W3P86-1686, Forwards First Part of Reload Analysis Rept,Containing Sections 1-6,9,11 & 12.Remaining Sections Will Be Included in Second Part to Be Submitted on 861001.Rept Submitted in Two Parts to Expedite Review Process,Per Discussion

Forwards First Part of Reload Analysis Rept,Containing Sections 1-6,9,11 & 12.Remaining Sections Will Be Included in Second Part to Be Submitted on 861001.Rept Submitted in Two Parts to Expedite Review Process,Per Discussion

ML20212Q107
Person / Time
Site:	Waterford
Issue date:	08/29/1986
From:	Cook K LOUISIANA POWER & LIGHT CO.
To:	Knighton G Office of Nuclear Reactor Regulation
References
W3P86-1686, NUDOCS 8609030452
Download: ML20212Q107 (51)

Text

o '

T LOUISi POWER & ANA LlGHT/ 317NEWBARONNESTREET e P. O. BOX 60340 ORLEANS, LOUISIANA 70160 * (504)5953100 f

j EWs!?sG August 29, 1986 W3P86-1686 3-A1.01.04 A4.05 QA Mr. George W. Knighton, Director PWR Project Directorate No. 7 Division of PWR Licensing-B Office of Nuclear Reactor Regulation

! Washington, D.C. 20555

SUBJECT:

Waterford 3 SES Docket No. 50-382 Reload Analysis Report (RAR)

Dear Mr. Knighton:

Provided herewith are three copies of the first part of the Waterford Reload Analysis Report. As previously discussed with you, the Reload Analysis Report will be submitted in two parts in order to expedite the regulatory review process. This part of the RAR contains sections 1.0, 2.0, 3.0, 4.0, 5.0, 6.0, 9.0, 11.0, and 12.0. The second part, which will contain the remaining sections of the RAR, is scheduled to be submitted to the NRC on October 1, 1986.

Please contact me or Robert J. Murillo should you have any questions.

Very truly yours, K.W. Cook V

Nuclear Support & Licensing Manager KWC/RJM/pim Attachmentu .

cc: B.W. Churchill, W.M. Stevenson, R.D. Martin, J.ll. Wilson, . I,L NRC Resident inspector's Office (W3) C ' (,'

'(0 t,

'#r.v 0609030452 060029 3 h (;

PDR P

ADOCK 05000302 PDR h I

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"AN EQUAL OPPORTUNITY EMPLOYER"

RELOAD AtlALYSIS REPORT FOR WSES-3 CYCLE 2 (Partlaj Submittal)

- - , _ . " " ' - ~ * - - - - . _ . _ , _ , ._ _ '* - , . , , , . , - - .

TABLE OF CONTENTS

'P' AGE l 1. INTRODUCTION AND

SUMMARY

1-1

2. OPERATING HISTORY OF THE REFERENCE CYCLE 2-1

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

! 5. NUCLEAR DESIGN 5-1 l

6. THERMAL-HYORAULIC OESIGN g 6-1 1 7. TRANSIENT ANALYSIS 7-1 i R. ECCS ANALYSIS 81

9. REACTOR PROTECTION AND MONITORING SYSTEM 9-1

10. TECHNICAL SPECIFICATIONS 10-1

11. STARTUP TESTING 11-1 .

12. REFERENCES 12 1

}

l i

1.0 INTRODUCTION

AND

SUMMARY

This report provides an evaluation of the design and performance of Waterforf Steam Electric Station Unit 3 (WSES-3) during its second . ,

cycle of operation at 100% rated core power of 3390 MWt and NSSS power

. of 3410 MWt. Operating conditions for Cycle 2 have been assumed to be consistent with those of the previous cycle and are sumarized as full f power operation under base load conditions. The core will consist of j trradiated Batch B and C assemblies, along with fresh Batch 0 assemblies. The Cycle 1 termination burnup has been assumed to be

between 13,400 and 14,400 MWD /T.

I The first cycle of WSES-3 will hereafter be referred to in this

, report as the " Reference Cycle." .

The safety criteria (margins of safety, dose limits, etc.) applicable for WSES-3 were established in the Cycle 1 FSAR (Reference 1-1). A j review of all postulated accidents and anticipated operational I

occurrences has shown that the Cycle 2 core design meets these safety criteria.

i .

i The Cycle 2 reload core characteristics have been evaluated with

respect to the Reference Cycle. Specific differences in core fuel

! Ioadings have been accounted for in the present analysis. The status

! of the postulated accidents and anticipated operational occurrences for Cycle 2 can be sumarized as follows:

l 1. transient input data are less severe than those of the Reference Cycle analysis; therefore, no reanalysis is necessary, or l 2. transient input data are not bounded by those of the Reference Cycle analysis, therefore, reanalysis is required.

j ,

For those transients requiring reanalysis (Type 2), analyses are I

presented in Sections 7 and 8 showing results that . meet the established safety criteria.

1-1

The Technical Specification changes requested for Cycle 2 are summarized in Section 10 and described ir greater detail in separate license mendment applications. ,

Modifications to the Core Protection Calculator (CPC) System are being made to imp rove performance and reflect the Cycle 2 core configuration. Algorithm changes are a result of the CPC Improvement Program (CIP) and are summarized in Section 9. The concurrent data base changes are a result of plant-specific application of the CIP to WSES-3 Cycle 2.

1 0

. I

! 1-2

- - - - - - ,n.- ,-. - . -- . . - - - - . - - - . - _ - - - - - - - - . . _ - - - - - - - _ - . . - - . - - - - . - - _ . - -

2.0 OPERATING HISTORY OF THE REFERENCE CYCLE WSES-3 is currently in its first fuel cycle which began with initial critical fey in March, 1985. power Ascension began in March, 1985 -

and on September 24, 1985 the unit was declared in commercial operation.

It is presently estimated that Cycle 1 will terminate on or about November 15, 1986. The Cycle 1 termination point can vary between 13,400 MWD /T and 14,400 MWD /T to accomodate the plant schedule and still be within the assumptions of the Cycle 2 analyses.

Outages occurred in October and December,1985, and in July,1986 for replacement of reactor coolant pump seals and in March, 1986 for maintenance and Technical Specification Surveillance. The capacity factor through July, 1986 was 77%.

f i

l I

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, =

1 3.0 GENERAL DESCRIPTION The Cyc1r 2 core will consist of those assembly types and numbers .

listed in Table 3-1. Seventy-three Batch A and nineteen' Batch B assemblies will be removed from the Cycle 1 core to make way for 92 fresh, Batch D assemblies. Sixty-one Batch B and all Batch C assemblies now in the core will be retained.

The reload batch will consist of 24 type 00 assemblies,12 type D1 assemblies with 4 burnable poison rods per assembly, 24 type D2 assemblies with 4 burnable poison rods per assembly and 32 type D3 assemblies with B burnable poison rods per assembly. These sub-batch types are zone-enriched and their configurations are shown in Figure 3-1.

The loading pattern for Cycle 2, showing fuel type and location, is displayed in Figure 3-2.

Figure 3-3 displays the beginning of Cycle 2 assembly average burnup distribution along with the initial assembly average fuel enrichment.

The burnup distribution is based on a Cycle 1 length of 14,400 MWD /T.

Control element assembly patterns and in-core instrument locations will remain unchanged from Cycle 1 and are shown in Figure 3-4 and Figure 3-5 respectively.

3-1

TABLE 3-1

~~ '

WATERFORO STEAM ELECTRIC STATION 3 Cycle 2 Core Loading Number Initial Total Number Ass embly Fuel Rods Initial Poison Poison of Desig. Number of per Enrichment Rod Loading Fuel Poison nation Ass emblies Assembly (w/o U-235) Assembly (gm BIO /in) Rods Rods B 61 208 2.41 16 .02276 12688 976 12 1.87 732 C 40 224 2.91 0 0 8960 0 12 2.41 480 C. 8 212 2.91 12 .01034 1696 96 12 2.41 96 C+ 16 208 2.91 16 .01034 3328 256 12 2.41 192 00 24 184 3.90 0 0 4416 0 52 3.40 1248 01 12 180 3.90 4 .019 2160 48 52 3.40 624 02 24 220 3.40 4 .019 5280 96 12 2.78 288 03 32 216 3.40 8 .019 6912 256 12 2.78 384 Total 217 49484 1728 3-2

1. M I - G M .. e X SUB-BATCH DO 24 ASSEMBLIES d E li s"b

=

xm - xx 0 3.90 w/o U 235 mr

,. , m m x m .' S 3.40 w/o U 235 1 JOC i I

- , , mr cm E' I xmo W i JG 3 6 JC M t JC3G YDIC W x '

,cr .

x SUB BATCH DI 12 ASSEMBLIES xx M% xx- -rx ,

b O 3.90 w/o U 235 i u 3 3.40 w/o U 235 m x 1

% = i E B 4C AL 023SHIM PIN xx- -xx 4 -E. Im b. ,

r x mr , m "O M 1 , ,, , i i SUB BATCH D2 24 ASSEMBLIES ll, ,  ; & O 3.40 w/o U 235 ii i S 2.78 w/o U 235

, i i i. ,

', , ,;  ; E B 4C AL 023SHIM PIN H I; ',, l ,

1 m  ! l1 i x

-, ,, .. ;c 23

,* l SUB BATCH D3 32 ASSEMBLIES 4 <l H' - 0 3.40 w/o U 235

l 1 i S 2.78 w/o U 235

',, , i; E B 4C AL 023SHIM PIN 7

i

>= , , ,

L'* X M I

l LGV151ANA Figure WATERFORD 3 CYCLE 2 PCWER & LIGHT CO. 31 Waterford Steam ENRICHMENT ZONING PATTERN FOR l BATCH D ASSEMBLIES Electric Station 33

i A B C D E F G H J K LM N P R S T V W X Y I l l l l l l 1 Il l l l l 1 I I I I I I

! 1- B 00 D0 ,B & NORTH 1 2- - C D0 D0 D3 B D3 D0 D0 C 3- B D1 C C C+ D2 C+ C C D1 B 4- B D1 C D3 8 D2 B D2 B D3 C D1 B 5- C D1 C D3 8 C B D3 8 C B D3 C 01 C i

6- DO C' D3 B D2 B D2 B D2 B D2 B D3 C 00 1 ,

7- -

DO C B C B D3 C- D3 C- D3 8 C B C DO -

8- B B 8- -

D3 C+ D2 B D2 C. C+ C+ C+ C. D2 8 D2 C+ D3 -

10~

] 00 00 11 - - B D2 'B D3 8 D3 C+ B C+ D3 8 D3 8 D2 8 -

12 -

00 00 13 - - D3 C+ D2 B D2 C- C+ C+ C+ C- D2 B D2 C+ D3 14 - B B 15 - - 00 C B C B D3 C. 03 C' D3 B C B C 00 16 - 00 C D3 B D2 B D2 B D2 B D2 B D3 C 00 17 - C D1 C D3 8 C B D3 B C B D3 C* D1 C

! 18 - B D1 C D3 8 D2 B D2 8 D3 C D1 B 1

19 - B D1 C C C+ D2 C+ C C 01 B 20 - C 00 D0 D3 8 D3 DO 00 C 21 -

3 00 DO B LOUIS!ANA Rgure POWER & l.lGHT CO. WATERFORD 3 Wotorford Steam CYCLE 2 CORE MAP 32 Electric Station 34

XXX INITIAL ASSEMBLY AVERAGE ENRICHMENT w/o U 235 YYYY BOC ASSEMBLY AVERAGE BURNUP (MWD /T) l EOC 1 = 14,400 MWD /T T.

2.38 3.79

16131 0 2.89 3.79 3.79 3.37 2.38 11428 0 0 0 15094

. 2.38 3.79 2.89 2.89 2.89 3.37 16162 0 9135 8310 16349 0 i

2.38 3.79 2.89 3.37 2.38 3.37 2.38 16162 0 11367 0 14648 0 16080 i

i 2.89 3.79 2.89 3.37 2.38 2.89 2.38 3.37 11428 0 11367 0 16717 8913 14388 0 I.

1 3.79 2.89 3.37 2.38 3.37 2.38 3.37 2.38

O 9135 0 16717 0 16869 0 16080

3.79 2.89 2.38 2.89 2.38 3.37 2.89 3.37

! 0 8310 14548 8913 16869 0 13987 0 l 2.38 16131 3.37 2.89 3.37 2.38 3.37 2.89 2.89 2.89 2 0 16349 0 14388 0 13987 14043 14043 3.79 0

2.38 3.37 2.38 3.37 2.38 3.37 2.89 2.38 k~~ 15094 0 16080 0 16080 0 14043 16542 4

1 l

I US NA WATERFORD 3 CYCLE 2 Figure PO T O*

Waterford Steam ASSEM8LY AVERAGE INITIAL ENRICHMENT 33 '

Electric Station AND AVERAGE BURNUP DISTRIBUTION i

35

i PLCEA PART LENGTH CEA BANK 6 - LEA.D REGULATING BANK 5 - SECOND REGULATING BANK -

4 -THIRD REGULATING BANK 3 - FOURTH REGULATING B ANK 2 - FIFTH REGULATING BANK 1 2 1 S

1 - LAST REGULATING BANK A S

BS gg08N 3 4 5 6 7 SA- SHUTDOWN 2 BANK A 8 9 10 11 12 13 SB 3 '4 15 16 17 18 19 20 1 S S A B 24 25 26 27 28 5 PLCEA 6 33 34 35 36 S 3 A B 42 43 44 4 1 52 53

• SHUTDOWN ROD IN POSITION Sj 52 IS AVAILABLE FOR ONLY TWO DIAGONALLY OPPOSITE CORE 62 QUADR ANTS .

2 I

4 WATERFORD 3 CYCLE 2 Rgure POkR& GHT CO*

Waterford Steam CEA BANK IDENTIFICATION 34

Electric Station 3-6

A N e e e e e .,

9 9 9 e 9 9 e

9 e @ e 9 9 e

! e e 1

4 e 9 @ 9 9 9 e e e e e e e el e e i

e e e el el el e l I I e e e e e i

l 1

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1 i

LOUI5!ANA WATERFORD 3 CYCLE 2 Rgure

! POWER & LIGHT CO. IN CORE INSTRUMENT ASSEMBLlES Waterford Steam 3,5 j CORE LOCATIONS Electric Station

3-7

4.0 FUEL SYSTEM DESIGN 4.1 ME,CHANICAL E DESIGN 4.1.1 Fuel Design The mechanical design of the Batch D reload fuel assemblies is identical to that of the Core 1 fuel assemblies (Reference 4-1) with the exception of the design features listed bel ow. No changes in mechanical design bases have occurred since the original fuel design. However, the ' Batch D design incorporates a number of refinements for the purpose of improving fuel handling, improving the burnup capability of the fuel, and improving the reconstitution feature of the fuel. The specific changes are discussed in the following paragraphs:

1. The perimeter strip lead-in tabs of the lowest spacer grid (made of Inconel) have been changed f rom trapezoidal-shaped to curve-shaped to improve fuel handling by reducing the chance of grid hangup. The remaining spacer grids (made of Zircaloy) already have curve-shaped perimeter strip lead-in tabs.

2. The shoulder gap in the Batch 0 fuel has been increased to improve the burnup capability of the fuel. The Batch 0 shoulder gap is 2.382 inches, compared to 1.332 inches for part of Batch i

B and 2.032 inches (after shimming) for the remainder of Batch B and Batch C. The Batch 0 shoulder gap was achieved by lengthening the guide tubes by 0.9 inches and shortening the fuel i rods by 0.15 inches. The changes do not result in the violation of any design criteria.

3. The designs of the CEA guide tubes and wear sleeves have been modified to f acilitate fuel bundle reconstitution. Reference 4-3 is the NRC's acceptance of the design change. ,

41 4

_ . _ _ , - - , - - - , . _ _ _ . ~ _ - , . _ _ _ - - - - . .

.~ .- - _ _ _ _ - - . _ . _ .

I J

4.1.2 Clad Collapse The fuel _to be included in the Cycle 2 core satisfies the bases for not requiring clad collapse analyses, as documented in Ref.orence 4-5. '

Therefore, no cycle specific clad collapse analysis was performed for Cycle 2. ,

4.2 MITIGATION OF GUIDE TUBE WEAR All fuel assemblies which will be placed in CEA locations in Cycle 2 will have stainless steel sleeves installed in the guide tubes to prevent guide tube wear.

4.3 THERMAL DESIGN The thermal perfonnance of composite fuel rods that envelope the rods of fuel batches 8, C and D present in Cycle 2 have been evaluated using the FATES 3A version of the C-E fuel evaluation model (References 4-7 and 4-8) as approved by the NRC (Reference 4-9). The analysis was performed using a power history that enveloped the power and burnup levels representative of the peak pin at each burnup interval, f rom beginning of cycle to end of cycle burnups. The burnup range analyzed is in excess of that expected at the end of Cycle 2.

Results of these burnup dependent fuel perfonnance calculations were used in the Transient Analysis presented in Section 7 and in the ECCS Analysis presented in Section 8.

4.4 CHEMICAL DESIGN The metallurgical requirements of the fuel cladding and the fuel assembly structural members for the Batch 0 fuel are identical to those of the fuel batches included in Cycle 1. Thus , the chemical or metallurgical performance of the Batch 0 fuel is expected to remain unchanged from the performance of the Cycle 1 fuel (Reference 4-1).

4-2

---.-,--,..-.--..n-, n_ ,. , , - - - , - . . . - - .

5.0 NUCLEAR DESIGN 5.1 pHYSICSCT4ARACTERISTIES 5.1.1 Fuel Management The Cycle 2 core makes use of a low-leakage fuel managment scheme, in which previously burned Batch B and C assemblies are placed on the core periphery. Most of the fresh Batch 0 assemblies are located throughout the interior of the core where they are mixed with the previously burned fuel in a pattern that minimizes power peaking.

4 With this loading and a Cycle 1 endpoint at 13,800 MWD /T, the Cycle 2 reactivity lifetime for full power operation is expected to be ,14,750 MWD /T. Explicit evaluations have been performed to assure applicability of all transient analyses to a Cycle 1 termination burnup of between 13,400 and 14,400 MWD /T and for a Cycle 2 length up to 16,000 MWD /T.

l Characteristic physics parameters for Cycle 2 are compared to those of the Reference Cycle in Table 5-1. The values in this table are intended to represent nominal core parameters. Those values used in the safety analyses (see Sections 7 and 8) contain appropriate 1

uncertainties, or incorporate values to bound future operating cycles, 4

and in all cases are conservative with respect to the values reported in Table 5-1.

Table 5-2 presents a summary of CEA reactivity worths and allowances for the end of Cycle 2 full power steam line break transient with a comparison to the Reference Cycle data. The full power steam line break was chosen as a reasonable illustation of the differences in

! CEA reactivity worths for the two cycles.

The CEA core locations and group identifications remain the same as in the Reference Cycle. The assumed power dependent insertion limits (p0!L) for regulating groups and part length CEA groups are shown in Figures 5-1 and 5-2 respectively. Table 5-3 shows the reactivity worths of various CEA groups calculated at full power conditions for Cycle 2 and the Reference Cycle.

5-1

5.1.2 Power Distribution  ;

Figures 9'-3 through 5-5 illustrate the calculated All Rods Out (ARO) planar radial power distributions during Cycle 2. The one-pin planar radial power peaks presented in these figures represent the mid-plane of the core. Time points at the beginning, middle, and end of cycle were chosen to display the variation in maximum planar radial peak as a function of burnup.

Radial power distributions for sel ected rodded configurations are given for BOC and EOC in Figures 5-6 through 5-11. The rodded configurations shown are: part length CEAs (pLCEAs), Bank 6, and Bank 6 plus the PLCEAs. As is the case for unrodded configurations, the largest planar radial peak for each of these rodded configurations occurs at beginning of Cycle 2.

The radial power distributions described in this section are calculated data which do not include any uncertainties or allowances.

The calculations performed to determine these radial power peaks explicitly account for augmented power peaking which is characteristic of fuel rods adjacent to the water holes.

Nominal axial peaking f actors are expected to range f rom 1.20 at 80C2 to 1.09 at E0C2.

5.2 SAFETY RELATED DATA 5.2.1 Augmentation Factors As indicated in Reference 5-1, the increased power peaking associated with the small interpellet gaps found in modern fuel rods (non-densifying fuel in pre-pressurized tubes) is insignificant compared to the uncertainties in the safety analyses. The report concluded that augmentation f actors can be eliminated f rom the reload analyses of any reactor loaded exclusively with this type of fuel. Therefore, augmentation f actors have been eliminated for Waterford 3 Cycle 2.

5-2

a 5.3 PHYSICS ANALYSIS METHODS 5.3.1 Analytical input to In-Core Measurements in-core detector measurement constants to be used in evaluat.lfg the reload cycle power distributions will be calculated in accordance with Reference 5-4.

5.3.2 Uncertainties in Measured Power Distributions The planar radial power distribution measurement uncertainty of 6.92%,

based upon Reference 5-4, will be applied to the cycle 2 COLSS and CPC on-line calculations which use planar radial power peaks. The axial and three dimensional power distribution measurement' uncertainties are determined using the values in Reference 5-4 in conjunction with

] other monlioring and protection system measurement uncertainties.

4

5.3.3 - Nuclear Design Methodology The Cycle 2 nuclear design was performed with two and three dimensional core models using the ROCS computer code with the MC module and employ-ing DIT calculated cross sections. The RDCS-DIT and the MC module was described in Reference 5-5.

I E

5-3

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

TABLE 5-1 WATERFORD 3 CYCLE 2 NOMINAL PHYSICS CHARACTERISTICS Reference

~

Dissolved Boron Units Cycle Cycle 2 Dissolved Boron Concentration for Criticality, CEAs Withdrawn, Hot Full Power PPM 452 1156 Equilibrium Xenon, BOC Boron Worth Hot Full Power, BOC PPM /% ao 79 107 Hot Full Power, EOC PPM /% ap --

85 ,

Moderator Temoerature Coefficients Hot Full Power, Equilibrium Xenon Beginning of Cycle 10-4ac/0F -1.2 -0.1 End of Cycle 10-44/ 0F -2.3 -2.5 Doppler Coefficient Hot Zero Power, BOC 10-53 ,foF -1.5 -1.7 Hot Full Power, BOC 0 10-bac/ F -1.1 -1.2 Hot Full Power, E0C 10-53 ,f oF -1.2 -1.4 Total Delayed Neutron Fraction,8eff BOC ------- 0.0072 0,0063 E0C ------- 0.0053 0.0051 Neutron Generation Time,t*

BOC 10-6 sec 30.0 23.9 EOC 10-6 sec ----

30.0 5-4

TABLE 5-2 WATERFORD 3 CYCLE 2 LIMITING VALUES OF REACTIVITY WORTHS AND ALLOWANCES FOR HOT FULL POWER STEAM LINE BREAK, %'ao END-0F-CYCLE (E0C) .

Reference Cycl e Cycle 2

1. Worth of all CEAs Inserted -11.3 -12.1

2. Stuck CEA Allowance +1.3 +2.3

3. Worth of all CEAs Less Highest Worth CEA Stuck Out -10.0 -9.8.

4 Full Power Dependent Insertion Limit CEA Bite +0.2 +0.3

5. Calculated Scram Worth -9.8 -9.5

6. Physics Uncertainty +.75 +1.0

7. Other Allowances (worth losses due to voiding and moderator teraperature axial redistribution) +0.2 +0 . 2

8. Net Available Scram Worth -8.85 -8.3 5

5-5 c . _ , , _

TABLE 5-3

-- WATERFORD 3 CYCLE 2 ,

REACTIVITY WORTH OF CEA REGULATING GROUPS AT HOT FULL POWER, % ao Beginning of Cycle End of Cycle Regulating Reference Reference CEAs Cycle Cycle 2 Cycle Cycle 2 Group 6 0.4 0.4 0.4 0.5 ,

Group 5 0.4 0.5 0.5 0.6 Group 4 1.0 1.0 1.0 1.1 l Note:

Values shown assume full sequential group insertion to 100%.

5-6

FIGURE 51 WATERFORD 3 CYCLE 2 PDIL FOR REGULATING GROUPS

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- ASSEMBLY RELATIVE POWER DENSITY .. .

1 0.37 0.75 I

1 1

0.48 0.99 1.19 1.12 0.86 0.43 1.13 1.14 1.17 1.08 1.32 X

, 0.43 1.12 1.09 1.29 0.96 1.31 . 0.99 0.48 1.13 1.09 1.29 0.91 1.01 0.94 1.23 0.99 1.14 1.29 0.91 1.20 0.89 1.24 0.92 i

1.18 1.16 0.96 1.01 0.89 1.18 0.99 1.16 0.37 1.12 1.07 1.30 0.94 1.24 0.99 0.89 0.86 0.75 k -

0.86 1.31 0.98 1.23 0.93 1.16 0.84 0.68 l

l X = LOCATION OF MAXIMUM 1 PIN PEAK = 1.56 LOUISLANA WATERFORD 3 CYCLE 2 Figure POWER & LIGHT CO. ASSEMBLY RELATIVE POWER DENSITY, l Waterford Steam HFP AT BOC, EQUILIBRIUM XENON, ARO 53 Electric Station i

5-9

9.

AS3EMBLY RELATIVE PO'NER DENSITY 0.40 ' O.77 1

i j

O.47 0.89 1.06 1.11 0.87 I

0.45 1.07 1.04 1.06 1.03 1.26 O.45 1.08 1.05 1.27 0.96 1.27 1.00 0.47 1.07 1.05 1.29 0.94 1.04 0.99 1.30 0.89 1.04 1.27 0.94 1.24 0.96 1.30 1.00 4

1.06 1.06 0.96 1.04 0.96 1.29 1.07 1.28 I

0.40 X

1.11 1.03 1.27 0.99 1.30 1.07 1.00 0.99 0.77

( . 0.87 1.26 0.99 1.30 1.01 1.28 0.96 0.82

! X = LOCATION OF M AXIMUM 1 PIN PEAK = 1.48 i

l l

l N

po[g U[3g gHT CO*

WATERFORD 3 CYCLE 2 ASSEMBLY RELATIVE POWER DENSITY Figure i

Waterford Steam 54

HFP AT 8 GWD/T, EQUILIBRIUM XENON, ARO l

Electric Station l 5-10

1 ASSEMBLY RELATIVE

- POWER DENSITY 0.46 .

- 0.84 -

9 O.51 0.90 1.06 1.16 0.92  ;

0.50 1.10 1.02 1.04 1.02 1.25 0.50 1.11 1.06 1.26 0.95 1.23 0.98 X

0.51 1.10 1.06 1.29 0.96 1.02 0.97 1.26 0.90 1.03 1.26 0.95 1.22 0.95 1.24 0.98 a

1.06 1.04 0.95 1.02 0.95 1.25 1.03 1.25 0.46 1.16 1.02 1.23 0.97 1.24 1.03 0.98 0.97 l 0.84

k -

0.92 1.25 0.98 1.26 0.99 1.25 0.95 0.83 l

l l X = LOCATION OF MAXIMUM 1 PIN PEAK = 1.45 l

LOU 151ANA Figure WATERFORD 3 CYCLE 2 POWER & LIGHT CO. ASSEMBLY RELATIVE POWER DENSITY Waterford Soom 5,5 HFP AT EOC, EQUILIBRIUM XENON, ARO Electric 9etion

5-11

f

, I ASSEMBLY RELATIVE POWER DENSITY

/ '

/ ,

LOCATION OF PLCEAS O.38 - -

0.75 -

//j (75% INSERTED) 1 0.46 0.99 1.18 1.13 0.87 X

0.43 1.13 1.15 1.18 1.10 1.32 0.43 1.12 1.10 1.28 0.96 1.29 1.00 VN/

0.48 1.13 1.10 1.28 0.90

/p/A

/

0.93 0.93 1.22 0.99 1.15 1.28 0.90 1.17 0.88 1.23 0.93 1.18 1.18 0.95 0.93 0.88 1.18 1.02 1.18 0.38 i 1.12 1.09 1.29 0.93 1.23 1.02 0.93 0.90 0.75

9. -

0.87 1.32 0.99 1.21 0.94 1.18 0.89 0.73 X = LOCATION OF MAXIMUM 1 PIN PEAK = 1.57 i

i LOU 15IANA Figure WATERFORD 3 CYCLE 2 POWER & LIGHT CO ASSEMBLY RELATIVE POWER DENSITY 56 Waterford Steam HFP AT BOC, EQUILIBRIUM XENON.PLCEAS Electric Station 5-12

1 ASSEMBLY RELATIVE

- POWER DENSITY 0.40 ' O.78 LOCATION OF BANK 6 (100% INSERTED) 1 0.52 1.06 1.24 1.16 0.89 4 0.47 1.22 1.22 1.22 1.10 1.30 i

0.47 1.22 1.19 1.37 0.98 1.21 0.86 0.52 1.22 1.19 1.38 0.96 1.00 0.80 i

1.06 1.22 1.37 0.96 1.22 0.87 1.11 0.77 1.24 1.22 0.98 1.00 0.87 1.14 0.95 1.09 0.40 1.16 1.09 1.21 0.80 1.11 0.95 0.88 0.85 0.78

$ -- 0.89 1.29 0.85 0.78 1.09 0.84 0.69 X = LOCATION OF MAXIMUM 1-PIN PEAK = 1.67 i

I LOU 151ANA Figure WATERFORD 3 CYCLE 2 l POWER & LIGHT CO* 57 ASSEMBLY RELATIVE POWER DENSITY Waterford Steam HFP AT BOC, EQUILlBRIUM XENON,WITH BANK 6 I

Electric Station 1

5-13

LOCATION OF PLCE AS (75% INSERTED) '

ASSEM8LY RELATIVE POWER DENSITY LOCATION dF BANK 6 0.41 . . 0.81 '

(100% INSERTED) 9 0.53 1.08 1.27 1.19 0.91 0.48 1.24 1.24 1.24 1.11 1.32 X

0.48 1.24 1.21 1.37 0.97 1.21 0.86 0.53 1.24 1.21 1.37 0.94 0.91

/ 0.78 .64

/

1.08 1.24 1.37 0.94 1.18 0.84 1.08 0.76 1.27 1.24 0.97

//

0.90 0.84 1.12 0.94 1.08 0.41 1.19 1.11 1.20 0.77 1.08 0.94 0.88 0.85 T. -

0.91 1.31 0.85 0.77 1.08 0.84 0.69 j X = LOCATION OF MAXIMUM 1 PIN PEAK = 1.69 l

LOU 151ANA WATERFORD 3 CYCLE 2 Fi8ure POWER & LIGHT CO. ASSEMBLY RELATIVE POWER DENSITY Waterford Seom HFP AT BOC, EQUILIBRIUM XENON, WITH 58 1 Electric 9ation BANK 6 AND PLCEAS

.; 5-14 i

1 ASSEMBLY RELATIVE POWER DENSITY O.47 . 0.87 -

LOCATION OF PLCEAS (75% INSERTED) 1 0.52 0.92 1.08 1.19 0.95 0.51 1.12 1.04 1.06 1.03 1.26 0.51 1.13 1.06 1.26 0.94 1.23 0.98 0.52 1.12 1.06 1.29 0.93 NNy2 0 0.95 1.25 h.9\

0.92 1.04 1.26 0.93 1.18 0.92 1.22 0.97 1.08 1.06 0.94 0.92 0.92 1.24 1.03 1.25 0.47 \

1.19 1.03 1.23 0.95 1.22 1.03 0.98 0.97 0.87

$ - 0.95 1.26 0.98 1.25 0.98 1.25 0.95 0.83 X = LOCATION OF MAXIMUM 1 PIN PEAK = 1.46 i

i i

LOU 151ANA WATERFORD 3 CYCLE 2 Figure l

j POWER & LIGHT CO. ASSEMBLY RELATIVE POWER DENSITY

Waterford Steam HFP AT EOC, EQUILIBRIUM XENON,WITH PLCEAS 59

Electric Station J

5-15

1 ASSEMBLY RELATIVE

- POWER DENSITY

.9 - 0.M -

LOCATION OF BANK 6 (100% INSERTED) 1 0.56 0.98 1.14 1.23 0.97 0.55 1.21 1.11 1.10 1.03 1.24 0.56 1.23 1.15 1.35 0.96 1.14 0.83 i

0.56 1.21 1.15 1,39 0.99 0.99 0.80 0.98 1.11 1.35 0.99 1.24 0.91 1.10 0.80 i

1.14 1.10 0.96 0.99 0.91 1.21 0.97 1.16 7

0.49 1.23 1.03 1.14 0.80 1.10 0.97 0.94 0.93 0.91 1 -

0.97 1.24 0.83 0.80 1.16 0.91 0.80 X = LOCATION OF MAXIMUM 1 PIN PEAK = 1.58  ;

WATE'RFORD 3 CYCLE 2 Figure POW R L GHT CO* ASSEMBLY RELATIVE POWER DENSITY a or m 5 10

HFP AT EOC, EQUILl8RIUM XENON WITH BANK 6 '

5-16

LOCATION OF PLCEAS (75% INSERTED) ASSEMBLY RELATIVE '

- POWER DENSITY 0.51 . 0.93 .

LOCATION OF BANK 6 (100% INSERTED) 4 0.58 1.00 1.16 1.26 1.00 1 0.57 1.23 1.12 1.11 1.05 1.26 0.57 1.25 1.16 1.34 0.95 1.13 0.83 i

0.58 1.23 1.16 1.38 0.97 0.78 1.00 1.13 1.34 0.97 1.20 0.89 1.08 0.79 1.16 1.11 0.95 0.89 1.19 0.97 1.16 0.51 '

1.26 1.06 1.13 0.78 1.08 0.97 0.94 0.94 O.93 T. -

1.00 1.26 0.83 0.79 1.16 0.92 0.81 X = LOCATION OF MAXIMUM 1 PIN PEAK = 1.59 LOUISLANA WATERFORD 3 CYCLE 2 i POWER & LIGHT CO, ASSEMBLY RELATIVE POWER DENSITY p;8"

Waterford Seem HFP AT EOC, EQUILIBRIUM XENON, WITH 5 11 Electric 9ation BANK 6 AND PLCEAS 5-17 l

6.0 THERMAL-HYORAULIC DESIGN 6.1 DNBR ANIlYSIS ,,

Steady state DNBR analyses of Cycle 2 at the rated power level of 3390 MWT have been performed using the TORC computer code described in Reference 6-1, the CE-1 critical heat flux correlation described in Reference 6-2, the simplified TORC modeling methods described in Reference 6-3, and the CETOP code describe in Reference 6-4 Table 6-1 contains a list of pertinent thermal-hydraulic design parameters. The Statistical Combination of Uncertainties (SCU) methodology presented in Reference 6-5 was applied with Wat e'rfo rd-3 specific data using the calculational factors listed in Table 6-1 and other uncertainty factors at the 95/95 confidence / probability level to define a design limit of 1.26 on CE-1 minimum DNBR This Cycle 2 DNBR limit includes the following allowances:

1. inC specified allowances for TORC code uncertainty and CE-1 CHF l correlation cross validation uncertainty as discussed in Reference 6-10.

2. NRC imposed 0.01 DNBR penalty for H10-1 grids as discussed in References 6-6, 6-7 and 6-8.

3. Rod bow penalty as discussed in Section 6.2 below.

6.2 EFFECTS OF FUEL ROD BOWING ON ONBR MARGIN Effects of fuel rod bowing on DNBR margin have been incorporated in l the safety and setpoint analyses in the manner discussed in References 6-5 and 6-9. The penalty used for this analysis is~

valid for bundle burnups up to 30,000 tND/MTU. This penalty is included in the 1.26 DNBR limit.

S 6-1

6.0 THERMAL-HYORAULIC DESIGN

~

6.1 DNBR ANALYSIS .-

Steady state DNBR analyses of Cycle 2 at the rated power 1.evel of 3390 MWT have been performed using the TORC computer code described in Reference 6-1, the CE-1 critical heat flux correlation described in Reference 6-2, the simplified TORC modeling methods described in Reference 6-3, and the CETOP code described in Reference 6-4 Table 6-1 contains a list of pertinent thermal-hydraulic design parameters. The Statistical Combination of Uncertainties (SCU) methodology presented in Reference 6-3 was applied with Wate'rford-3 specific data using the calculational factors listed in Table 6-1 and other uncertainty f actors at the 95/95 confidence / probability level to define a design limit of 1.26 on CE-1 minimum DNBR This Cycle 2 DNBR limit includes the following allowances:

1. NRC specified allowances for TORC code uncertainty and CE-1 CHF correlation cross validation uncertainty as discussed in Reference 6-10.

2. NRC imposed 0.01 DNBR penalty for HID-1 grids as discussed in References 6-6, 6-7 and 6-8.

3. Rod bow penalty as discussed in Sectio,n 6.2 below.

6.2 EFFECTS OF FUEL R00 BOWING ON ONBR MARGIN Ef fects of fuel rod bowing on DNBR margin have been incorporated in the safety and setpoint analyses in the manner discussed in References 6-5 and 6-9. The penalty used for this analysis 17

~

valid for bundle burnups up to 30,000 MWO/MTU. This ' penalty is included in the 1.26 OhBR limit.

=

6-1

For assemblies with burnup greater than 30 GWD/T sufficient available margin exists to offset rod bow penalties due to the lower radial 4 power peaks in these higher burnup batches. Hence the rod bow penalty based uptm Reference 6-9 for 30 GWD/T is applicable for al1 assembly - ,

burnups expected for Cycle 2.

l 6-2

- ---r - -. _, , . - , - - - . ---------,..--r--.w-_- ---.,--.-,.% ,..-,..,,s - - _ _ . _ . . - -, . - - - . - - , - - - - - - . - - ,

TABLE 6-1 WSES-3 Cycle 2 Ibermal Hydraulic Parameters at Full Power ,

Reference General Characteristics Units Cycl e Cycle 2 Total Heat Output (Core only) MWg 3390 3390 10 Btu /hr 11,570 11,570 Fraction of Heat Generated in --

0.975 0.975 Fuel Rod Primary System Pressure psia 2250 2250 Nominal Inlet Temperature (Nominal) 0F 553.0 553.0 Total Reactor Coolant Flow 396.000 396,0'00 (Minimum Steady State) gpglb/hr 10 148.0 148.0 Coolant Flow Through Core (Minimum) 106 !d/hr 144.2 144.2 Hydraulic Diameter (Nominal Channel) ft 0.039 0.039 6 2 Average Mass Velocity 10 lb/hr-ft 2.64 2.64 Pressure Drop Across Core (Mininum psi 15.4 15.4 steady state flow irreversible P over entire fuel assembly)

Total Pressure Drop Across Vessel psi 41.8 41.8 (Based on nominal dimensions and minimum steady state flow) 2 Core Average Heat Flux (Accounts BTU /hr-ft 182,400 182,700

for fraction of heat generated in f uel rod and axial densifica-tion factor) 2 Total Heat Transfer Area (Accounts ft 61,900*** 61,700*

i for axial densification factor)

Film Coefficient at Average BTU /hr-ft20F 6200 6200 Conditions O

Average Film Temperature Difference F 29.2 29.3 Average Linear Heat Rate of Unden- kw/ft 5.34 5.34 sified Fuel Rod (Accounts for .

fraction of heat generated in fuel rod)

Average Core Enthalpy Rise BTU /lb 80.3 80.3 6-3 4

- -_r- - . - , . , . . _ . - - . _. ,._ ,, , _ , - . _ - _ . . . _ - . _ - . - . , - _ , . - . . . _ , _ _ -

TABLE 6-1 (continued)

Reference Calculational Factors Unit Cycl e Cycle 2 Maximum Clad Surface Temperature CF 656.7 656.7 Engineering Heat Flux Factor 1.03 1.03**

Engineering Factor on Hot Channel 1.03 1.03**

Heat Input Rod Pitch, Bowing and Clad Diameter 1.05 1.05**

Factor Fuel Densification Factor (Axial) 1.002 1.002 1

NOTES: ,

• Based on 1728 poison rods.

• • These factors have been combined statistically with other uncertainty f actors at 95/95 confidence / probability level to' define a new design limit on CE-1 minimum DNBR when iterating on power as discussed in Reference 6-5.

• • • Based on 1632 poison rods.

e 6-4 I

W e

0 9 7.0 TRANSIENT ANALYSIS 8.0 ECCS ANALYSIS (THESE SECTIONS WILL BE PROVIDED LATER IN THE FINAL RELOAD ANALYIS REPORT)

=- .-- . - - . . _ .

9.0 REACTOR PROTECTION AND MONITCRfNG SYSTEM

9.1 INTRODUCTION

The Core Protection Calculator System (CPCS) is designed to provide the .

low DN8R and high Local Power Density (LPD) trips to (1) ensure that the

^

specified acceptable fuel design limits on departure from nucleate boiling and centerline fuel melting are not exceeded during Anticipated Operational Occurrences (A00s) and (2) assist the Engineered Safety Features System in limiting the consequences of certain postulated accidents.

The CPCS in conjunction with the balance of the Reactor Protection System (RPS) must be capable of providing protection for certain specified design basis events, provided that at the initiation of these occurrences the Nuclear Steam Supply System, its sub-systems, components and parameters are maintained within operating limits and Limiting Conditions for Operation (LCOs).

9.2 CPCS SOFTWARE M00!FICATIONS The CPC/CEAC sof tware for WSES-3 is being modified for operation in Cycle 2. This modification is being made to implement the CPC Improvement Program (CIP) including algorithm and plant-specific data base changes, changes to the list of addressable constants and implementation of Reload Data Block (RDB).

The CPC/CEAC algorithms for WSES-3 Cycle 2 are the same as those implemented at SONGS-2 and 3 (Cycl e 3) and at ANO-2 Cycle 6 and

, described in References 9-1 and 9-2. The revised list of addressable constants are defined in Reference 9-3. The Rel oad Data Block (Reference 9-5) is a group of constants that are located in protected memory of the CPC and the CEAC, separate f rom other non-addressable constants. The RDB constants are loaded from a separate RDB disk and can be changed without requiring a CPC/CEAC sof tware change.

4 l 9-1

I 1

j 1

l The modifications for WSES-3 Cycle 2 relative to the Cycle i software are described in References 9-3, 9-4, 9-5 and 9-9. The modifications described in References 9-3, 9-4, and 9-9 are incorporated in References 9-1 and 9-2. The implementation of all changes will be done in -

accordance with the established software change procedures, References 9-6 and 9-7.

Cycle dependent values of the data base and RDB constants will be detennined for WSES-3 Cycle 2 consistent with the cycle design, performance and safety analyses. The RDB constants will be installed on the Cycle 2 RDB Disk for loading at the site as described in Reference 9-8.

9.3 ADDRESSABLE CONSTANTS .

Certain CPC constants are addressable so that they can be changed as required during operation. Addressable constants include (1) i constants that are measured during startup (e.g., shape annealing matrix, boundary point power correlation coefficients, and adjustments for CEA shadowing and planar radial peaking f actors), (2) uncertainty factors to account for processing and measurement uncertainties in DNBR and LPD calculations (BERRO through BERR4), (3) trip setpoints and (4) miscellaneous items (e.g., penalty f actor multipliers, CEAC penalty f actor time delay, pre-trip setpoints, CEAC inoperable flag, calibration constants, etc.).

l l

Trip setpoints, uncertainty f actors and other addressable constants have been determined for WSES-3 Cycle 2 consistent with the sof tware l and methodology established in the CIP (Reference 9-3, 9-4 and 9-5).

j The CIP methodology includes the application of statistical combination of uncertainties.

9.4 O!GITAL MONITORING SYSTEM (COLSS) l l

l The Core Operating Limit Supervisory System (COLSS) is a ' monitoring system that initiates alarms if the LCO on DNBR, peak linear heat rate ,

l core power, or core azimuthal tilt are exceeded. The COLSS data base i

l and uncertainties will be updated to reflect the Cycle 2 core design, i

9-2

10.0 TECHNICAL SPECIFICATIONS (THIS SECTION WILL BE PROVIDED LATER IN THE FINAL RELOAD ANALYSIS REPORT).

4

l1.0 l STARTUP TESTING The planned startup test program associated with core performance is outlined below. These tests verify that core performance is consistent with the Engineering design and safety analysis. Some of th,e,, tests also provide the data needed for adjustment of addressable constants on the Core Protection Calculators (CPCs).

11.1 PRE-CRITICAL TEST 11.1.1 Control Elemen t Assembly (CEA) Trip Test Pre-critical CEA drop times are recorded for all full length CEAs at hot, full flow conditions. The drop times will be verified to be within Technical Specification limits.

II.2 LOW POWER PHYSICS TESTS II.2.1 initial Criticality Initial criticality is obtained by fully withdrawing all CEA Groups except Group 6 (which is withdrawn to approximately 75 inches), then diluting the Reactor Coolant System (RCS) until the reactor is critical, l1.2.2 Critical Boron Concentration (CBC)

The CBC is obtained for the All Rods Out (ARC) condition and for a partia!!y rodded configuration. Comparison to the predicted CBC is performed by compensating for the residual CEA worth (f rom the actual CEA position to the predicted CEA position). The measured CBCs will be verified to be within the equivalent of 1% DeltaK/K of the design predictions, 11.2.3 Temperature Reactivity Coefficient The isotherral temperature coefficient (lTC) is measured at the Essentially Ali Rods Out (EAR 0) configuration and at a partially rodded configuration. The average coolant temperature is varied and,the re-activity feedback associated with the temperature change is measured.

11-1

The measured value will be verified to be within 2 0.3 x 10-4 Del ta/K/ K/degF of the predicted value.

The moderator temperature coefficient (MTC) is calculated by subtracting the predicted value of the fuel temperature coefficient from.the ITC. The MTC value is tehn verified to be within Technical Specification limits.

11.2.4 CEA Reactivity Worth CEA worths will be measured using the CEA Exchange technique. This technique consists of measuring the worth of a " Reference Group" via standard boration/ dilution techniques, then exchanging this group with other groups to measure their worths. All full-length CEAs will be in-cluded in the measurement groups. Due to the large differences in relative CEA group worths, two reference groups (one with very high worth and one with medium worth) may be used. The groups to be measured will be exchanged with the appropriate reference group, depending on their predicted worth.

4 The individual measurement group worths will be verified to be within 15% or 0.1% DeltaK/K (whichever is larger) of predicted values. In

{ addition, the total worth of all the measurement g oups will be verified to be withia 10% of the predicted total worth.

l l1.3 POWER ASCENSION TESTING

Following completion of the Low Power Physics Test sequence, reactor power will be increased in accordance with normal operating procedures.

The power ascension will be monitored by an off-line NSSS performance and data processing computer algorithm. This computer code will be continuously executed in parallel with the power ascension to monitor I

CPC and COLSS performance relative to the processed plant data against i which they are normally calibrated. if necessary, the power ascension will be suspended while necessary data reduction and equipment call-

brations are performed. Thus the monitoring algorithm continuously a

! ensures conservative CPC and COLSS operation whl:e optimizing overall efficiency of the test program.

• 1 1

1 a

ll-2 i

1 W

l lI .3. I ' Reactor Coolant Flots Reactor coolant flow will be measured by calorimetric methods at steady

i. state conditions in accordance with Technical Specifications. Acceptance criteria sill require that the COLSS Indicated flow be conservative with

respect to the measured flow and that the CPC' indicated flow le con-servative with respect to the COLSS indicated flow.

I 11.3.2 Fuel Symmetry Verification Fixed incore detector' data will be examined at low power to verify that

no detectable fuel mistoadings exist. Individual instrumented fuel  !

10% of the symmetric assembly powers will be verified to be within group average.

11.3.3 Core Power Distribution

(.

1 Core Power distribution data using fixed incore neutron detectors will be

! used to further verify proper fuel loading and to verify consistency

]. between the as-buiit core and the engineering design modeis. This is i accomplished using measurement data from two power plateaus.

I l Compliance with the acceptance criteria at the interrediate power plateau (between 40 and 70% reactor power) gives reasonable assurance that the I

power distribution will remain within the design limits while reactor power is increased to 100%.

i The final power distribution comparison is performed at full power.

Axial and radial power distributions are compared to design predictions j- as a final verificatien that the core is operating in a manner consistent with its design within the associated design uncertainties.

I l The measured results are compared to the predicted values in the following ,

1 j manner.for both the intermediate and the full power analyses:

A. The root-mean-square (RMS) error between the measured and predicted

{ radial power distribution for each of the 217 fuel assemblies will

. be verifled to be less than or equal to 0.05. -

4 11-3 i

i i  :

r-

8. The RMS error between the measured and predicted axial power distribution for each of the 51 axial nodes.will be. verified to be less than or equal to 0.05. "' )

C.

The n7asured values of planar radial peaking factor (Fxy ),e integrated radiaj peaking factor (Fr), core average axial peak (Fz ),

and the 3-D power peak (F q ) will be verified to be within 10% of their predicted values. i 11.3.4 Shape Annealing Matrix (SAM) and Boundary Point Power Correlation Coefficients (BPPCC) Verification The SAM and BPPCCs are determined from a linear r2gression analysis of the measured excore detector readings end corresponding core power

. 3 distribution determined from the incore detector signals. 1Since these values must be representative for a. rodded and unrodded core.throughout the cycle, it is desirable to use aE wide a range of axial shapes'as are available to establish their values. The spectrum of axial shapes encountered during the power ascension has been demonstrated to be '

adequate for the calculation of the matrix elements.

The incore, excore, and relaTeo data is compiled and analyzed throughout the power ascension W the of f-line NSSS performance and data processing algorithm. The results. of tile analysis are used to modify the ap-propriate CPC constants if necessary.

11.3.5 Radial Peaking Facter (RPF) and CEA Shadowing Factor (RSF) herification The RPFs and the RSFs are calculated using fixed incore detector an,d excore detector data from the following CEA configurations:

- AlI Rods Out

- Group 6 fully inserted

- Group 6 fully inserted & PLCEAs @ 37.5 in, withdrawn

- PLCEAs @ 37.5 in, withdrawn Appropriate CPC and/or COLSS constants are modified based on the measured values. The rodded portions of this test may be deleted from the test program if appropriate margin penalties are incorporated into'the CPC and COLSS uncertainty constants.

11-4

11.3.6 . Temperature Shadowing' Factor Verification Excore detector response as a function of RCS cold leg temperature .

during the power ascension will be analyzed by the off-line'NSSS per-formance 2' ode to verify the adequacy of the CPC Temperature. Shadowing Factor constants.

3- II.3.7 Reactivity Coefficients at Power W-

^

The isothermal temperature coef ficient (lTC) is measured at approxi-mately full power by swinging turbine load to alternately increase and decrease core inlet temperature. The swings in core temperature and power are used along with the predicted power coefficient to calculate j the ITC. The predicted fuel temperature coefficient is then subtracted from the ITC to obtain the MTC. The measured MTC is then used to I

verify compliance with the Technical Specifications.

11.4 PROCEDURE IF ACCEPTANCE CRITERIA ARE NOT MET The results of all tests will be reviewed by the plant's reactor

- engineering group. If the acceptance criteria of the startup physics tests are not met, an evaluation will be performed by the plant's reacter engineering group with assistance from the fuel vendor, as needed. The

, results of this evaluation will be presented-to the Plant Operations Review Committee. Resolution will be required prior to power escalation.

, if an unreviewed safety question is involved, the NRC would be notified.

t

\l 3

4 7

I 4

I s

[ i i-5 v.,- , , ,, ngn----,~ ,

,w--v, ,n_,-----~,,...n- ,,_-w.-w,,.-, . - - g-g r-- n e,, ww n y wn,,w,-~ n,., m _ w ~- e m m- .m.,,v-

12.0 References 12.1 Section5.0 References .-

(1-1) "Waterford Steam Electric Station, Unit No. 3 Final Safety Analysis Report," Docket No. 50-382.

12.2 Section 2.0 References None 12.3 Section 3.0 References None 12.4 Section 4.0 References (4-1) "Waterford Steam Electric Station, Unit No. 3, Final Safety Analysis Report," Docket No. 50-382.

(4-2) (Deleted)

(4-3) C. O. Thomas (NRC), to A. E. Scherer (C-E), " Acceptance for Referencing of Licensing Special Report LD-84-043, CEA Guide Tube Wear Sleeve Modification," September 7, 1984 (4-4) (Deleted)

(4-5) EPRI NP-3956-CCM, "CEPAN Method of Analyzing Creep Collapse of Oval Cladding Volume 5: Evaluation of Interpellet Gap Formation and Clad Collapse in Modern PWR Fuel Rods," April, 1985.

12-1

. i (4-6) (Deleted)

(4-7) CENPD-139-P-A, "C-E Fuel Evaluation bbdel Topical Report,"

July, 1974.

(4-8) CEN-161(B)-P, "improvrements to Fuel Evaluation Model," July, 1981.

(4-9) R. A. Clark (NRC) to A.E. Lundval l, Jr. (BG&E), " Safety Evaluation of CEN-161 (FATES 3)," March 31, 1983.

12.5 Section 5.0 References (5-l) EPRI NP-3966-CCM, "CEPAN Method of Analyzing Creep Collapse of Oval Cladding, Volume 5: Evaluation of Interpellet Gap Formation and Clad Collapse in Modern PWR Fuel Rods," April, 1985. ,

(5-2) (Deleted)

(5-3) (Deleted)

(5-4) MSS-NA3-P. " Verification of CECOR Coefficient Methodology for Application to Pressurized Water Reactors of the Middle South Util ities System", August I, 1984.

(5-5) CENPD-266-P-A, "The ROCS and DIT Computer Codes for Nuclear Design," April, 1983.

12-2

12.6 Section 6.0 References

~

(6-1) CENPD-161-P-A, " TORC Code, A Computer Code for ,0etermining the Thennal Margin of a Reactor Core," April,1986.

(6-2) CENPD-162-P-A, " Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids, Part 1 Uniform Axial Power Distribution," September,1976.

(6-3) CENDP-206-P-A, " TORC Code, Verification and Simplified Modeling Methods ," June,1981.

(6-4) CEN-160(S)-P, Rev. 1-P, "CETOP Code Structure and Mo,deling Methods for San Onofre Nuclear Generating Station Units 2 and 3," September, 1981.

(6-5) CEN-283(S)-P, " Statistical Combination of Uncertainties, Part 1, Combination of System Parameter Uncertainties in Thermal Margin Analyses for San Onof re Nuclear Generating Station Units 2 and 3," June, 1984 (6-6) CEN-155-(S)-P, "CE-1 Applicability to San Onof re Units 2 and 3 HID-2 Grids, Response to NRC Questions," March,1981.

(6-7) CEN-165(S)-P, " Responses to NRC Concerns on Applicability of the CE-1 Correlation to the SONGS Fuel Design," May,1981.

(6-8) NUREG-0787, Supplement 1. " Safety Evaluation Report Related to the Operation of Waterford Steam Electric Station, Unit No. 3," Docket No. 50-382, October, 1981.

(6-9) CENPD-225-P-A, " Fuel and Poison Rod Bowing," June 1983.

(6-10) Robert A. Clark (NRC) to William Cavanaugh III, (AP&L),

" Operation of ANO-2 During Cycle 2," July 21, 1981 (Safety Evaluation Report and License Amendment No. 26 for ANO-2).

12-3

9 12.7 Section 7.0 References (later) 12.8 Section 8.0 References (Later) 12.9 Section 9.0 References (9-1) CEN-304-P, Rev. 01-P, " Functional Requirments for a Control Element Assembly Calculator," May,1986.

(9-2) CEN-305-P, Rev. 01-P, " Functional Requirements for a Core Protection Calculator," May, 1986. .

(9-3) CEN-308-P-A, "CPC/CEAC Sof tware Modifications for the CPC Improvement Program," April,1986.

(9-4) CEN-310-P-A, "CPC and Methodol ogy Changes for the CPC

?mprovement Program," April,1986.

(9-5) CEN-3330-P, Rev. 00-P, "CPC/CEAC Sof tware Modifications for the CPC Improvement Program Reload Data 81ock," May,1986.

(9-6) CEN-39(A)-P, Rev. 03, "CPC Protection Algorithm Software Change Procedure," January,1986.

(9-7) CEN-39(A)-P, Supplement 1-P, Rev. 03-P, "CPC Protection Algorithm Software Change Procedure Supplement 1," April, 1986.

(9-8) CEN-323-P, " Reload Data Block Constant Installation Guidelines ," February,1986. .

(9-9) CEN-281(S)-P, "CPC/CEAC Software Modifications for San Onofre Nuclear Steam Generating Station Units No. 2 and 3,"

July , 1984.

12-4

- -- ._-____ =__ - - - _- . .____ -_ - -_ - ___- - _ - _ _ _ _ _ _ _ - _ . _ _ . _ _ _ _ _ _ _ .

. .g.

12.10 Section 10.0 References (later) 12.11 Section 11.0 References None I

I 12-5

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