. - -___,;.____ . .x. a. .

._ _f

. Table 5-2. shutdova Mars.tn calculation for occeea 3

.3 ,

Cycle 5 ,

BOC. EOC.

t ti/k I ak/k Arailable Rod Worth Total rod worth. RZF 8.45 8.30 Worth red'n due to poison bu*nup -0.42 -0.42 Maxf === stuck rod. HZF -1.81 -1.75 Net worth 6.22 6.13 less 10% uncertainty -0.62 -0.61 Total available worth 5.60 5.52 Required Rod Worth Power deficit. EFF to MIP 1.34 2.11 Max allowable inserted rod worth 0.52 0.52 Flux redistribution 0.53 1.03 Total required worth 2.39 3.66 Shutd e Martin Total avail worth - total req'd 3.21 1.86 worth  !

Notes Required shutdown margin = 1.CCI ik/k.

5-5 Babcock .WI cox

_ _ _ . _ . _. _ ._ _ . , . _w

b

~ !,

Figure 5-1. BOC (4 EFFD) Cycle 5 Two-Dimensional Relative Tower l Distribution - Full Power. Equilibrium Teon. AFS,Es  ;

Inserted -

8- 9 10 11 12 13 14 15 I 5 0.80 1.00 0.98 1.30 1.01 1.03 0.81 0.78 E 1.11 1.33 1.20 1.11 I

1.23 1.04 0.80 i 8 31 -

L 1.04 1.29 1.11 1.04 1.15 0.66 m g- e M 1.25 1.02 0.73 0.86 I

E 0.81 0.97 0.59 I} .

8}: ,

0 -

0.67 '

B E;

I B-!

e I! i l

O' Inserted Rod Group No.

li I i

0.00 Balative Power Density f i

I I-

,.. -.m c_ - --.-u___. _ ._._._..___m. u.______ ,_,__ __  ;

l

6. THERMAL-iTJ*.A".LIC DESIGN The thermal-hydraulic design evaluation supporting cycle 5 operation used the methods and models described in references 1. 9.' and 10. The incoming batch 7 fuel is hydraulically and geometrically similar to the fuel remaining in the core from previous cycles. The cycle 4 and 5 ment == design conditions and significant parametere are shown in Table 6-1. The minimum DNBE shown at the design overpower is unchanged for cycle 5 and la based on 106.5% of EC design flow.10.4% ==rt== core bypass flow, the Mark-B4 fuel assembly, and includes the effects of in-core fuel densification.

For cycle 5 operation, a flux-to-flow trip setpoint of 1.08 is established.

This setpoint and other plant operation limits based on minimus DNBP. criteria contain a DNBR margin of 10.2% from the design minimum DNBR limit of 1.30.

In response to reference 11. B&W has coastitted to prepare a topical report ad-dressing the potential for, and ef fects of, fuel rod bow. In addition. B&W has submitted an interim rod bow penalty evaluation procedure12 for use until the topical report is completed and reviewed. As shown in reference 10, when this interim procedure is used, there is no DNBR penalty due to fuel rod bow for fuel burnup to approximately 21.300 mwd /stU. For oconee 3. cycle 5. the 11aiting (highest power) fuel assembly is always in fuel burned less than 21.300 MW4/stU and no DNER rod bow penalty is required.

6-1 Babcock & WHecx

. _. -3 :

g

.~-

Table'6-1. Marimas Deefs conditions for cycles 4 and 5 cr.l. ,

Design power level, NWt g

2568  ;

System preset.re, pela Reactor ecolant flow, Z design 2200

-g 106.5 -

• Vessel inlet / outlet coolmat temp at 555.6/602.4 g 1001 power F g

Raf design radial-i m i power ^

1.71 peaking factor '

Raf design axial flux shape 1.5 cosine .

Not channel factores Fmeh=1py rise 1.011 Beat flux 1.014 Flow area 0.98 Active iual lea 8th, 1s. -140.2 Avg beat flux at 100% power, Brn/h-ft2(a) 175 = 103 g

3 Max best flux s' we, Stu/h-ft2@) 450 = 103 CEF correlatiew BAW-2 Mia DER at design overpower (112%) 1.98

• Heat flux based on danaified length (in botteet core location).

I sased on average heat flux with referenca peaking.

I I

8 I

E I

I

. v. .

. 6., -

m.m .

w _ . -.w -: ~ .c_w_ .

if - w.._ a. u wuze + -- - -a .- E

.7. ACCIDENT A.VJ TRANSIENT AKALYSIS 7.*. Ceneral Safety Analysis Each FSARI accident analysis has been examined wit's respect to changes in cycla 5 parameters to detsruine et.e effect of the cycle 5 reload and to ensure that thermal performance during hypothetical transients is not degraded. The effects of fuel densification on the FSAR accident results have been evaluated and are i reported i:n reference 10. i Since batch 7 reload fuel assemblies contain fuel rods with a theoretical density higher than those considered in reference 10

)

the conclssions in that reference are still valid.

1Eo new does calculations ,were performed for the reload report. The dose ron- i

)

ciderations in the FSAR were based os maximum peaking and burnup for all core l

cycles; therefore, the dose considerations are indeps5 dent of the reload batch.

7.7 Accident Evaluation The key parameters that have the greatest erfeca; on determining the out< ?se of a transient can typically be classified in three major areast core thermal parameters, thermal-hydraulic parameters, and kinetics parameters. .c1? ding the reactivity feedback coefficients and control rod wurths.

Fuel thermal analysis parameters for each batch in cycle 5 are given in Table 4-2. Table 6-1 compares the cycle 4 and 5 thermal-hydraulic maximum design conditions. Table 7-1 compares the key kinetics parameters from the FS H. and cycle 5. Generic LOCA analyses have been performed for the B&W 177-FA lowered-loop WSS using the Fisial Acceptance Criteria ECCS evaluation hel rep: rted in reference 13. These analyses are generic in nature since the lialting values of the key parameters for all plants in this category were used. Furthermore, the combination of the average fuel temperature as a function of linear heat rate and the lifetime pin pressure data used in the 14CA limits analysesI3*I" is ccuservative compared to those calculated for this reload. Thus, the anal-yses and the LOCA limits reported its references 13 and 14 provide conservative 1

7_g Babcock & WImox i

e 2 results for the operation of Oconee 3. cycle 5 fuel. A tabulation showing the bonding valuss for allowable 14CA peak Lats for Oconee 3. cycle 5 fuel is pro- .

vided in Table 7-2.

From the e1tastination of cycle 5 core thermal properties and kinetics propertise with respect to acceptable previous cycle values. it is concluded that this core reload will not adversely affect the safe operation of the Oconee 3 plant during cycle 5. Considering the previously accepted design basis used in the FSAR and cubsequent cycles, the transient evaluatica of cycle 5 is considered to be bounded by previously accepted enalyses. The initial conditions of the transienta in cycle 5 are bo mded by the FSAR 'and/or the fuel densification report.18 I

. . g I

8 I

I I

I I

I I

E

,., -mm .

..____..-...___.__._.ua............__..._._.._.--__.....-_. --R

- ~ . . - ~ - - - . - - - . - - - - - - - ~ ~ - - - - - -

I Table 7-1. Cosparison of Key Parameters for Accident Analysis t FSARI Predicted Parameter value cycle 5 value BOC Doppler coeff,10-5 Ak/k/*F -1.17 -1.51 EOC Doppler coeff, 10-5 Ak/k/*F -1.33I * -1. 5 ' -

BOC moderator coeff,104 , Ak/k'F +0.5(D -0.66 EOC modcrator coeff, 10-" ak/k/*F -3.0 -2.69 All rod bank worth, EZP, 1 Ak/k 10.0 8.45 8 Boron reactivity worth, 70*F.

pps/11 Ak/k 75 76 Max. ejected rod wc rth, EFP, I Ak/k O.65 0.29 Dropped rod worth, RFP, I ak/k 0.46 0.20 laitial boron conc HFP, ppa 1400 1161 t b I"

-1.2 x 10-5 .1k/k/F was used for steam-line failure analysis.

k..

-1.3 x 10-5 Ak/k/F was used for cold water accident (pump start-up).

'?!.

r. ^k (b)+0.94 x 10~4 Ak/k/F was used for the moderator dilution accident. ,

hl k- (* Using reference 6.

i J

.(

[ ,g' J

g Table 7-2. Bounding Values for Allowable IDCA Peak 1.inear Heat Rates iges i

j. .

Allevable peak linear Core elevation, ft.

k' heat rate. kw/ft 1 2 45 15.5 j -

4 16.6 6 18.0 y 8 17.0 10 16.0 2

0

[

~ s$

?

.g

~ . ~ .. - . - . - - . - . - . - - =. .~

E

~

t -  !

1 1

• s j

W ] '

l 2

g 8. FROPOSED MODIFICATICIIS TO TIQBtICAL 1

j g SPECIFICATIONS j

9 I The Technical Specifications have been revised for cycle 5 operation in accord-  :

.j ance with the methods of references 15-17 to account for minor changes in power '

t peaking and control rod worths inherent with non-equilibrium cycles. In addi-tion: I q'

1. Oconee 3 will be changed from a rodded to a feed-and-bleed mode of opera-  :

) tion for cycla 5. This change is not regarded as a major change in cperat- I

$ ing ande since Oconee 3 was operated essentially in a rods-out configuration

,] during the latter part of the previous cycles. A similar change from redded to feed-and-bleed mode was approved for Three M114 Island Unit 1 for cycle

]h 4.18

2. The power level cutoff restrictico applied in previous cycles to the con-trol rod position limits of Figures 8-3 through 8-5 have been removed for cycle 5 operation. The change has been accomplished by designating the y power level cutoff at 100% FP. Any operating restrictions resulting from transient zenon-induced power peaks, including xenon-free startup are in-q herent.ly included in the control rod position and axial imbalance limits d of Figures 8-3 through 8-14.

3. A flux / flow trip setpoint of 1.08 is established for cycle 5 operation.

Based on the Technical Specifications derived from the analyses presented in this report. The Final Acceptance Criteria ECCS limits will not be exceeded.

q nor will the thermal design criteria be violated. Figures 8-1 threugh 8-14 are revisions to previous Technical Specification limits.

a -

n

.It i

fi S-1 Ba M a h

$L - -- - _ . ~ u. . . . - .-- --- - . . , .. . - - ~ . . . - -~=a a.-.- - --. ]

l 1

. .v

• 7 " ,-g_e' .E,}',

j s , + ,

+

g

, q Figure 8-1. Core Protection Safety 1.inits.

Utsit 3 ,

t i Thermal Poest Level. 5

\

UNACCEPTABLE l orERATl0N l

-44.3.312 37.832 ,

ACCEPTABLE

- 110 '

-52.102.5 /4PsWP OPERAftes , , ggg 49.2.101.8 44.8.87.2 -- IO 37.87.2 52,78.j ACCEPTASLE - - 80 384 P5WP #9.2.76.3 OPERATt0B ACCEPTABLE

,0 .......

g 2 384 PnWP 52.50.3 0 Etation - - 50 ss.2.as.s

_ _ 40 CetTE a

REACTOR COOLANT Ft0W, SP4 sn.sa0 (soo$)* g

- - 30 5 2 289.035 (74. 1) 3 183.690 (49.0%) , _ gg 10 l 1 I  ! I l

-60 -40 -20 0 20 40 60 Reactor Power labalance. 5

• 108.5% OF FiktT C0tt DEslGs FLOW I

I I

I 8-2 Bahk &hx

} _

', I

^

'r  :

Figree 8-2. Protectivs system hximus Allowable . 7. .

Setpoints. Unit 3 ..

~

t 'q Tkstsal Psust Level, 5 '

p ORACCEPTABLE '.'

l OPERATIDM , . . . .

T

(-2s.ios) (20.e. i os. o )

I s

, , gig

+

Y ACCEPTAsLE D ,.'

ib a

+%' I4 PEMP goPERATios

. ]W l #,

a3 (-s e. t . s s. o ) p l I (ss.,,,s,,3

. 33 g h-2s.so.s) gg (ts.so.s) .

(ACCEPTABLE A 384 PUNP

(-se.g.ss.s) loPtaAT40s -

- 78 I '

Q g l (ss.s.ss.s)

- , 80 g

-2s.52.s) (2s. s2. s) <

ACCEPTAsLE -

- 53

[z l {

g (-,o.i..s..) [o.sasPta Tion PowP- 4. i j

, (,s.s.3,.,,

I I -

- as f l -

I

' l g - - 29 I.  ;

L l J II - ,- t o le e I I" I I I

-60 -40 -20 0 20 48 80 f

Pcesr Imbalance. 5 i 4

1 8-3 N & h

.8 . . m .,a ... .-> . .w k - e.h .a - -

-mem--waw wwaw -me - w.. mat u- d .W. w*dh * - ' ' - -

i ,

\

> 5jj

( ~, p .. ; b i  !

j. 6 2 Figure S-3. Rod Poettion Limits for Four-Pump Operation From 0 to 100 t 10 ErPD - Oconee 3. Cycle 5 it.

(see ses) g{, i IN - (are.ses) i gg - '

(ast.ee) t-J 3NilT8911 bat 8111 LIElT te - '

i (ass,se)

OPERAfl811 IN Twit Attigli

~ il - 4g\@  !

it NOT ALLOVtB E

Power Level 8

• 50

• E Cutoff =

100% Fr 5} '

i

54 - I"****I en  !

MINISSIRE OPERATING tidim 30 -

g 1 re 7,,, 3 i

< e s. n) {

38 -

t!37tlCT[3 g (e.Ie )

$ I f f f f i t t t n , , '

O 20 48 64 IS e

164 120 148 150 ISO 20G 220 248 288 288 383 W Red lases. 5 Withdrawn y >r ii.

. 2. ..

l- f

,, I., I

.~> .,se, ,

f I e I

8 25 58 75 108 g

i t t i

Grows 6 I .

I I

" "'"** ~

. . - . . . . ~ - . - .- - - -- -- -- E-

l wm t ,

l Figure 8-4. Eod Position Limits for Tour-Pump Operation From '

100 2 10 to 200 2 10 EFFD - Oconee 3. Cycle 5 ,

lit (165.102) (I7e. e st ) ,

3Mlligosu aAR$1N 58 -

Ligli -

(tse.se)

=

II p *

.....e.)

OMRATION IN TMIS REGl0N

T 15 NOT ALLCvt0 Power Level se - Cutoff =

$ 100% FF

5, _ (......,

.: 48 -

E E 38 -

MRS113tKE OMRATins 28 -

REGION (e.s) ggs,es) 18 _

RESTRICTED 8 f(0.01i i e i i e , , , .

9 20 40 El 20 100 123 148 150 180 200 228 248 280 288 303 Bad lades. 5 Withdrava 8 5,5 100 2,5 7,5 8 25 SO 75 les I t t 9 i Group 5 graig ;

O i

25 59 75 108 f 1 1 t Crevs 5 1

8-5 Babcock & Wilcox

-~. .... ~. .._ n. u . ~ :s. a .....__ ,- 4 .._ __ _ _,_.: _ _ _ _ __ _ _ _ _ _ . _

l l

.}- f ',i :9 ?

~

~

'} ' ,

.;1. 'M

.l

..: v, .;.  : .;r

+

L' Figure 8-5.

Mod Position Limits for Four-Pump Operation After '

, .- 200 1 10 EFF3 - Oconee 3. Cycle 5

. . Ill, (ass. set) (27s.let)

.~7 1N ~ .

__; y

,.N

~

(28e.H)

• ~

~SPERATICN IN THIS REClelt (rse.es) 15 bl0T ALL0tta (tes.75) 18 -

G .

3 88 Power Level C -

Cutoff =

, 103

FF

= H -

(sw.se)

{ 43 S MTOCfN EARClu g Llulf g 5

~

gg,,,,,, PEREIS$1BLE CriaATiks REGION IS -

(e.e)

W 3 i f f I f f I I f f a f l 0 23 43- 63 84 100 122 140 165 180 2E8 223 243 28B 2to 333 E M Inder. 5 Sitt' drawn 8 25 58 75 t a e 10,0 ,8 2,5 50 75 100

, g g Ersup $ ggggy y 8 25 50 75 183 i f I t t Eroup 8 I

I E

B.

8-6 Babcock s. Wilcox g

---

L~---..w--.a..-..,.. -.~.:--. ..--......-~.....~..E i

l

. ..? ,

~* -

l

~

Figure 8-6. Rod Position Limits for Two- and Three-Pump Operation -

From 0 to 100 1 10 EFFD - Ocones 3, Cycle 5 118 (see. set) (its.se2) -,

188 - pc.W, E 98 ~

• I OPERAfl0ll IN Tills I""I 3 g, ,_

SECl0N l$ IIST ALLOSED 3

II -

$Hi!T83fil BARCIN LIElf g f

= 88 _

O

[ 58 _ (es.se) 3 48 ,.

2 4 811155t K E OPERATINE SEClON

38 -

l e

28 -( e .

} (le.le) 18 RESTSICit8 TOR 2 & 3 PUEP l 8 IbfI t t t I t t t t t , . t t 8 28 48 88 88 188 128 148 III IIS 208 222 283 288 288 388 Sed lades. S titherste 8 25 58 75 183 8 25 58 15 188 I I t i e f I f f a Eroup 5 Stsap 7 25 58 75 188 l 9 t t t i Ersup 8 ,

l l

1 i

. I s-7 Babcock a,Wilcox

_. .. __.--_. .___  :;_.,a.~._____.A_1_ m. e -- - - - - -

l

.. .. . . . . . m _

lp,.,

j.g!;7,.('MU,r-Nr9.ji.

7,.

, .< q - 1

.: . v , c.c C,

7. x. ,

,.; . r .< t. ,7 ,.. 1., <<.r .

.. 5Ajia  : . npf.t . .e::, ,. ' ,, :. .

u a.,,-

.v.. , ,

. ,e .,r

• : f9_ ,'.e; .-

, - ;- , ; s ' ;t ;e .. ,g ,,;-.-, ' ;. :..

.? ~+. , - -

c, w *..c ,

.n,',,,,c; q yg5ere 8-72 .' Rod Position Limits for Two- and Three-TW Operatice

- - ije n4 m

. 3;<

'From 100 1 10 to 200 2 10 ZFFD - Oconea 3, Cycle 5

.- j pg l (als,est)

-f g , _

1 55 '#. 1 3RT90Eli PANIN L1ElT

, " g. .

s 3

.3

>4 8PERATICII IN Tills R:2tSe II -

15 IIST ALLORES

- ~X 3

,_ CS -

. (144,ne)

- 54 -

g .

m.

• E a

. ~

g

= 3g .

PERfl53tti.E SPERATING RtGl84

. 29 - -

g (ss,ss)

_a -

I.

E!!TRICTED TCR 2 1 3 PtsP g l O *) I f f f I f f f t , , ,

S 28 48 58 63 IGG 120 140 188 150 202 220 248 233 233 333 Red lades. S tithdrsan W 8 25 58 75 183 8 25 54 75 les ill t

• 1 Grasp 5 f i I t , e i g gre,, y a 25 58 75 188 t I  !  ! 1 Gresp 8 8

I I

I I

s-8 Babcock & Wilcox

_ . . _ . _ _ _ _ . __ _ _ _._ B 1 l

\

a

~

Figure 8-8. Rod Position I.inits for Two- and Three-Fuap Operatico ,

After 200 2 10 EFFD - Oconee 3, Cycle 5 ,

Ilt (res.ler)

Its -

1 se - .'

E. . .

E OPERATI0li IN Till! RttlDel

)

88 - '

IS NOT ALLOWED gg a- 78 SIGUIDMII BARCIN 2* LIEIT

SG -

\

=

Se - (ese.se) 5 48 -

~ '

53 -

• PEtsl55tglE SPERATileG REGISit

% 20 -

(Iss as)

It ,

(8.e) s e e t t t t i e i e e t j 8 28 40 la 80 100 128 140 150 180 2C8 22 240 259 284 300 tes laces, s gitssteen 4 25 58 75 tot 8 25 58 75 100 m t t t t I g g r Group 5 Grose 7 8 25 50 75 ISO I f f I f Ctsup 8 4

89 Babcock &WUcox

_ . . . _ _ . _ _ . _ _ _ . _ ~ . .. _ __..__-.m._-.._.._ _ _ ___._._..__ __ -

_ _ , _ . . - - - - , m -- ~ - - - - '

lGh,(jf;cg; t

. s .. :

-a .

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9.

I STARTUP PROGRAM - PHYSICS TESTI2iG The planned startup test program associated with core performance is outlined below. These tests verify that core performance is within the assumptions of the safety analysis and provide confirmation for continued safe operation of the unit.

9.1. Precritical Tests, 9.1.1. Control Rod Trip Teg Precritical control rod drop times are recorded for all control rods at hot full-flow conditions before zero power physics testing begins. Acceptance criteria state that the rod drop time from fully withdrawn to 75% inserted shall be less than 1.66 seconds at the conditions above.

It should be noted that safety enalysis calculations are based on a rod drop time of 1.40 seconds fron fully withdrawn to two-thirds inserted. Since the mat accurate position *ndication is obtained from the zone reference switch

{ at the 751-inserted position, this positioc is used instead of the two-thirds inserted position for data gathering. The acceptance criterion of 1.40 seconds corrected to a 75%-inserted position (by rod insertion versus time correlation) 4 is 1.66 seconds.

9.2. Zero Power Physics Tests 9.2.1. Critical Boron Concentration Criticality is obtained by deboration at a constant dilution rate. Once crit-icality is achieved, equilibrium boron is obtained and the critical boron con-centration deternined. The critical boron concentration is calculated by cor-recting for any rod withdrawal required af ter achieving equilibrium boron. The acceptance criterion placed on critical boron concentration is that the actual boron cone.cntration must be within 2100 ppm boron of the predicted value.

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a 9.2.2. Temperature Reactivity Coefffcient The isothermal temperature coefficient is measured at approximately the all-rods-out configuration and at the hot zero power rod insertion limit. The average coolant temperature is varied by first decreasing then increasing ten-perature by 5'F. During the change in temperature, reactivity feedback is com-pensated by discrete change in rod motion, the change in reactivity is then calculated by the summation of reactivity (obtained from reactivity calculation on a strip chart recorder) associated with the temperature change. Acceptance criteria state that the measured value shall not differ free the predicted value by more than 10.4 x 10-* (ak/k)/*F (predicted value obtained from Physics Test Manual curves).

The susderator coefficient of reactivity is calculated in conjunction with the temperature coefficient measurement. After the temperature coefficient has been measured, a predicted value of fuel Doppler coefficient of reactivity is added to obtain moderator coefficient. This value must not be in excess of the acceptanse criteria limit of +0.5 x 10~" (ak/k)/*F.

9.2.3. Control Rod Croup Rs* activity Worth l Control bank group reactivity worths (groops 5, 6, and 7) are measured at hot 5 zero power conditions using the boron / rod swap method. The boron / rod swap method consists of establishing a deboration rate in the reactor coolant sys-5 W

tem and compensating for the reactivity changes of this deboration by inserting control rod groups 7, 6, and 5 incresental steps. The reactivity changes that occur during these measurements are calculated based on Reactimeter data, and differential rod worths are obtained from the measured reactivity worth versus the change in rod group position. The differential rod worths of each of the controlling groups are then russed to obtain integral rod group worths. The acceptance criteria for the cent;ol bank group worths are as follows:

1. Individual bank 5, 6, 7 worth:

predicted value - measured value = 100 s 15 measured value

2. Sua of groups 5, 6 and 7:

predicted value neasured value measured value = 100 10 i

Babcock s,Wilcox 9-2 3

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9.2.4 Elected Control Rod Reactivity Vorth After the CRA groupa have been positioned near the minimum rod insertion limit, the ejected rod is borated to 100% withdrawn and the worth obtained by adding the incremental changes in reactivity by boration.

After the ejected rod has been "sorated to 100% withdrawn and equilibrium boron established, the ejected rod is then swapped in versus the controlling rod group and the worth determined by the change in the previously calibrated con-trolling rod group position. The boron swap and rod swap values are averaged and error-adjusted to determine ejected rod worth. Acceptance criteria for the ejected rod worth test are as follows:

1. predicted value measured value measured value = 100 s 20

2. Measured value (error-adjusted) i 1.0% Ak/k The predicted ejected rod worth is given in the Physics Test Manual.

9.3. Power Escalation Tests 9.3.1. Core Power Distribution Verification at 4 0, 75 and 100% FP With Nominal Control Rod Position Core power distribution tests are performed at 40, 75, and 100% full power (FP). The test at 40 FP is essentially a check on power distribution in the core to identify any abnormalities before escalating to the 75% FP plateau. -

Rod index is established at a nominal full power rod configuration at which the core power distribution was calculated. APSR position is established to 1 provide & core power imbalance r:orresponding to the imbalance at which the core power distributfor, calculations were performed.

The following acceptance criteria are placed on the 40% FP test:

1.

The worst-case maximum linear heat rate must be less than the 14CA limit.

2. The minimum DNBR aust be greater than 1.30.

3. The vslue obtained from the extrapolation of the minimum DNBR to the next power plateau overpower trip setpoint must be greater than 1.30 or the extrapolated value of imbalance must fall outside the RPS power / imbalance /

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4 The value obtained from the extrapolation of the worst-case maximum linear heat rate to the next power plateau overpower trip setpoint must be less than the fuel melt limit or the extrapolated value of imbalance aust fall outside the RPS power / imbalance / flow trip envelope. l

5. The quadrant power tilt shall not exceed the limits specified in the Tech-nical Specifications. .

6. The highest measured and predicted radial peaks shall be within the follow-ing limita: I predicted value measured value , , g measured value

7. The highest measured and predicted total peaks shall be within the follow-ing limits:

predicted value measured measured value value = 100 sM Items 1. 2, 5. 6. and 7 above are established to verify core nuclear and ther-mal calculations 1 models, thereby verifying the acceptability of data from these models for input to safety evaluations.

Itema 3 and 4 establish the criteria whereby escalation to the next power pla- l teau may be accomplished without exceeding the safety limits specified by the cafety analysis with regard to DNBR and linear heat rate.

The power distribution tests performed at 75 and 100% FP are identical to the j 4CK FP test except that core equilibrium xenon is established prior to the 75 I and 100% FP tests. Accort.ingly, the 75 and 100% FP measured peak acceptance .

criteria are as follows:

1. The higheet measured and predicted radial peaks shall be within the follow-ing limita: '

' predicted value acanured value measured value x 100 s5

2. The highest measured and predicted total peaks shall be within the follow-ing limits:

predicted g lue measured value neasured value w 100 s 7.5

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9.3.2. Incore Vn Excore Detector Imbalance Correlation Verif f eation at s40% FP Imbalances are set up in the core by control rod positioning. Imbalances are read simultaneously on the incore detectors and excore power range detectors fer various imbalances. The execre detector offset Vs incore detector offset clops must be at least 1.25. However, this slope criterion may be reduced to 1.15 if the test is repeated at 75% FF af ter a suitable period of operation ct 100% FF. If the excore detector offset Vs incore detector offset slope criterion is not met, gain amplifiers on the excore detector signal processing equipment are adjusted to provide the required gain.

9.3.3. Temperature Reactivity Coefficient at 4100% FP The average reactor coolant temperature is decressed and then increased by cbout 5'F at constant reactor power. The reactivity associated with each ten-perature change is obtained from the change in the controlling rod group post-tion. Controlling rod group wr rth is measured by the fast insert / withdraw metnod. The temperature reactivity coefficient is calculated from the mes-cured changes in reactivity and temperature.

Acceptance criteria state that ti.e moderator temperature coefficient shall be n2gative.

_9.3.4 Power Doppler Peactivity Coefficient at s100T FP Reactor power is decreased and then increased by about 5% FP. The reactivity change is obtained from the change in controlling rod group position. Control rod group worth is measured using the fast inse-t/withdrau method. Reactivity corrections are made for changes in xenon and reactor coolant temperature that occur during the measurement. The pcwer Doppler reactivity coefficient is cciculated from the measured reactivity change, adjusted as stated above, and' the measured power change. --

The predicted value of the power Doppler reactivity coefficient is given in ths Physics Test Manual. Acceptance criteria state that the measured value chall be more negative than -0.55 x 10* (ak/k)/I FP.

9.4 Procedure for Use When Acceptance Criterin Are Not Met If acceptance criteris for any test are not met, an evaluation is performed before the test program is continued. This evaluation is performed by site

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test personnel with participation by Babcock & Wilcox technical personnel as required. Further specific actions depend on evaluation results. These ac-tions can include repeating the tests with more de: tailed attention to test prerequisites, added tests to search for anomalies, or design personnel per-W forming detailed analyses of potential safety problems because of parameter deviation. Power is not escalated until evaluation shows that plant safety will not be compromised by such escalation. -

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i REPDLENCES I

Oconee Nuclear Station, Units 1, 2, and 3 Final Safety Analysis Report, Docket Nos. 50-269, 50-270, and 50-287.  ;

1 2

A. F. J. Eckert, H. W. Wilson, and K. E. Toon, Progran to Determine In- l l

reactor Performance of B&W Fuels - Cladding Creep Collapse, BAW-10084P-A, j Rev 2, Babcock & Wilcox, January 1979.

3 BPRA Retainer Design Report BAU-1496, Babcock & Wilcox, May 1978.

i j C. D. Morgan and H. S. Rao, TAFT - Fuel Pin Temperature ond Cas Pressure Analyals, Babcock & Wilcox, BAW-10044 Hay 1972.

5 Oconee Unit 3, Cycle 4 - Reload Report, BAu-1486, Babcock & Wilcox, May 1978.

6 B&W Version of PDQ07 Code, BAW-10117A, Babcock & Wilcox, January 1977.

7 Core calculational Techniques and Procedures, BAW-10118 Babcock & Wilcox, October 1977.

' s Assembly Calculations and Fitted Ev. lear Data, BAW-10116A. Babcock & Wilcox, May 1977.

9 Oconee Unit 3 Cycle 4 - Reload Report, Amendment, BAW-1486, Babcock &

Wilcox, June 12, 1978.

10 Oconee 3 Fuel Densification Report, BAW-1399, Babcock & Wilcox, November 1973.

l l

11 D. B. Vassallo (NRC) to J. H. Taylor (B&W) letter, " Calculation of the l Ef t'ect c,f Fuel Rod Bowing on the Caitical Heat Flux for Pressurized Water Reactors" (Revised September 15, 1978), June 12, 1973.

12 J. H. Taylor to D. B. Vassallo, letter, " Determination of the Fuel Rod Bow DNb Penalty " December 13, 1978.

13 ECCS Analysis of B&W's 177-FA Lowered-Loop NSS, BAW-10103. Rev. 2 Babcock

& Wilcox, September 1975.

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. r 1* J.' E. Taylor to S. A. Varga (NRC), 14tter, July 18, 1978.

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15 Power Peaking Nuclear Rallability Factors, RAU-10119, Rabcock & Wilcos, '

January 1977. .

IE Normal Operating Controls, RAU-10122, Rabcock & Wilcox, Augut 1978.

17 Verification of the Three-Dimensional FIAME Code, EAU-10125A, Rabcock & ~

Wilcox, Augu.st 1976.

Is Three Mile Island Unit 1, Cycle 4 Reload Report, RAU-1473. Rev. 3 Rabcock

& Wilcos, May 1978.

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