ML20095H935
| ML20095H935 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 04/30/1992 |
| From: | ENTERGY OPERATIONS, INC. |
| To: | |
| Shared Package | |
| ML20095H930 | List: |
| References | |
| NUDOCS 9204300366 | |
| Download: ML20095H935 (76) | |
Text
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RELOAD ANALYSIS REPORT FOR WATERFORD 3 CYCLE 6 APRIL 1992 4
9204300366 920430 PDR P
ADOCK 05000382 PDR
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RELOAD ANALYSIS REPORT FOR WATERFORD 3 CYCLE 6 !
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MAY 1992 i
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TABLE OF CONTENTS *
- 1. INTRODUCTION MID
SUMMARY
1
- 2. CPERATING' HISTORY OF THE CURRENT CYCLE
- 3. GENERAL DESCRIPTION ,
- 4. FUEL SYSTEM DESIGN
- 5. NUCLEAR DESIGN [
6, t THERMAL-HYDRAULIC OESIGN l
7 NON-LOCA SAFF.TY ANALYSIS ,
- 3. ECCS ANALYSIS f
- 9. REACTOR-PROTECTION AND MONITORING SYSTEM i
- 10. ' TECHNICAL 3PECIFICATICNS -i '
- 11. SIARTUP TESTING i
- 12. REFERENCES e
APPENDIX A - EVALUATION OF CHANGE TO NUCLEAR DESIGN METHODS !
APPENDIX B - DEBPIS RESISTANT FUEL DESIGN
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LIST CF TABLES .
3-1 Cycle b Core Loading 5-1 Cycle 6 Nominal Physics Parameters 5-2 Cycle 6 Limiting Values of Reactivity Worths and Allcwances for Hot Full Power Steam Line Break 5-3 Cycle 6 Reactivity Worth of CEA Regulating Groups at Hot Full Power 6-1 Cycle 6 Thermal Hydraulic Parameters at Full Power 7,0-1 Design Basis Events Ctasidered in the Cycle.6 Safety Analysis 7.0-2 DBE's Evaluated with Respect to Offsite Dose Criterien 7.0-3 DBE's avaluated with Respect to RCS Pressure Criterien 7.0-4 DBE's Evaluated with' Respect to Fuel Performance 7.0-5 DBE's Evaluated with Respect to Shutdcen Margin Criterion-7.0-6 Core Parameters Input to Safety Analysis 7.1,3-1 Sequence of Events for the Increased Main Steam Flow in Jombination with a Loss of AC Power i 3-1 Cycle 6 ECCS Analysis Significant System Parameters 3-2 Cycle 6 ECCS Analysis Significant Fuel Pin Parameters LISTHCF FIGURES h V 6'& s L*,
3-1 Cycle 6 Fresh Fuel 3-2 Cycle 6 Quarter Core Loading Pattern 3-3 Cycle 6 Assembly Averaged Burnup 3-4 CEA Bank Identification 3-5 In-Core Instrument Assemblies - Core Locations '
5-1 PDIL for Regulating Groups 5-2 Part Length CEA Insertion Limit vs Thermal Power 5-3 Assembly Relative Power Density, HFP at DOC,'Unrodded 5-4 Assembly Relative Power Density, HFP at MOC, Unrodded 5-5 Assembly Relative Power Density, HFP at EOC, Unroaded
- 5-6 Assemoly Relative Power Density, HFP at BOC with PLCEA's 5-7 Assembly Relative Power Density, HFP at BOC with Bank 6 5-8 Assembly Relative Power Density HFP at BOC with' Bank 6 and -
PLCEA's 5-9 Assem.bly Relative Power Density, HFP at EOC_with PLCEA's 5-10 AssemDly Relative Power Density, HFP at EOC with Bank 6 5-11 Assembly Relative- Power Density, HFP at-EOC with Bank 6 and PLCEA's
-S-1 ECCS' Analysis, PCT versus Time for the Hot Spot Location l
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1.0 INTRODUCTION
-AND
SUMMARY
i This report provides an evaluation of the design and performance of Waterford Steam Electric Station Unit 3 during its sixth-cycle of_ operation at 100%-
rated core power cf 3390 MWt ana NSSS power of 3410 MWt. Operating conditions for Cycle 6 are assumed to be consistent with those of previous Cycles and are i summarized as full power operation under base load conditions. The core.will consist of irradiated Batch C, '!
F, and G assemblies, along with fresh Batch H 'I assemblies. The Cycle 5 termination burnup has been assumed to be between 427 and 457 Effective Full Power Days (EFPD).
The safety criteria (margins of safety,. dose limits, etc...) applicable to j Waterford 3 were established in the FSAR (Reference 1-1) for Cycle 1. A review ~
of all postulated cccidents and anticipated. operational occurrences (AOO's) '
was performed for Cycle 2, resulting in revisions to several safety analysis-events, and for cycles 3, 4, and 5, ;
resulting in a negative 10CFR50.59 !
licensing finding. The-FSAR as amended.with information'from the Cycle 2. [
Reload Analysis Report (References 1-2 and11-3) constitutes the analyses of j record for Waterford 3.
The Cycle 6 reload core characteristics have 'n1 evaluated with respect-to 4
j the Reference Cycle. Specific differences-in core fuel. loadings have been accounted for in the present analysis. The Cycle 6 analysis results for .[
postulated accidents and AOO's are summarized in Sections 7 and 8. i Two transients reanalyzed for Cycle 6 met NRC acceptance criteria, but I
exceeded the consequences determined in the Reference Cycle 1 analyses ;
documented and (2) in the Waterford 3 FSAR. These transients are . (1) Large Break LOCA-
-l Excess Luad). Excess Steam Demand with Loss of Offsite Power (also referre'd to as ,
Because the consequences of these two events exceeded that of the Reference Cycle, .
the Cycle-6 Reload Analysis Report is being submitted.for NRC-review. '
All.other transients resulted in consequences.which were: bounded by the FSAR ~
Reference Cycle analyses.- {
S The.Waterford: I 3 Cycle 6 reload required two methodology. changes. The high burnup methodology of Reference 4 provides justification for peak : rod [
bvrnups 58,700 MND/T.
of up to 60,000 MWD /T!-the= peak, cod burnup for:Cy'cle 6 is-In addition, the fission gas release model tsed will be the NRC approved version of FATES 3B-(Reference 1-5)'. l' b
L The Cycle 6 reload required no Technical Specification' changes.
L Application of the Modified Statistical Combination of' Uncertainties (MSCU)-
methodology (Reference l-b) is anticipated for Cycle 6. Information~ presented. "
in the following sections of this report will'remainivalid with.the.
application'of_MSCU to Cycle 6, except that Reference.6-5;would be revised to ._ _
l the MSCU report, .;
Reference 1-6. Application of.the MSCU methodology would. ;
preserve-the operating margins. assumed in.the Cycle 6. safety analyses. ",
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. OFF. RATING HISTORY CF THE CURRENT--CYCLE t Waterford 3 is currently in its fifth fuel cycle which began with initial criticality on May 23, 1991. Full power operation was achieved on June 2, 1991. '
It is presently estimated that Cycle 5 will terminate-by September 18, 1992. !
The Cycle 5 terminaticn poir.t can vary between 427 EFPD and 457 EFPD to f
accommodate the plant schedule and still be within-the burnup assumptions of !
the cycle 6 analyses.
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3.0 GENERAL DESCRIPTION The. Cycle 6 core will consist of those assembly typesiandinumbers-listed in Table 3-1, one Batch C1 assembly, forty-eight Bttch_E' assemblies, and thirty- .
six 84 Batch F assemblies will be removed from the Cycle 5 core;to make way for fresh Batch H assemblies plus one previously discharged C1 assembly. All eighty-four Batch G assemblies and forty-eight Batch F' assemblies now in the core will be retained..One twice burned Batch C1 assembly' discharged at the-End of Cycle 2 will be reinsertec.
The reload batch will consist of B type Ho asseenblies, 20 type H1 assemblieu with S burnable poison shims per assembly, and 56 type H2 assemblies withf16 curnable poison shims per assembly. These .sub-bat;ch types are zone-enriched and their configurations are shown in F.iqure 3-1.
The core,loading i pattern fer Cycle 6, showing fuel type and location in the quarter s displayed in Figure 3-2. The full core will be--loaded with quarter-core-rotational symmetry. '
rigure 3-3 displays the beginning of Cycle 6 and end-of Cycle 6 (490 EFPD) assembly average -burnup distributions. These burnup distributions are based en a Cycle 5 length of 457 EFPD.
Control element assembly patterns and in-core instrument-locations will' remain unchanged from Cycle 5 and are shown in Figure 3-4 and Figure 3-5 respectively.
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CYCLE 6 CORE LOADI!1G Numoet !nitial To t a l ,'awce r Assemo.y feel has :rl.ia. Poison Poison ef
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- 20 176 4.05 8 0.016 3520 160 52 3.65 1040
.2 48 16e 4.05 16 0.024 8064 768 52 .3.65 2496 n0 4 ~34 '..t 3 0 1472 0 32 .c5 416 20 '. 7 6 4 05 i 0.016 3520 160 52 . 65 1040-82 56 '. 6 B 4.05 ;6 0,024 9400 .996 52 3.65 2912
.0tal 21I 48864 2346 3-2
FIGUAE 3-1 WATERFORD 3 CYCLE 6 FRESH FUEL
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4 Figure 3-2 WATERFORD 3 CYCLE 6 QUARTER CORE LOADING PATTERN X A - Fuel Batch X - QC Location (Cycle 6) 1 2 F2 H0 3 4 5 16 7 F1 F1 G1 '
H1 G2 8 9 10 11 12 13 F1 H1 GO H2 G2 H2 14 15 16 17 18 19 20 F2 HI -G2 H2 G1 H2 G2 21 22 23- 24 25- 26 27 28 F1 H1 G2 GO- G2 H2 F0 H2 29 30 31 32 33 34 35 36 F1 G0 H2 G2 GO F0 H2 G2 37 38 39 40 41 42 43 44 G1 H2 Gl~ H2 '
F0 H2 G2 H2 F2 46 47 48 49 50 51 52 53 g v.1 G2 H2 F0 H2 G2 G1- G2 H0 55 56 57 58 59- 60 61 6E*
G2 H2 G2 H: -G2 H2 G2 Cl REINSERTED FROM CYCLE 2 l
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Assembly Averaged Burnups (EOC5 457 EFPD, E006 - 490 EFPD) 1 F2 2 HQ
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37300 0 43900 14500 t
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35100 36900 18000 0 22300 l 40000 44300 31700 21400 41400 i 8 F1 9 H1 10 40 11 H2 12 G2 13 H2 36300 0 15100 0 22800 0 "
42700 16300 32700 23100 43600 25300 :
14 F2 15 H1 16 G2 17 H2 18 G1 19 H2 20 G2 ,
39400 0 23200 0 17300 0 21800 45600 16900 41100 24400 39500 25700 1
43600 21 F1 22 H1 23 G2 24 GO 25 G2 26 H2 27 F0 28 H2 #
l 35100 0 23200 12200 21300 0~ 34400 0 l 40000 16100 41100 33300 42500 25100 52200 25200
?29 F1 30 GO 31 H2 32 G2 33 GO 34' F0 15 H2 36 G2 36900 15100 0 21300 12200 31900 0 23100 44300 32700 24400 42500 34000 50000 25200 44600 37 G1 38 H2 39 G1 40 H2 41 FO 42 H2 43 G2 44 H2 18000 0 17300 0 31900 -0 18600 0 ;
45 F2 31700 23100. 39500 25100 49900 25100 40700 25500 l 37300 l 43900 46 H1 47 G2 48 H2 49 FO 50 H2 51 G2 52 1 G1 53 .G2 0 22800 0 34400 0 18500 17500 .23100
- 54 HO 21400 43600 25700 52200 25200 40700 38100 42200 0
14500 55 G2 56 H2 57 G2 SS H2 59 G2 60 H2 61 G2 62 C1 !
22200 0 21800 0 '23100 0 23100 29000 41400 25300 43600 25200 44600 25500- 42200 42800 l
X Y X = Quarter Core Assembly Number A Y = Fuel Batch B
A = Assembly Average Burnup (HWD/T),'BOC B = Asseebly Average Burnup (MWD /T), EOC
. . _ . 3-5
- g i PLCEA - PART LENGTH CEA BANK 6 - LEAD REGULATING BANK 5 SECOND REGULATING BANK 1 2 i 4 - THIRD REGULATING BANK i 3 FOURTH REGULATING BANK
- S 2 - FIFTH REGULATING BANK A 1 LAST-REGULATING BANK '
SB - SHUTDOWN BANK B 3 4 5- 6 7 ,
SA = SHUTDOWN BANK A 2 I 8 9 10 11 12 13 i S 3 B 4 !
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SRUTDOWN R00 IN POSITION 52-- 53 52 13 AVAILABLE FOR ONLY TWO DIAGONA'.LY'0PPOSITE CORE S*
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LOUISIANA' Figure ~
F0WER & LIGHT.00.,
WATERFORD 3 Waterford Steam- CEA BANK IDENTIFICATION- 3-4 -
. Electric Station.
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41 42 43 64 ia 17 isi 144 tee - 170 in 172 m 174 m- m. 177 its in sa0 141 182 153 144 15 14 146 187 188 189 190 191 '
45 44 4FE 192 193 44 49 50 - 51 194 195 196- 197 194 19 199 200 201 202 255 - 204 a 205 206 20F tot 209 210 211 212' 213 52 s3 u - $5 w --
l 21 214 215 214 - 21T LOUISIANA POWER & LIGHT CO. figure ,
WATERFORD 3 Waterford Steam IN-CORE INSTRUMENT ASSEMBLIES 3-5
~ Electric' Station CORE LOCATIONS 3-7
4.0 FUEL-SYSTEM' DESIGN 4.1 Mechanical Desion The mechanical design of the Waterford 3 Batch H reload fuel is essentially the sane as the debris-resistant Batch G fuel that was introduced in Cycle 5.
The changes associated with the debris-resistant design are descr'. bed in Appendix B, since Cycle 2. whien also describes the minor changes to the mechanical design The primary differences between the Batch H and Datch G fuel designs are minor I changes in the skirt that is located at the bottom of the upper end fitting's outer posts, and in the flange located at the upper end of the outer guide tube assembly. The changes improve the fabrication process. The outer post is ,
screwed into the guide tube flange to secure the-fuel assembly upper and fitting to the-fuel assembly grid cage, and-the post skirt is expanded into-holes in the flange to prevent the post-from unscrewing. These. changes will minimize distortion of the guide posts during fabrication while still meeting the anti-rotation torque requirements. This design change is not expected to impact the in-reactor performance of the fuel.
4.2 Mitigation of G3 tide Tube Wear e
All fuel assemolies in Cycle.6 will have stainless steel sleeves installed in the guide tubes to prevent guide tubs wear.
4.3 Thermal Design The thermal performance of composite fuel rods'that envelope the fuel rods of the Batenes present in Cycle 6 have been evaluated using the FATES 3B version of the CE fuel evaluation model (References 4-3. 4-4, ~4-5, and 4-6) . The ;
analysis was performed using a power history that enveloped the power and -
burnup levels representative of the peak fuel rod:at-each burnup interval, '
from Beginning of Cycle (BOC) to End of Cyclef(EOC) _ burnups. Therburnup range it analyred burnup dependent is in excesafof-that expected at the end'of. Cycle =6. Results of-.these
- Analyses (Section 7) fuel performance calculations were used in the Transient .
(Section 8) performedand for in the Emergency Core Cooling System-(ECCS) Analysis Cycle 6. -
4.4 Chemical Design ,
?-
- The material specifications that are used for the fuel cladding and-the other fuel assembly components in Batch H fuel are~the same.as those-that were used~ -
in Batch G fuel and essentially the same as those that-were used for. Batch:D (Cycle 2) fuel.
4.5 Shoulder Gap ' Adequacy.
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All of the fuel assemblies that.are to be used.in Cycle 6iwere evaluated to confirm that there is adequate shoulder-gap clearance through the end of Cycle'6. The method used for Cycle 6 is.that described in Reference 4-2. The initial shoulder gap 'in Batch F, Batch G, and Batch;H fuel;-is 2.382 inches, .
L ;i the came as reported in.the Cycle 2 Reload Analysis Report _for Batch D. The 3,
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5.0 NUCLEAR DESIGN 5.1 Physics Characteristics 5.1.1 Fuel Manacement The Cycle 6 core makes use of a very low-leakage fuel management scheme in which twice burned Batch F assemblies are placed en the core periphery Most of the fresh Batch H assemblies e.re located throughout the interior of the core where they are mixed with other previously burned fuel in a pattern that minimizes power peaking. With this loading and a Cycle 5 endpoint at 441 EFPD, the Cycle 6 reactivity for full power operat .on is expected to be 162 EFPD i [
(17,600 MWD /T) . Explicit evaluations have been performed to assure applicability of all analyses to a Cycle 5 termination burnup of between 427 '
and 457 EFPD and for a Cycle 6 length up to 490 EFPD (18,600 MWD /T)
Characteri7 tic physics parameters for cycle 6 are compared to those of the s Reference represent Cycle in Table 5-1. The values in this-table are intended to i nominal core parameters. Those values used in the safety analyses
.ontain appropriate uncertainties, or incorporate values to bound future i Operating cycles, and in all cases are conservative with respect (3 6 reported in Table 5-1. }
to -he va}IEVhs ,
Table 5-2 presents a summary of Control Element Assembly (CEA) reactivity worths (MSLB) and allowances for the end of Cycle 6 full power Main Steam Line Break '
ansient with a comparison to the Reference Cycle data. The Cycle 6 .
values ore explicitly calculated with 3D ROCS (Reference 5-2) while the Reference Cycle values were based en 2D POCS with adjustments for 30 effects. '
The full power steam line break was chosen to illustrate differences in CEA reactivity worths for thc two cycles. '
The CEA core locations and group identifications remain the same as in the ,
Reference Cycle. The power dependent insertion limits (PDIL's) for regulating '
groups and part-length CEA groups are shown in Figures 5-1 and 5-2 respectively and remain the same as in the Reference Cycle. Table 5-3 shows the reactivity worths of various CEA groups calculated at power conditions for- ;
Cycle 6 and the Reference Cycle.
-i
-5.1.2 Power Distributions-Figures 5-3 through 5-5 illustrate the calculated All Rods Out (ARO) planar ,
radial power distributions during Cycle 6. The one-pin -planar radial power [
peaks presented in these figures are obtained rom the middle-region of'the core. Time points at the beginning, middle, and end of cycle were chosen since l the variation in maximum-planar radial peak as a function of burnup is small.
I Radial power distributions for selected rodded configurations are given for
'- BOC and EOC in Figures 5-6 through 5-11. The rodded configurations shown are:
part-length CEA's (PLCEA's): Bank 6; and Bank-6-plus thelPLCEA's. '
The radial power distributions described in this section are calculated data which do not include any uncertainties or allowances.'The fine mesh calculations performed to determine these radial power peaks explicitly:
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. account for augmented power peaking wnich is characteristics of fuel rods adjacent to the water. holes.
L Nominal axial peaking factors are expected to range from.1.16 at BOC to 1.07 '
at Middle of Cycle, to 1.10 at EOC.
5.1.3 Maximum Fuel Rod Burnup '
The maxim 2m fuel rod burnup of 58,700 MWD /T projected _for the Waterford 3 Cycle 6 safety analysis is less than the 60,000 MWD /T limit presented.in 1:
Reference 5-3. Reference 5-3 has been transmitted to the NRC.by ABB/CE for generic approval. The physics data which are input to the Cycle 6 safety and r
fuel performance analyses are developed from explicit fine mesh calculations of fuel rod power and exposura.
i Those physics data which are burnup dependent, for example, maximum fuel red fluence and fuel rod power histories for FATES 3B analyses, conservatively envelope core and fuel rod behavior at the maximum burnups as well as lower burnups. Also, the power levels of the high burnup rods are more than 30%
below the EOC peak rod power levels.
5.2 Physics Analysis Methods 5.2.1 Analytical Input to In-core Measurements In-core detector measurement constants toibe used?in evaluating the reload
~
cycle power distributions will be calculated in accordance.with Reference 5-1.
5.2.2 Uncertainties in Measured Power Distributions.
The planar rad'al power-distribution measurement uncertainty of 6.92%, based .
upon Reference 5-1, will be applied to.the_ Cycle 6 COLSS and CPC cn-line 7 calculations which use planar radial power peaks. The'axia11 and three- I dimensional power distribution measurement uncertainties.are-determined using p
the values in. Reference 5-1 in-conjunction with other monitoring and protection system measurement' uncertainties.
5.2.3 Nuclear Desian Methodology L
As in the Reference. Cycle, the Cycle 6 nuclear design was performed with two-and three-dimensional core models using the ROCS computer code.and employings ,
! DIT calculated crosa sections. The ROCS-DIT code-and.the MC module were; l
described in Reference-5-2.
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Recent developments in ABB/CE physics methodology (Reference 5-5)' explicitly-t account _
f or the .f ollowing additional physical phenomena : : (1)' anisotropic scattering within the pin cells, (2)L anisotropic neutron current at cell +
interfaces, (3)' assembly discontinuity factors, and (4) the utilization'of the
- Nodal _ Expansion Method'(NEM) -in. ROCS instead of the' previous Higher Order Difference (HOD) solution. The ' Cycle '6 - design was _ performed' using ' the *
- Reference 5-5 methodology. }
The Reference 5-3 methodology produces'more accurate core. power distribution !
predictions by improving: 1) the global radial power distribution, where power .
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~4 4
sharing between neighboring assemblies is better modelled,-and 2) the local fuel pin poweridistribution within an assembly, where the predicted-distribution is-ilatter.
The ROCS-DIT methods and theories of Reference 5-2, as supplemented by Reference 5-5, have been reviewed by the NRC. In its acceptance of Reference 5-2, the NRC approved the. general physics theories employed by ROCS-DIT and the analytical technique used to obtain the, calculational uncertainties (95/95 limits) associated with the methodology, rather than the actual values of the uncertainties. ABB/CE has reviewed the available data '
base and has dotermined that the' calculational uncertainties are not -
substantially different as a result of the new methods. The ROCS-DIT topical and its SER anticipated that improvements to the methodology would be made.
The NRC, in its approval only required CE to reevaluate the uncertainties associated with such analytical changes. Consequently, . the ROCS-DIT Reference 5-5 methodology c 1 be applied without additional NRC review. Additional details are provided in Appendix A.t' this report.
Negative reactivity insertion as a function of CEA scram bank position was ,
calculated using the one dimensional-space-time code FIESTA (Ref erence 4) .
1 5
L q
r
}
e h
l l 54
- i s
' dr
- a. -- - - en ---n--- ., ,.,,v-w~-.--w.,, y -,-r-- - .-- e .w . m,...-,e--,,,-, o w- ,,. .,,,,.:,.,me e e.v..w-w e , p ,a, , --v n w- p pg
. . . ~ . . . - ... . . ~ -. _ . - . .- . . .- _ .-. . . . . - . - . . - . . - . . . -. .
r
-.-- [
,- - !=
TABLE 5-1 1 WATEPSORD 3 CYCL.E 6 +
NOMINAL PHYSICS CHAFACTERISTICS - i t
Reference Dissolved Boron Units Cycle Cycle 6 '
Dissolved Boron Concentration i for Criticality, CEA's
Hot Full Power, BOC ;
PPM /%Ap 107 130 Hot Full Power, EOC ',
PPM /%Ap '85 96 l Moderator Temperature Coefficients i
Hot Full Power, Equilibrium Xenon 1
' Beginning.of Cycle [
10
- 4 Ap/ ' F -0.1 -0.7 End of Cycle i 10-4 Ap/ ' F - -2.5- -3.1' !
Doppler Coef ficient Hot I:ero Power, LOC t'
'10 4Ap/*F- -1.7 -1,7 I Hot-Full Power, BOC 104 Ap/
- F -l.2 -1,2 Hot Full Power, EOC 10* sap /
- F -1.4 -1.4 .
5 Total Delayed Neutron Fraction, Deff r t
-0.0051 0'.0051 !
Neutron Generation Time, -:
l' I
(
1,
- BOC ,
10-* sec EOC 23.9 19.9 .!
i
.109 sec ' .3 0 . 0 26.6
- l. n I
t j
l -
I L- .;
t l =
l 5
+
b i I r
i
.5-4 2
4- - - -mv-. 5.,.e,v_w_m--.,
%.c .,,y- .m,...,, . . , , - .p., ,3,w.,-,.,,-,.y_ . , . . . _ , , , ,, y,,p77, p,y--g--.ee it 9 , ( w tr+e .. -n r- S -m 4 y
- . - .. - -. -- - - . . .~.. . . _ _ . . - - - -
TABLE 5-2 i
WATERFORD 3 CYCLE 6 LIMITING VALUES OF REACTIVITY WORTHS AND ALLOWANCES FOR HOT FULL POWER STEAM LINE BREAK,-lap, END-OF-CYCLE (EOC)
Reference C y*:l e Cycle 6 s
- 1. Worth of all CEAs Inserted -12.1 -
10.40
- 2. Stuck CLA Allow 6nce + 2.3 + 1.06
- 4. Full Power Dependent !asertion .
Limit CEA Bite + 0.3 + 0.24
- 5. Calculated Scram Worth 74
- 9.5 -
9.g '
13; 6, Physics Uncertainty + 1.0 + 0.57
- Other Allowances + 0.2 + 0.1
- 8. Het Available Scram Worth
- 8.3 - 8.43 >
l I
h I
-_ s t
2 b
5 .5 i
--y--+r- +- , q % y y .-- w-- y ery, y .w.m->p
. l
. l TABLE 5-3 WATERTORD 3 CYCLE 6 REACTIVITY WORTH OT CEA REGULATING GROUPS AT HOT FULL POWER, lap i l
i l
PeQihning of Cycle I End of Cyc_le_ 1 I
Regulating Reference Reference I CEA's i
Cycle Cycle 6 Cycle _ Cvele 6 !
l l
- roup 6 0.4 0.3 0.5 0.5 u Grcup 5 0.5 0,$ 0,6 0.4 !
Group 4 1.0 0.9 1.1 .1.1 1
'l l 1 l
Mote:
Values shown assume sequential ;roup insertion.
i
.1 l
l i
]
l l
l l
1
-l 1
5+6 i
-- - U
. _ . . . - . - - ~ - - . - . - - - - ----~... - - -. -.. - ~ _.._ .-~.-.. - -.-. ~..-. - -
( i i
- 4 4 l TABLE t-3 WATERFORD 3 CYCLE 6 i
PEACTIVITY WORTH CP CEA REGULATI!!3 GROUPS -
AT 1107 FULL PCWER, i .ip l e
i Beginning of Cycle End of Cycle '
t Regulating Referenco Reference CEAS Cycle Cycle 6 Cycle Cycle 6 i Group C 0.4 0.3 0.5 0$
Group 5 0.5 0.5 0.6 -0.4 l
1.0 Group 4 ') . 9 1.1 1.1
?
! Jot e :
Values otown assure sequential group insortion.
t Z
r i
i i
4 t-7 l
... .---.:....-.. . . . . . - . . - . . - - - - . . . . . . . . . . - . ~ , . . . . . , . - , . . .,z.-...,.-...,.. - , . - , , . . - . - ~ , . - - . . - - . ~ - - . , ~ , .
t i
i 8-9 FRACTION OF RATED THERMAL POWER o o o 'o o o o w b s in
'o o -
-a o o o o o o m 4 he to o'
, E- o o o o o u . . , . ~ , .
5 ;
4
/ LONG TERM STEADY STATE INSERTION LIMIT 1
m a E E g .
-[ A f
x o f- x
- e -
'a - / 'o r
i "
m g i !
[ SHORT TERM STEADY STATE /
i E_
. INSERTION LIMIT- I
/
u a l /
m m
o .
- o -- /
D /
- E _
5 g 8 -o- /
/ [
I 7 j R , -
/ 8
$ / 3
? m ,.
~
. / /'$.
o E' s '
" W 5 -
m
- -2 1 5A -i o ~
b m
- - 3
-H-
~
O l o-g -
- 2 C
o n=
. 9' wh 4
=
.=
b-14 t
4 O-l SdnOHD DNI1V70DBU UOd 110d CGWOdH31VM L 9 3HnDid .
t
,,...m-,..,,.---,,,-..-,-,,,...-.,_---,m--m,
I i
i 1.00 ,
i m 0.90 '
- - i e m m 0.80 -- - 2 3:
. a.
o-
+ ~ 112.5** (75%) $
z l,
_a 0.70 -
O '
4 2 .
E __
- z m I 0.60 mE mym
-o W 0.50 bmh 0 3m -
g F
l \ TRANSIENT Z om !
r- ei
- 0.40
' ' I E O j i '*
i 2
' O I s 0.30 LONG TERM d b I INSERTION LIMIT m i
< f .g- ,
m i
k 0.20 ' ' #
- ----------- - - ..- [
22.5" (15%) r-l o
0.10 m
m O i 150 140 130 120 110 100 90 80 70 60 y 40 30 10 0
,20 ,
PART LENGTH CEA POSITION, INCHES WITHDRAWN i
-m- -
,-w .. y ~-r r , , - - , ,,e ,nn, w ,- ,. ,,. e-,,, ,,,,,nr. ,, -- .,, - -. ,- -- .,, - ,, - , . - .,.--,,,,--r- -- n-,, , . . . - - .
-. .. _ _ _ - ._=.. --- - - . . . . . . _ - - .- ._
l l
Figure 5-3 WATERFORD 3 CYCLE 6 Assembly Relative Power Density HFP at BOC, Unredded 1 F2 2 H0 i
0.348 0.841 3 F1 4 F1 5 G1 6 H1'7 G2 0.237 0.371 0.720 1.175 1.095 8 F1 9 H1 10 G0 11 H2 12 G2 13 H2 0.324 0.852 0.953 1.180 1.127 1.303 14 F2 15 HI 16 G2 17 H2 18 G1 19 H2 20 G2 O.311 0.913 1.025 1.301 1.211 1.290 1.139 :
r 21 F1 22 HI 23 G2 24 GO 25 G2 26 H2 27 F0 28 H2 l 0.235 0.844 1.026 1.283 1.251 1.311 0.917 1.238
29 F1 30 GO 31 H2 32 G2 33 GO 34 F0 35 H2 36 G2 I l 0.370 0.951 1.302 1.250 1.301 0.993 1.278 1.138 l 37 G1 38 H2 39 Gl 40 H2 41 F0 a2 H2 43 G2 44 H2 -
45 F2 0.719 1.180 1.210 1.307 0.989 1.321 1.224 1.340 0.348 46 HI 47 G2 48 H2 49 F0 50 H2 51 G2 52 G1 53 G2 54 H0 1.175 1.127 1.289 ' O.916 1.276 1.224 1.191 1.104 ;
0.841 55 G2 56 H2 57 G2 58 H2 59 G2 60 H2 61 G2 62 Cl i
1.095 1.303 1.139 1.238 1.138 . 1.340 1.104 0.769 Maximum 1-Pin Peak = 1.503 in-Assembly 24 X Y X - Quarter Core Assembly Number !
Y = Fuel Batch ;
Z Z = Integrated Power Density !
5-10
_. __ _- , _ _ ,_ .- ~ _ .- -.
1 Figure 5 4 i WATERFORD 3 CYCLE 6 Assembly Relative Power Density HFP at MOC, Unrodded 1 F2 2 H0 0.351 0.771 3 F1 4 F1 5 G1 6 HI 7 G2 0.257 0.396 0.732 1.148 1 C34 8 F1 9 H1 10 G0 11 H2 12 G2 13 H2 ;
0.335 0.857 0.941 1.237 1.127 1.369 14 F2 15 H1 16 G2 17 H2 18 G1 19 H2 20 G2 1 0.323 0.89G 0.957 1.306 1.197 1.388 1.184 21 F1 22 HI 23 G2 24 G0 25 G2 26 H2 27 F0 28 H2 0.255 0.851 0.957 1.132 1.143 1.364 0.569 .1.364 T' F1 30 GO 31 H2 32 G2 33 G0 34 F0 35 H2 36 G2 j 0.395 0.939 1.306 1.142 1.174 0.978 1.366 1.170 37 G1 38 H2 39 G1 40 H2 41 FO 42 H2 43- G2 44 H2 I45 F2 0.732 1.237 1.197 1.351 0.976 1.361 1.207 1.386 i '
O.351 46 HI 47 G? 48 H2 49 FO 50 H2 51 G2 52 G1 53 G2 54 H0 1.148 1.126 1.388 0.968 1.365 1.207 1.119 1.035 O.771 55 G2 55 li?. 57 G2 58 H2 59 G2 60 H2 61 G2 62 Cl v
1.034 1.369 1.184 1.364 1.170 1.366 1.035 0.739 l
l l
l Maximum 1 Pin Peak 1.485 in Assembly 19 l
X Y X = Quarter Core Assembly Number Y = Fuel Batch Z Z = Integrated Power Density 5 11
e figure 5 5 WATERFORD 3 CYCLE 6 Assembly Relative Power Density HFP at EOC, Unrodded 1 F2 0 H0 0.389 0.771 3 F1 4 F1 5 G1 6 H1 7 G2 0.321 0.465 0.782 1.138 1.001 8 F1 9 H1 10 GO 11 H2 12 G2 13 H2 0.405 0.9".4 0.987 1.300 1.104 1.364 14 F2 15 H1 16 G2 17 H2 18 61 19 H2 20 G2 0.394 0.973 0.974 1.333 1.168 1.391 1.152 21 F1 22 HI 23 G2 24 GO 25 G2 26 H2 27 F0 28 H2 0.320 0.949 0.973 1.069 1.075 1.344 0.975 1.370 29 F1 30 GO 31 H2 32 G2 33 GO 34 F0 35 H2 36 G2 0.464 0.986 1.334 1.075 1.078 0.949 1.344 1.115 37 G1 33 H2 39 Gl'40 H2 41 F0 42 H2 43 G2 44 H2 45 F2 0.782 1.300 1.159 1.344 0.948 1.310 1.108 1.311 0.389 46 HI 47 G2 48 H2 49 F0 50 H2 51 G2 52 G1 53kG2 54 H0 1.138 1.104 1.391 0.975 1.344 1.108 1.004 0 T39 l 0.771 55 G2 56 H2 57 G2 58 H2 59 G2 60 H2 61 G2 62 Cl j1.001 1.364 1.152 1.370 1.115 1.311 0.939 0.705 Haximum 1 Pin Peak = 1.488 in Assembly 48 X Y X = Quarter Core Assembly Number Y = Fuel Batch Z l'= Integrated Power Density 5-12
l 1
Figure 5 6 WATERFORD 3 CYCLE 6 Assembly Relative Power Density !
HFP at BOC with PLCEA's .
l l
l 1 F2 2- H0 0.355 0.837 3 F1 4 F1 5 G1 6 HI 7 G2 0.244 0.382 0.733 1.176 1.112 l 8 F1 9 H1 10 G0 11 H2 12 "G2 13 H2 O 333 0.858 0.966 1.177 1.142 1.302s ~
s-1.-
t 14 F2 15- H1 16- G2 17 H2 18 G1 19- H2 20 G2 0.319 0.921 1.048 1.290 1.198 1.272 1.148 e
21 F1 22 HI 23- G2 24= GO 25 G2 26' H2 27 F0 28 H2 0.242 0.850 1.048 1.298 1.246 1.191 0.909 1.227 29 F1 30 GO 31 H2 32 G2 33 GO 34 F0 35 H2 36 G2 O.381 0.963 1.290 1.245 1.295 0.989 1.266 1.154 i f
37 G1 38 H2 39 G1 40 H2 41 F0 42 H2 43 G2 44 H2
! 45 F2 0.733 1.176 1.197 1.187 0.985 1.318= 1.246 1.352 I 0.355 46 HI 47 G2 48 H2 49 i0 50 H2 51- G2 52 G1 53 G2, 54 H0 1.176 1.142 1.27) 0.908 1.265 1.245 1.229 '1.148 0.838 55 G2 56 H2 571 G2 58 H2 59- G2 60 H2 61 G2 62 Cl t
1.112 l'.302 , 1.148 1.227 1.154 1.352 1.148 0.789 :
Maximum-1-Pin Peak = 1.513 in Assembly 24-
[ .
t r
rt X ; Y. X = Quarter Core Assembly Number )!
Y = Fuel Batch Z- Z~= -
Integrated Power Density:
t Ia W
5 13
. . . _ _ . - . _ . - . . . _ . , _ , _ .._._..._.._...____.___._.._-__;._-._..._._..-...-.,_2-
Figure 5-7 WATERFORD 3 CYCLE 6 Assembly Relative Power Density HFP at BOC with Bank 6 1 F2 2 H0 0.364 0.856 3 F1 4 F1 5 G1 6 HI 7 G2 0.260 0.401 0.754 1.194 1.124 8 F1 9 H1 10 GO 11 H2 12 G2 13 H2 0.357 0.914 1.015 1.208 1.136 1.274 14 F2 15 H1 16 G2 17 H2 18 G1 19 H2 20 G2 0.342 0.987 27 1.362 1.233- 1.209 1.014 21 F1 22 HI 23 G2 24 GO 25 G2 26 H2 27 F0 28 HT.
0.257 0.905 1.117 1.381 1.323 1.304 0.816 0.724 29 F1 30 GO 31 H2 32 G2 33 GO 34 F0 35 H2 36 G2 0.399 1.012 1.363 1.322 1.356 1.003 1.183 1.005 37 G1 38 H2 39 G1 40 H2 41 F0 42 H2 43 G2 44 H2 45 F2 0.753 1.206 1.232 1.300 0.999 1.306 1.205 1.291 0.364 46 HI 47 G2 48 H2 49 F0 50 H2 51 G2 52 53 G1 G2 54 H0 1.194 1.135 1.208 0.815 1.182 1.205 1.199 1.119 0.856 55 G2 56 H2 57 G2 58 H2 59 G2 60 H2 61 62 G2 Cl 1.124 1.274 1.014 0.724 1.005 1.291 1.119 0.772 Maximum 1-Pin Peak o 1.623 in Assembly 24 X Y X = Quarter Core Assembly Number Y = Fuel Batch Z Z = Integrated Pcwer Density 5 14
Figure 5 8 WATERFORD 3 CYCLE $
Assembly Relative Power Density HFP at BOC, Bank 6 and PLCEA's 1 F2 2 H0 0.376 0.885 3 F1 4 F1 5 G1 6 HI 7 G2 0.264 0.408 0.770 1.226 1.156 0 F1 9 H1 10 GO 11 H2 12 G2 13 H2 0.362 0.924 1.023 1.220 1.155 1.299 14 F2 15 HI 16 G2 17 H2 18 G1 19 H2 20 G2 0.346 0.996 1.122 1.355 1.216 1.208 1.022 21 F1 22 HI 23 G2 24 GO 25 G2 26 H2 27 F0 28 H2 0.262 0.915 1.121 1.374 1.296 1.187 0.800 0.722 29- F1 30 GO 31 H2 32 G2 33 GO 34 F0 35 H2 36 G2 0.406 1.021 1.356 1.295 1.326 0.981 1.175 1.007 37 G1 38 H2 39. G1 40 H2 41 FO 42 H2 43 G2 44- H2 45 F2 0.770 1.219 1.215 1.183 0.377 1.303 1.214 1.305
'~
0.376 46 HI 47 G2 48 H2 49 F0 50 H2 51 G2 52 G1 53 G2 54 H0 1.226 1.154 1.207 0.799 1.174 1.214 1.217 1.139 0.885 55 G2 56 H2 57 G2 58 H2 59 G2 60 H2 61 G2 62 C1 1.156 1.299 1.022 0.722 1.007 1.305 1.139 0.7B8 Maximum 1-Pin Peak =-1,602 in Assembly 24 X Y' X = Quarter Core Assembly Number Y = Fuel Batch Z Z = Integrated Power Density -
5 -
figure 5-9 WATERFORD 3 CYCLE 6 Assembly Relative Power Density HFP at EOC with PLCEA's 1 F2 2 H0 0.402 0.797 ;
3 F1 4 F1 5 G1 6 HI 7 G2 0.328 0.474 0.800 1.168 1.030 8 F1 9 H1 10 GO 11 H2 12 G2 13 H2 0.412 0.968 0.998 1.315 1.122 1.390 14 F2 15 H1 16 17 G2 H2 18 G1 19 ~H2 20 G2 0.401 0.986 0.981 1.326 1.152 1.389 1.160 21 F1 22 H1 23 G2 24 GO 25 -G2 26 H2 27 F0 28 H2 0.326 0.963 0.980 1.065 1.048 1.219 0.958 1.368 29 F1 30 GO 31 H2 32 G2 33 GO 34 F0 35 H2 36 G2 0.474 0.997 1.327 1.048 1.050 0.926 1.334 1.117 37 G1 33 H2 39 G1 40 H2 41 F0 42 H2 43 G2 44 H2 45 F2 0.800 1.315 1.152 1.219 0.925 1.305 1.115 1.325 0.402 16 HI 47 G2 48 H2 49 FO 50 H2 51 G2 52 G1 53 G2 54 H0 1.169 1.122 1.389 0.958 1.334 1.115 1.019 0.956 0.798 55 G2 56 H2 57 G2 58 H2 59 G2 60 H2 61 G2 62 Cl 1.030 1.390 1.160 1.368 1.117 1.325 0.956 0.721 Haximum 1-Pin Peak - 1.506 in Assembly 56 X Y X - Quarter Core Assembly Number Y = fuel Batch Z Z = Integrated Power Density 5-16
Figure 5-10 WATERFORD 3 CYCLE 6 Assembly Relative Power Density HFP at E0C with Bank 6 1 F2 2 H0 0.417 0.825 3 F1 4 F1 5 G1 6 HI 7 G2 0.353 0.503 0.831 1.197 1.049 8 F1 9 HI 10 GO 11 H2 12 G2 13 H2 0.448 1.042 1.057 1.358 1.123 1.36 [t, 14 F2 15 H1 16 G2 17 H2 18 G1 19 H2 20 G2.
0.435 1.068 1.057 1.412 1.188 1.312 1.009 21 F1 22 HI 23 G2 24 G0 25 G2 26 H2 27 F0 28 H2 0.351 1.037 1.056 1.144 1.122 1.334 0.844 0.762 29 F1 30 GO 31 H2 32' G2 33 GO 34 F0 35 H2 36 G2' O.502 1.057 1.412 1.122 1.103 0.935 1.230 0.947 37 G1 38 H2 39 G1 40 H2 41 FO 42 H2 43 G2 44 H2 t
45 F2 0.831 1.358 1.188 1.333 0.934. 1.288 1.071 1.252 0.417 46 HI 47 G2 48 H2 49 FO 50 H2 51 G2 52 1 G1 53 G2 54 H0 1.198 1.120 1.312 0.843 1,229 1.071 0.990 0.928 l
0.825 55 G2 56 H2 57 G2 58 H2 59- G2 60 H2 51 l G2 62 Cl l 1.049 1.363 1.009 0.762 0.947 1.252 0.928 0.703 a-
' Maximum 1-Pin Peak - 1.561 in Assembly 31 X Y X - Quarter Core Assembly Number Y = Fuel Batch I Z = Integrated Power Density 5-17
Figure 5-11 WATERFORD 3 CYCLE 6 Assembly Relative Power Density HFP at E0C, Bank 6 and PLCEA's 1
1 F2 2 H0 0.431 0.853 3 F1 4 F1 5 G1 6 HI 7 G2 0.361 0.514 0.851 1.229 1.079 8 F1 9 H1 10 GO 11 H2 12 G2 13 H2 0.457 1.058 1.070 1.374 1.139 1.389 14 F2 15 HI 16 G2 17 H2 18 G1 19 H2 20 G2 0.444 1.08" 1.065 1.406 1.172 1.309 1.016 21 F1 22 HI 23 G2 24 GO 25 G2 26 H2 27 F0 28 H2 0.359 1.052 1.064 1.141 1.095 1.209 0.824 0.757
~29 F1 30 GO 31 H2 32 G2 33 GO 34 F0 35 H2 36 G2 0.513 1.069 1.400 1.095 1.073 0.910 1.216 0.947 37 G1 38 H2 39 G1 40 H2 41 F0 42 H2 43 G2 44 H2 45 F2 0.851 1.374 1.172 1.208 0.908- 1.280 1.076 1.263 0.431 46 HI 47 G2 48 H2 49 FO 50 H2 51 G2 52 G1 53 G2 54 H0 1.230 1.139 l.309 0.824 1.216 1.076 1.003 0.943
. 0.853 55 G2 56 H2 57 G2 58 H2 59 G2 60 H2 61 G2 62 Cl 1
i 1.079 1.389 1.016 0.757 0.947 1.253 0.943 0.717 L
Maximum 1-Pin Peak = 1.564 in Assembly 31 X Y X = Quarter Core Assembly Number Y = Fuel Batch Z Z = Integrated Power Density 5-18
- . - - - . . . ~ . ~ . . - - . - - - - ~ . - - - - _ . . . . . . . - . _ _ - - - . ~ _ . . . . . . . .-
f 4 i
. , i E
6.0 THERMAL-HYDRAULIC OESIGN !
6.1 DNDR Anclysis '
t Steady State Departure f rom Nucleate Boiling Ratio (tNDR) analyses of Cycle 6 at the core rated power level of 3390 MWt have been performed using the TORC '
computer code descrited in Reterence 6-1, correlation described in Reference 6-2, thethe CE-1 Critical Heat riux (CHF)
- described in R21erence 6-3, simplified TORC modeling methods and tne CETOP code described in Reference 6-4.
Table 6-1 presents a compatison of pertinent thermal-hydraulic design parameters Ur.ce rt aint ies (SCU) for Cycle 6 and the Reference Cycle. The $tatistical Combination of methodology presented. in Reference 6-5 was applied with daterford 3 specific data. This-was done using the calculational factors listed in Table 6-1 and other uncertainty factors at the 95/95 1
confidence DNBR .
/ probability level to define a design limit of 1.26 on CE-1 minimum -
The Cycle 6 DNDR limit includes the following. allowances:
1.
IIRC specifitd allowancis for TORC code uncertainty.and CC-1 CHF correlation crossLvalidation uncertainty, as discussed in Reference 6-10.
2.
An NRC imposed 0.01 DNPR penalty for HID-1 grids as discussed in Reference 6-6, 6-7, and 6-0.
- 3. Rod bow penalty as discussed in Section 6.. below.
6.2 Effects of Fuel Pod Bowino on DNBR Marcin &
Effects 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 tnis analysis, 1.75% on minimum DNBR, is valid for bundle burnups up to 30,000 MWD /T.
This penalty is included in the 1,26 DNBR: limit. l For assemblies with burnups greater than 30,000.MYlD/T,
- exists to offcet rod bow penalties due to the lower radial' sufficient margin-power peaks in these higher burrup batches. Hence, _
the rod bow' penalty based upon. Reference-
-Cycle 6-9 for 6. 30,000 mwd /T.is applicable for all assembly-burnups expected for I
s f
t F
E
[. '
.,,,,4.,.,,- ,_a.......-a,~----- t
,_m ew*-*~aw'"*e' ~ "" "-"~" "' ~ ~~ , . , . -- - " ~ ~ ' ' ~ ~ ~ '
. - . __ _ _ . .._ _ _ _ . . _ . _ . ~ . _ _ _ _ _ _ . _ _ _ _ _ . _ . _ _ _ . . - _ _ _ . _ _ _
e TABLE 6-1 '
Waterford 3 Cycle 6 Thern.a1 Hydraulic Ptirameters at Full _ Power I Reference General Characteristics " nits Cycle 12) Cycle 6 l
Total Heat Output (Core only) MWt 3390 3390 10' BTU /hr 11570 11570 L
Fraction of Heat Generated in --
0.975 0.975 E
Fuel Rod ,
h Primary System Pressure psia 2250 (Nominal) 2250 inlet Iemperature (Nominal) 'F 553.0 553.0 Total Reactor C olant Flow gpm 396,6v0 396,000 (Minimum Steady State) 106 lb/hr 148.0 Coolant Flow Througn Coro 148.0-106 lb/hr 144.2 144.2 (Minimum)
Hydraulic D.iameter ft. 0.039 0.039 (Nominal Channel)
Average Mass Velocity 106 lb/hr-ft2 2 64 2.64 Pressure Drep Across Core psi- ~ 15.4 15.4 (Mittimum steady state flow trreversible AP over entire '
fuel assembly)
Total Pressure Across psi 41.5 - 41.5
'.'e s s e l (Based on nominal dimensions and minimum steady state flow) P Core Average Heat Flux BTU /hr-ft2 I 182,700*** 185,400*
(Accounts f or f raction of ,-
heat generated in fuel rod
' and axial densification factor)
Total Heat Transfer Area 14.2 61,700'** 60,800*
(Accounts for' axial' densification factor) 2 6-2 v
-E.,r.., -c4 .a.,ei-wrrmm.,.w-*+w.3-+ws+wmwassw.-* =-esv-m"-4"t*++tme fi*-+-Ws ww-" k T v -'t y e-w me r t m - -TN v--l- *-mN-+-me+--s va- m e '
MvM - we + ve w 'vv< w e' r wn e e +W'v.+t + re
.. . . . . _ . _ . . . . - - . . - - - . = _ . . . - . . . . . - - . . _ - . _ . . . - . ~ . - ~ _ _ _ . .. _ -..-
_Tablo 6-1 (continueal Reference General Characteristics Units Cycle (2) Cycle 6 Film Coefficient at BTU /hr-ft 2 *F 6200 6200 Average Conditions Average Film Temperature *F 29.3 Difference 29.9 Average Linear Heat Rate hW/ft 5.34*** 5.42*
of Undensified Fuel Rod (Accounts for fraction of heat generated in fuel rodi Averaf,e Core Ehthalpy Rise BTU /lb 60.3 80.3 Maximum Clad Surface 'F 656.7 ks Temperature 656.l j2'
'M Engineering Heat Flux Factor ----
1.03** 1.03**
Engineering Factor on Hot ----
1.03**
i Channel Heat Input 1.03**
Rod Pitch, Bowing and Clad ----
1.0$**
Diameter Factor -1.05**
Fuel Densif.:atien Factor ----
1.002 1.00.
(Axial)
I 40TES:
Based on 2348 poison rods.
These factors at factors have been ccmbined' statistically with other uncertainty limit the 95/95 confidence / probability level to define a new design on CE-1 minimum D!JBR when iterating on power as discussed in Reference 6-5.
Based on 172B poison rods.
4 4
l l
l 6-3
__._..____.m.____ _____._._ _ - .. . _ . . _ _ _ _ _ . . - .
9 4 .
7.0 NON-LOCA SAFETY ANALYSIS 7.0.1 Introduction
'I i
This section analyses at apresents rated power the of results 3410 of MWt.the Waterford 3 Cycle 6 Non-LOCA safety ~
i The Design Basis Events (DDE's) considered in the safety analyses are listed in Table 7.0.1. These events are categorized into three groupst Moderate Frequency, Infrequent, and Limiting Fault Events. For the purpose of this report, the Moderate Frequency and Infrequent Events are referred to.as Anticipated Operational Occurrences (AOO's). The DBE's were evaluated with
' respect to four criteria: Offsite Dose, Reactor Coolant System Pressure,. Fuel Performance Limits (S AFDL's (DNBR )),
and Fuel Centerline Melt Specified Acceptable Fuel Design and Loss of Shutdown Margin. Tables 7.0-2land 7.0-5 precent the list of events analyzed for each criterion. All events were reevaluated to assure that they meet their respective criteria for Cycle 6. The DBE's chosen for analysis for each criterion are the limiting events with respect to that criterion. !
I For the event analyses presented, a discussion of the reason (s) for the reanalysis, a discussion of the cause(s) of the event, a description of the analyses performed, results, and conclusions are included.
7.0.2 Methods of Analysis The analytical methodology used for Waterford 3 Cycle 6 non-LOCA safety analyses is the same as previously presented unless.otherwise stated in the event presentation. Only the methodology that has been previously reviewed and approved on the Waterford 3 docket or on other dockets is used. The Reference Cycle for the individual DBE's is taken to be the last Cycle that the results ,
for the individual events were presented to the NRC (Cycle 1, Reference 7-1, or Cycle 2, Reference 7-2).
Changes in inputs to the non-LOCA safety analyses, whether f rom the~ Cycle 6E core loading or from other changes to the plant configuration are assessed to determine if a reanalysis of any DDE is. required.
7.0.3 Mathematical Models The f ollowing mathematica1' models and computer codes were used to analyze the DBE's for Waterford 3 Cycle 6.
Plant response for non-LOCA events was simulated'using the CESEC III computer code (Reference 7-3). Simulation of the fluid conditions within the hot y channel of the reactor core and calculation of DNBR was, performed using the '
CETOP-D computer- code described in Reference. 7- 6.
The HERMITE computer code (Reference 7-7) was used:to; simulate the. reactor
~
- point core for analyses which required more spatial-detail than is provided by a kinetics model.
74 .
5
- . . ..:._---.-- -..-u..- -.. -. -, _.-.,. ,- - .-.. u.-,- .a=--. .,,. -. .
. -1 The TCRC computer code (References 7-8 and 7-9) is used to simulate the fluid conditiens within the reacter core and to calculate fuel pin DNBR for the sheared shaft event.
Determination of DNBR for the post-trip return-to-power portion of the Main Steam Line event is based on the correlation developed by R. V. Macbeth (Reference 7-4) with corrections to account for non-uniform axial heat flux developed by Lee (Reference 7-5). This methodology is the eamo as that employed in the Reference Cycle analysis.
The number of fuel pins predicted to experience clad failure is taken as the number of pins which have a CE-1 DNBR value below 1.06, ex9ept for analyses in 5 which the method of statistical convolution (Reference 7-10) has been presented to the NRC on the Waterford 3 docket.
7.0.4 Input rarameters and Analysis Assumptions Table 7.0-6 summarizes the core parameters assumed in the Cycle 6 transient
-analysis and compares them to the values used in the previous Cycle (Reference 7-11). Specific initial conditions for_each event are tabulated-in the section of the report summarizing that event. For some of the pnge s, certain initial core parameters were assumed to be more limiting *.han the actual calculated Cycle 6 values (e.g., CEA worth at-trip, moderator temperature coefficient).
7.0.5 Conclusion All DBE's were evaluated for Waterford 3 Cycle 6 to determine whether their results are bounded by the Reference Cycle. It was determined that the consequences of all Cycle 6 non-LOCA transients are bounded by the results already on the Waterford 3 docket except for the Increased Main Steam riew with Loss of of fsite Power event. A discussion of this event for Cycle 6 la included in Section 7.1.3.
T-$
- . . - - - . - . - - . - . . . _ .. . - . . . - - - . ~ . . . ~ _ _ . .~ ~.-.- - - - ~ ~.- - _ _.
Table 7.0-1 I i
I waterford 3 Desian Basis Events Considered in the Cyc.le 6 Safety Analysis 7.1 Increase in Heat Removal by the Secondary System 7.1.1 Decrease in Feedwater Temperature ,
7.1.2 Increase in Tendwater Flow 7.1.3 Increased Main Steam Flow 7.1.4 Inadvertent Cpening of a Steam Generator Safety Valve or Atmospheric Dump Velve 7.1.5
- Steam System Piping Failures 7.2 Decrease in Heat Pemoval by the Secondary System "
7.2.1 Loss of External Load 7.2.2 Turbinc Trip 3 7.2.3 Loss of Condenser Vacuun 7.2.4 Loss of Normal AC Power V.2.5 Loss of Normal Feedw+ter 7.2.6 *' Feedwater System Pipe v waks 7.3 Decrease in Reactor Coolant Flowrate 7.3.1 7.3.2 Partial Loss of Forced Reactor Coolant Flow 7.3.3
- Total Loss cf Forced Reactor Coolant Flow Single Reactor Coolant Pump Shaft Seizure / Sheared
~
i Shaft 7,4
' Reactivity and Power Distribution Anomalies 7.4.1 Uncontrolled CEA Withdrawal from a Suberitical or Low Power Condition 7.4.2 7.4.3 Uncontrolled CEA Withdrawal at Power CEA Miscperation' Events 7.4.4 _
Chemical and Volume Control System (CVCS) Mal...netion '
7.4.5 (Inadvertent Doron Dilution)
' 4.6
- Startup of an Inactive Reactor-Coolant System Pump CEA Eiection 7.5 Increase in Reactor' Coolant fystem Inventory i
7.b.1 CVCS Malfunction ,
7.5.2 Inadvertent Operation. of the ECCS' During Power Operation 7.6 Decrease in Reactor Coolant System-Inventory 7.6.1 7.6.2
- Pressurizer Pressure Decrease Events Small Primary Line Break Outside Containment 7.6.3
- Steam' Generator Tube Rupture 7.7 MAscellaneous 7.7.1 Asymmetric Steam Generator Events-Categorized as Limiting Fault Events 7-3 i
_ _ . _ . . ~ . _ . - . - _ . - .;_ __.___ u .~ _ _- ,.a,._., _ . _ . , . , . .
. . . _ . - . _ . . . - . , _ _ - . - - . - _ - . . .- - - . - - - . . _ - . . - . - . . . . _ ~ _ . . . . ~ . . - . ,- ~
- eble 7.0-2 DBE's Evaluated with Paspect_t;2_offsite Dese Criterien ,
Section Event i Results Al Anticipated Cperational Occurrences 7.1.4 1) \nadver ent opening of a Steam ,
Bounded by Generator Safety Valve or Reference Cycle Atmospherac Dump Valve 7.2.4 2) Loss of Normal AC Poser Bounded by Reference Cycle B) Limiting Fault Etents
- 1) Steam System Piping railures:
7.1.5.a a) Pre-Trip Power Excursions Bounded by 7.1.5.b Reference Cycle b) Post Trip Analysis Bounded by y 7 . 2 . f- 2)
Reference Cycle , ', s feedwater System Pipe Breaks Bounded by t' 7.3.3 3)
Reference Cycle Single Reactor Coolant Pump Bounded by Shaft Soizure/ Sheared Shaft Reference Cycle 7.6.? 4) $ mall Priiury Line Break Outside Bounded by Containment. Reference Cycle 7.6.3 5) Steam Gbnerator Tube Rupture Sounded by Reference Cycis l
l l
7-4 i
l t
. i i
i
- able 7 0-3 . !
DBE's Evaluated with Fampect to LCS Pressure Criterion 'i Section !
Event
,f Results I Al Anticipated 0peracional Occurrences i 7.7.1 1) Lose of External Load 1
Bounded by i 1.0.0 Ref9tence Cycle
- 2) Carb ae Trip Bounded by i
7.2.3 .3)
RefSLence Cycle I Loss of Condenbar Vacuum Bounded by 7.2.4 4)
Reference Cycle .
Loss of Normal AC Power Bounded by Reference Cycle 7.2.5 5) Loss of Normal reedwater Bounded by 7.4.1 6)
Reference Cycle i Uncontrolled CEA Withdrawal Bounded by from SuDeritica) or Reference Cycle Low Power Conditions-1.4.2 7) Uncontrolled CEA Withdrawal Bounded by at Power ,
7.5.1 8) Reference Cycle CVCS Malfunction Bounded by i
I 7.5.2 Reference Cycle
- 9) Inadvertent Operation of the Bour.ned-by "
ECCS During Power Operation- Referonce Cycle t
B) Lim 1 ting Fault Events 7.2.6 1) Feedwater System Pipe Breaks Bounded by '
Reference Cycle p
l b
1-s. ,
4 i
vy . - -,gwn, ww.mv-,w,-g----y- ,-,--wy-.f ,4 ~ , i r m y ,
, ,-wy----y...,y. wwy,.,..,mn-+y, ,w r,m, --r .w w n ,9 w, . w sw, m y e.
p %,w ,, v .p . vge 3,%i Ls,%_r%, es--g.,-wmw_,,-.-.-,.w.,
Table 7.0-4 DBE's Evaluated with Pespect t o Fuel Fo rf o rmance Sectien Event Results A) Anticipated CPerational Occurr ences t 7.1.1 1) Decrease in reedwater Temperature Bounded by 7.1.2 2)
Referenes Cycle Increase in Feedwater Flow Bounded by 7.1.3 3)
Reference Cycle Increase in Main Steam riow Not D:unded by Reference Cycle 7.1.4 (presented)
- 4) Inadvertent Opening of a Steam Bounded by Generator Safety Valve or Reference Cycle Atmospheric Dump V41ve 7.3.1 5) Partial Loss of Forced Reactor Bounded by i Coolant Flow Reference Cycle 7.3.2 6) Total Loss of Forced Reactor Bounded by Coolant Flow Referenc.e Cycle 7.4.1 7) Uncentrolled CEA Withdrawal Dounde,by from Suberitical er Reference Cycle Low Power Conditions 7.4.2 B)
Uncontrolled CEA Withdrawal Bounded by at Power 7.4.3 Reference Cycle
- 9) CCA Misoperation Events Bounded by 7.6.1 10) Reference Cycle Pressuriter Pressure Decrease Bounded by Events 7.7.1 11) Reference Cycle Asymmetric Steam Generator Events Bounded by l
Reference Cycle el Lim 1 ting Fault Events 1;
Steam System Piping Failures:
l 7.1.5.a a) Pre-Trip Power Excorsions Bounded by 7,1.5.b Reference Cycle l
b) Post Trip Ar. clysis Bounded by 7,3.3 2) Reference Cycle Single Peactor Coolant Pump Bounded by 7.4.6 Shaft Seizure / Sheared Shaft Reference Cycle
- 3) CEA Ejection Bounded by Reference Cycle 1-6
- . _ . _ _ . ...__,_..m.._._ ,- _ . .m. _ ...__ _ ._ ____ _ ._.._ _. _ . _ _ . . . _ . . . _ . . . _ . _ _ _ _ _ . -
4 4
Table 7.0~5 I DDE's Evaluated with Fespect to Shutdown Marcin Criterion $
Section Event t Results ,
A) Anticipated Operat:lCnal Occurrences 7.1.4 1) Inadvertent opening of a Steam Bounded by Generator Safety Valve or Reference Cycle Atmoophoric Dump Valve 7.4.4 2) CYC5 Malfunction 3ounded by '
7.4.5 (Inadvertent Boron Dilution) Reference Cycle
- 3) Startup of an Inactive Reactor Bounded by Coolant System Pump -Reference Cycle -
B) I,imiting Fault Events t
- 1) Steam System Piping Tallures: .
7.1.5.b i b) Post Trip Analysis Bounded by '
. Reference Cycle- '
i i
r I
s
?
I p
- f
+
P l' ;i 1
s 7
i
- i
, - _ , . _ , . . _ _ _ _ _ - . . . . . - . _ . . _ . . . . . _ . . , . . , . _ , . _ . - . , _ _ . . _ _ . _ . . ~ . . . _ _ . , . . . . _ , , . . - , _ . _ . . . _ . . -
Table 7.0-6 Waterford 3 Cycle 6 Core Parameters Input to Safety Analyses ,
Previous Cycle Safety Parameters Values Units .(Cycle 5) Cycle 6 Values Total F.CS Power MWt 3478 3478 (Core Thermal Power
+ Pump Heat) .
Core Inlet steady state *r 542 to 560 7emperature -542 to $60 (70% power and (70% power and above) above) 520 to 560 520 to~560 (below 70% power) (below 70% power)
Steady State psia RCS Pressure 2000 - 2300 2000 - 2300 g g:-'
e Rated Peactor GPM 396,000 to coolant riow 396,000 to 410,000 410,000 Axial Shape Index LCO ASI - ,3 to +.3 .3 to +.3-Band Assumed for Units All Powers Maximum CEA Insertion % Insertion 28 at rull Power 29 of Lead Bank-
% Insertion 25 25 of Pa rt-Length Maximum Initial Linear KW/ft 13,4 Heat Rate 13.4
!' Steady State Linear KW/ft 21.0 Heat Rate for ruel 21.0 Center Line Melt Minimum DNBR-CC-1 (SAFDL) 1.26 Macbeth (ruel failure 1.26 1.30 1.30 limit for post-trip SLB with LOAC) 7-8 i
t y- t---' , - * +-.- e,.-e-v., ---e .-. .-
r ---r
_.__ ._ _ ___ _ .._.. _ .- ._ _____ _ . _ _ _ . . . . . . . _ _ . . _ . _ _ _ . _ . _ _ _ . . . _ . _ _ _ . ~ . _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _
i i
a s
Table 7.0-6 (continued)
Previous Cycle i Safety Paran.eters Values Units (Cycle 5) Cyclo _6 Values Moderator Temperature 10-4 Ap/
- r -3.3 to +0.5 Coefficient -3.3 to +0.5 (below 70%
(below 70% power) power) -3.3 to 0.t)
-3.3 to 0.0 (70% power and (70% power and above) above)
St.atdown Margin idp -5.15 (Value Assumed in -5.15 Limiting EOC t tero Power SLB)
-i Tine for the Average Seconds 3.0 3.0 t
CEA Position to Reach 90% Insertion Following Opening of the Trip 1.reakers t
P L
k E
f I
I l
l 7' g r
i
, ,%. --[-+ mm. Ee, w . .,v'y, .p,.,r, u ..ma, ,--.1,, , , ,.y,.m.,_wi,w4....e,_v. y...e,4r, ,*q er y , , ..n.,.y., , 7.- ,- ,w ,, h m , - - , e .w y - - . . , , . ,, , , y,
- . . _ . - _ - - - _ . _ _ _ - - .. - - - - -..- - _ _ ~.- -.~ ~ - .. - .- ,.
' 4 7.1 INCREASE IN HEAT REMOVAL BY THE SECCNDARY SYSTEM 7.1.1 Decrease in tenwater Temperature The results are bounded by the Reference Cycle 7.1.2 Increase in Feedwater Flow The results are bounded by the Reference Cycle 7.1.3 Increased Main Steam Flow r
The Increased Main Steam Flow event with a loss of AC Power and the associated coastdown of_the Reactor Coolant Pumps is an event for which the number of fuel pins predicted to experience DFB will be calculated. The flatter ,
distribution of fuel rod powers (" pin census") for Cycle 6 was evaluated to quantify the percent of fuel pins experiencing DNB.
Although the percent of fuel pins predicted to experience CNB remains small, '
the' Cycle 6 value for'the percent of fuel pina predicted to experience DNB is not event.
bounded by the value reported in the Reference' Cycle (Cycle 11 for this Using the method of statistical convolution, the FSAR reported 0.83% of the fuel pins will experience DNB.
7.1.3.1 Identification of Causes The Increased Main Steam Flow event.13 conservatively assumed to use up all the thermal margin preserved by COLSS and the other LCO's through the combination of increases in core power and decreases in RCS pressure. The riant is postulated to reach a temporary steady state condition with the peak pin in the core just above the conditions which would result in a CPC Low DNBR Trip. ,
The perturbation in the secondary system-due to the decreased. secondary system-pressure is then postulated to result.An a turbine trip. A' Loss of AC power .,
(LOAC) is postulated to immediately accompany the turbine trip. The LOAC is l postulated to interrupt power to the RCP's causing a 4 pump coastdown.
As soon as the decrease in flow is detected by the CPC's, the CPC. calculation 's of DlH3R will return a value lower than the trip.setpoint. The Cycle 6_ analysis credits the timing of the.CPC calculations of DNBR which yields a reactor trip-earlier than waiting for the RCP's-to_ reach the Low Pump Speed setpoint. The.
Low Pump Speed setpoint 1 was used for reactor trip in'the Reference Cycle analysis.
-7.1.3.2 Analvois of Effects and consequences
~
The Increased Heat Removal part of the transient has not been explicitly.
modelled- for Cycle 6 since it is conservatively- assumed that this portion of-the event use.= up all_the thermal margin initially preserved by the LCO's.
This assumption is' conservative'as explicit modelling of tha Excess Load portion of the event would-demonstrate that therral margin to the SAFDL exists - ,
at the-time of LOAC. The analysis performed for. Cycle 6'is equivalent to-the 7-10
_ . . , _ _ _ . , _ ,..._-._.u-,,,_,,._.,,- _ . , , , ~ , _ _ . - - , , , _ _ - , , . , _ _ - , - .- .--,,,,, 4 ,_
- l analysis of a 4 pump loss of forced reactor coolant flow initiated from conditions at or just above the CNBR SAFDL.
i To uss up the initial margin, the increased heat removal must occur in the i
presence of a negative Moderator Temperature Coefficient (MIC). However, a v negative MTC has a beneficial effect for a 4 pump loss of forced reactor flowt increased temperature rise across the core before reactor trip would provide r egative reactivity f eedback with a negative MTC present. ,
The lindting MTC f or the event is then a balance between these two competing effects. A previous parametric study on MTC valid for Cycle 6 determined that a value of -1.0$x10-4 Ap/'r was the most limiting MTC value.
i The following analytical steps have been used in the Cycle 6 analysis:
1.
The 1-D liERMITE code was used to model the flow ceastdown and reactor trip portion of the event.
2.
The CPC Low DNBR Trip la credited immediately after the beginning of the coastdown of the RCP's. This is done rather than_ waiting for the pumps to slow to the CPC Low Pump _ Speed setpoint. Modeled this way, the time from initiation of the cea tidown until the power is interrupted to the CEA holding coils is reduced from 0.86 seconds to 0.35 second.
- If the CPC Low DNBR trip setpoint was not[immediately reached,=it would. f mean that additional initial thermal margin must be present at the onset of the coastdown. This additional margin would result ~in fewer fuel rods experiencing DNB.
t 3.
To avoid unrealistically combining the worst case MTC with the worst case pin census, burnup dependence of MTC was also accounted for.
The Cycle was divided between the times'in life in which the MTC was 1 more negative than -2.0*10 4 Ap/'r and those in which the HTC was less i negative than this value. Thus, a .value of -1,05 *10 Ap/'F_was assumed-for-2.0*10-4 of approximately the first 10,300 MWD /T burnup of cycle 6 and a value ,
t Ap/'r was assumed f or burnups larger than this.
4 The results from HERMITE cases based upon these MTC. values were used to evaluate fuel pin DNBR. The-percent of fuel pins violating the SAFDL was calculated based on a bounding pin census chosen for each of ,
the two burnup ranges.
l Table 7.1.3-1-contains the sequence of events for'the Increased Main Steam Flow event with Loss of AC,. based on_this trip timing.'The Excess Load portion of the event begins at time _ T - 0. The _ initially preserved thermal margia is -
todaced until some time, AT, at which time the core has' reached a point 1)ust
the RCS.
the rate and~ severity with Which the-Increased Main-Steam' Flow is imposed upon '
' At time AT the Loss of Of fsite Power is assu:ned to occur. The CPC's sance the decreased flow and generate a Low DNBR Trip. The power;to *5e holding coils'is removed at 4T + 0.35 secondst At AT + 0.95 seconds the flux in_the holding coils _has decayed and the.CEA's begin-to drop into the core.
1-11 F
=w-eve.,--*ee,se-.m- . r-----*-,tr--.s---+ --w---v'r,w y -ir - 3 +- ,.---,,-rw--ee,ywr prw-,4-e-- i-av .-qm,- g wWy p- k .y e -++v y w'e e v y-t e.- w y- -+,w y ----w-----v*~9
t i
The transient DNBR reaches a minirnum value at AT + 2.1 seconds and then [
begins to increase. The value of this minimum DNBR is 1.01J, which is below the Cycle 6 DNBR SAFDL and more adverse than the Reference Cycle result '
! of 1.096. After this minimum value, event recovery proceeds as presented in the FSAR.
7.1.3.3 pesults r
The maximum number of pins predicted to experience DNB using the tr m thod of statistical convolutier, .s less than 3.0% for Cycle 6.
The predicted number of pins with DNBR values wMch decrease morrentarily below the SAFD1, could be reduced considerably if other analytical steps were included. Applying power penalties in the CPC neutron power calculations, detailed examination-of the coolant flow into the fuel assemblies with fuel pine predicted to fail, and an explicit calculation of the cooldown and actual loss of operating margin are analytical steps which, if further applied, would ,
reduce the reportori amount of fuel pins experiencing DNB.
4 As stated in the Watwrford 3 FSAR, for this event, eventhoughsomefuelpinsexperienceOffh temperature would b>,
f t.1 Aamage would not be expected since the maximum clad Vd, failure.
iar less than clad temperaturus whicn could lead to clad I 7.1.3.4 Gonclusions With the assumption that no additional thetm l margin has been-set aside at the beginning of the LCAC portion of the event, the maximum number of pins predicted to experience DNB is less than 3% for Cycle 6. This is acceptabl6 since this Further, results in only a small fraction of the fuel experiencing DNB.
while fuel fallure is not expected, a coolable core geometry would be maintained even if the 3.0% of fuel calculated to experience DNB failed. This event is presented because the predicted maximum number of fuel pins in DNH for Cycle 6, while acceptacle, i s not bounded by the Waterford 3 -
Ref erence cycle results f or this event.
1 l'
l 1-12
. . ' , , _ _ , - . ._ .. -.. -- ,,,,#. .~r.--y. r-,- -- v'C * * ' ' * ~ ' * ~ * " * #
._. . . . .. . - = . . . ~ - _ . . ~ . - _ . . - . . .
Table 7.1.3-1 ,
Sequence of Events for the Increased Main Steam riow. '
in Combination with a L,rs of AC Power-1 Time (sect Event Setpoint or Value 0.0 Malfunction of Control System causes ----
increased steam flow through the Turbine or the Turbine Bypass Valves AT Thermal Margin Initially Preserved by ----
COLSS Depleted, the Hottest Fuel Rod is Just Above the DNDR SAFDL as Calculated by the-CPC's. ,
Loss of AC Power 'is Assumed - ,
and the Coast Down of the RCP's'Begins. "
AT & 0.35 Low DNBR Trip Generated'by the CPC's, 1.26 Trip Breakers Open AT + 0.95 CEA's Begin to Drop ----
f AT + 2.10 Minimum DNER Occurs l'.076 .
- i. AT + 3.429 1.verage CEA Position 90% Inserted- ----
AT + 6.0 Stea.- Generatur Safety Valves open 1070 PSIA ST + 11.0 Maximum Steam Generator Pressure 1124 PSIA 1,800 operator t-akes contaal of plant ----
14,000 Shutdown Cooling Initiated ----
e l~
7-13
,. . . - . ~ . . . . - . . . . . . . - , . . . . . . . _ . ;.-_..._ .--a.. ._ ._ . - - -,
. . . - - . - . . . _ . - - - . ~ . . . . -.--- - - -.. - . ... .--~... .w .. - - , . .... - -.
- i
. i 7.1.4 Inadvertent Onenina of a Steam Generator Safety Valve or I Atmospheric Dumo Valve The'results are bounded by the Reference Cycle 7.1.5 Steam System Pipino Tallures:
The results are bounded by the Reference Cycle. _
7.2 s DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM-The results. are bounded by the Reference. Cycle' for all events in this category. ,
7.3 DECREASE IN REACTOR COOLANT TLONRATE '
The results are bounded by the Reference Cycle for all events in this category.
7.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES The results are bounded by the Reference Cycle for all events in i this category.
7.5 INCREASE IN REACTOR-COOLANT SYSTEM INVENTORY
. The results.are bounded by the Reference Cycle for allievents in this category.
7f6 DECREASE IN REACTOR COOLANT-SYSTEM INVENTORY The-resuits are bounded by the Reference Cycle for-a.11 events in this ategory.
7.7 ~
MISCELLANEOUS - ( Asymmet ric Steam Generator Events)
The resolts are bounded by the Reference Cycle. -
[.
6 i
I
'a
~
I 7'-14 p g---r,q- g.y e-*- --. y q g. -q ip- g.pg g y-g y .g .p- .wy&yg M,pey myw- g -N s-#4 y w- yr w-*-egy .ig-g- g g ) t g& y -1 <> -g--+-rwy-,.p'-gyw w w-wp e y
. . . _ - _. _ ~ _ . - . . _ _ _ __ _ _. _ -._ _ . , _ _ _ , _ _ - - -
8.0 ECCS ANALYSIS B.1 LARGE BREAK LOSS OF COOLANT ACCIDENT (LOCA) 3.1.1 Introducticn An Emergency Core Cooli.;g system-(ECCS) performanceJanalysis of the limiting break size was performed for Waterford 3 Cycle 6 to demonstrate compliance with 10CFR50.46 (Reference B-1), the NRC Acceptance' Criteria for Light Water Reactors. The analysis justifies an allowable Peak Linear Heat Generation Rato iPLHGR) of 13.4 kW/ft. This PLHGR is equal to tho existing Waterford 3 -limit.
The method of analysis and'cetailed-results which. support this value are presented in the following sections.
B,1,2 Methods of Anal _v sis' The ECCS performance analysis for .Waterford 3 Cycle 6 consisted of an evaluation of the dif ferences between Cycle 6 and the Reference Cycle .
analysis, Acceptable ECCS perf ormance was demonstrated for -the Reference Cycle in Reference inclume: B-2. The dif f erences between Cycle 6 and the Reference Cycle
- 1) diff.erences in fuel cycle related parameters, to the debris resistant fuel design wnich was introduced inand 2) c.4anget doen Cycle 5 and previously evaluated for its offect on ECCS p-erformance, The previous evaluation of the debris: resistant fuel design concluded that.the blowdown and refill /refloca-.hydrau.lica calculations performed for the Reference peak Cycle remain applicable by including a 3*F penalty to the calculated _
clad temperature.
Ournup dependent NRC approved fuel-performance calculations were code performed using the FATES 3B version of CE's (References 8-4, B-5,
-STRIKIN-II (Reference 8-7) code to determine the limiting-.and 8-6) and:the-fuel rod conditions for use in the Cycle 6 ECCS perf ormance evaluation.
3,1.3 Results-l i_
I Table 8-1 presents aL ccmparison of the significant . parameters for Cycle 6 'and the Reference Cycle, The initial system. flow rate,'coreuflow rate,
-and outlet core' inlet temperatures used in the Reference Cyclo analysis,are the same as those for Cycle'6..The pressure drops across the core along with the n"mber of-plugged U-tubes that per Steam Generator- are also identical. Therefore, it is 1 concluded ll the' blowdown-and refill /reflood: hydraulic calculations employed -
lin the Reference Cycle. analysis apply:to the Cycle 6 analysis.
e The limiting . fuel rod conditions f orf the Ref erence . Cycle _ and ' Cycle ; 6 are -
i compared in Table 8-2. The: hot ~ rod gas pressure _at theJ11miting burnup for Cycle 6 'is not significantly dif ferent from the . corresponding- pressure for the .
Reference Cycle. The-PLHGR in the average channel ,ofL'.he hot-assembly increased sligntly. 'The average fuel and centerlinen temperatures decreased and
{ the gap conductance decreased f rom the Reference Cycle. ' A hot rod temperature ._
[. transient calculation demonstrated-that the less' favorable radiatien-heat
' transfer characteristics are not . offset by tne lower average-fuel and centerline temperatures .f or . Cycle 6. Figure '8-1 shows a plot, of PCT -versus time for the hot spot location, .
3-1
, _ - _ m. _ _ _ .- ,_ . _. . - _ . . .__ __ . . _ . - - _ . . _. . . _ , . -.- _ . . . , _ _ -
~
The peak clad temperature and maximum 1ocal clad oxidation values of 2173*F and B.4%, respectively, f or the Cycle 6 analysis exceeds the corresponding values f or the Ref erence Cycle analysis of 2150 *F and 7. 93% . The 2173
- F Cycle 6 peak clad temperature includes the 3*F PCT penalty due to the debris resistant fuel unchanged fromdesign.
that The core wide oxidation value for Cycle 6 is < 0.805%,
reported f or tae Reference Cycle.
B,1.4 Conclusion The ECCS performance analysis for Waterford 3 Cycle 6 resulted in a peak clad
~amperature of 2173*F, a maximum local clad oxidation of 8.4%, and a core wide oxidation of < 0.805%. While the POT and local clad oxidation values exceed the values previously reported for the Reference Cycle, these values are less than the acceptance limits of 2200*F and 17%, respectively. The core wide oxidation value is unchangcd from chat reported for the Reference Cycle and is less than the acceptance limit cf 11. Therefore, operation at a PLHGR of 13.4 kW/ft and a power level of 3458 MWt (102% of 3390 MWt) will result in acceptable ECCS performance for Cycle 6.
3.2 ..
I.N*
SMALL BREAK TOSS CF COOLANT ACClyENT S.2.1 Introduction #P An ECCS performance analysis of the limiting break size was performed for Waterford 3 Cycle 6 to demonstrate compliance with 10CFR50.46 (Reference B-1),
the NRC Acceptance Criteria for Light Water Cooled reactors ~.
B.2.2 Method of Analysis Tre ECCS performance analysis for Waterfnrd 3 Cycle 6 consisted of an-evaluation of the differences between Cycle 6 and Cycle 5 and a comparison to the Reference Cycle analysis. Acceptable ECCS performance was demonstrated for the Reference Cycle in Reference 8-3. Acceptable ECCS performance for Cycle 5 was documented in the Cycle 5 Reload Analysis Report. It was determined that all input data for Cycle 6 are bounded-by the' Cycle 5 data. '
B.2.3 Results The peak cladding temperature, maximum local cladding oxidation,'and core wide oxidation values of 1846*F, 1.7%, and 0.28%, re pectively, for the Reference- '
Cycle apply conservatively cc Cycle 6.
B.2.4 Conclusions The Cycle 6 Small Bream LOCA results Cycle remain bounded by the Reference Cycle.
6 Small Large Break LOCA, Break LOCA results are less severe than those for the Cycle 6 which remains limiting.
1 8-2
4 Table 8-1 z
Waterford-3 Cycle 6 ECCS Analysis.
Significant System Parameters' .,
Reference
. Parameter Cycle Cycle 6 i t
Reactor Power Level (102% of Nomina 1,) MWt 3458 -3458 System Flow Rate (Total), 1bm/hr 148.0x106- 148.0x106 i Core Flow Rate, Ibm /hr 144.0x106 -144.0x106:
-l Core Inlet Temperature, 'r 557,5 557,5. .i Core outlet Tempe ra tu re, 'F 618,6 }
613 :
Number of Plugged U-Tubes per Steam Generator 400 '400
-f
.1 i
.I i
I J
-t l
I l
)
l
Table B-2 Waterford 3 Cycle 6.ECCS Analysis Sionificant _jel F Pin Parameters-Reference Parameter III Cycle Cycle 6 Peak Linear Heat Generation Rate, Hot Assembly, Hot Channel, kW/ft .13.4 13.4 Peak Linear Heat ~ Generation Rate.
Hot Assembly, Average Channel, kW/ft 32.158 12,409 Fuel Average Temperature at PLHGR, 'T 2111.3 2102.1 Fuel Centerline Temperature at PLHGR. 'r '3321.6 3290.9 Gap Conductance at PLHGR, BTU /hr-ft2_.F 1534.0 1513.O Hot Rod Gas Pressure, psia
'1113.3- =1114,5.
' Hot Rud Burnup, MWD /MTU 1000,0' 1000.0 di l The values are at the limiting hot rod burrup-as calculated by STRIKIN-II I:
9-4
. . . - . . _ _ . - 2_.. .._.~ . , . . . . _ , - . s . . . . . . . . ..~..;-_,.--.,-.~.J.-..--,, . . , . ..=.__,_.....,_...-..a..
i i
. l I
figure 8-1 Waterford Unit 3 Cycle 6 ECCS Analysis PCT (Zire Reaion Temo) versus Time for the Hot Soot location 2200._ _
g -
1900 b
\ '
j'
/
x .' 3 o_ :
2 -
w H
1600 3
z o
.i.
a t' W1300 ce :
Im
- o '
CC -
N -
~
1000 g 700 [
l 400 '''' ''''
0 ' ' ' ' ' ' '1 0 0 200 300 400 500 TIME IN SEC
l 9.0 REACTOR PROTECTICN AND MONITORIN'i SYSTEM ,
9.1 Introduction The Core Protection Calculation (CPC) system is designed to provide the low DNBR and high Local Power .Dansity (LPD) trips to (1) ensure that the Specified Acceptable Fuel Design Limits (SAFDL's) on Departure from' Nuclear Boiling Ratio (DNER) and centerline fuel melting (i.e., Local Power Density (LPD) are not-exceeded during AOO's, and (2) assist the Engineered Safety Features System in limiting the consequences of certain postulated accidents.
The CPC system, 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, it.s subsystems, components, and parameters are maintainedfor Conditions within operating limits and Technical Specification Limiting.
Operation (LCO's).
9.2 CPC Software Modificarjans The algorithms associated witr tne CPC Improvement Fa 3-2, and 9-3) which were implt Program (References 9- (*
entedinCycle2areapplicableto' Cycle 6{.*?he values for the Reload Data Block (RDB) constants will be evaluated for waterford 3 applicability consistent with the Cycle design, performance, and safety analyses. Any necessary change to the RDB constants will be installed in accordance with Reference 9-4.
9.3 Addressable constants Certain CPC constants are addressable so that they can be changed as required during operation. Addressable constants include measurements (e.g., (1) . constants related' 'to coefficients, shape aanealing matrix, boundary point power. correlation factors), (2) and adjustments for CEA ' shadowing and planar radial peaking uncertainty f actors to account for processing and measurement
- uncertainties in setpoints, and (4) DNBR and LPD calculations miscellaneous (BERR0 through BERR4) , (3) trip t e.ns (e.g.
, penalty f actor multiplier, CEAC -
penalty factor time delay, pre-trip setpoints, CEAC inoperable flag,-
calibration constants, etc.). Trip setpoints, uncertainty factors, and other '
addressable constants will be dutermined for this-cycle consistent with the-sof tware ar d methodology established in the CPC Improvement- Program and the Cycle design, performance, and safety analyses.
9.4 Digital Monitoring System (COLSS)
The Core Operating Limit Supervisory System (COLSS), described in Reference 9-5, is a monitoring system that initiates alarms if ~ the LCO's on DNBR, peak linear' heat tilt rate, core power, axial shape index, or core azbmuthal are required, exceeded. The COLSS data base and uncertainties will be updated, as to reflect the reload core design.
l 9-1
.- . . . . . _ - . . _ . . . ~ . . . . . . . . . . - . - - . . . . ~ . . . . . . . . .~.. . .. . .. . . . . . . . _. .~.
= ,
10.0 TECHNICAL SPECIFICATIONS There are no. Technical Specification changes required as-a result of the Cycle 6 eore design and safety analyses.
i 9
I
+
t' ll l
l P
. I t
t +
L
~
to.1-L L
1
- , - - - - - - a- w a,y--- p a f,e-g --9 :-w-- z -y 4 r w ,* , --+,w e 4 *. e+vs -p + w. e.92 =g ,%,- e=--ww-e
4 11.0 _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 the tests also provide the -
data needed for adjustment of addressable constants in the Core Protection Calculator._ System (CPC'S). and in COLSS.
11.1 Pre-Critical Test 11.1.1 Control Element Assembly (CEA) Trip lest Pre-critical CEA drop times are recorded for all full length'CEA's at hot, full flow conditions. The drop times will be verified to be within Technical Specification limits.
11.2 Low Power Physics Tests 11.2.1 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 S/ stem (RCS) until the reactor is critical, i
11.2.2 Critical Boron Concentration (CBC)
The CBC is obtained for the All Rods Out (ARO) condition and for a partially rodded configuration.
Comparison to the predicted CBC is performed by c:mpensating predicted CEA position).
for the residual CEA worth (from the actual CEA position to the equivalent of +/- 1% aK/K The of the measured CBC's will be verified to be'within_the.
design predictions, 11.2.3 Temeeraturo Reactivity Coefficient The isothermal temperature coefficient (ITC) is measured at the Essentially All Rods Out (EARO) configuration and at a partially rodded configuration.
The average coolant temperature is varied and the reactivity feedback-associated with the temperature change is measured.
The measured value will be verified to be within +/---0.3 x-10-4-aK/K/*r of the predicted value.
The moderator temperature coefficient - (MTC) is calculated by subtracting the predicted value of the fual temperature: coefficient frem the ITC. .The-MTC-value is then verified to be within Tachnical 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 CEA's will be included' in thei measurement groups. Due to the large differences in relativeLCEA-group worths, two reference groups -(one with very. high worth a d one with medium 1
'11-1 ll l
, , _ . . . - , -- s 6 ---- A-~ - " - ~ " ' ~ ~ " ~ ~ ~' "~~ '
worth) may be used.
The group
- to be measured will be exchanged with the appropriate ref erence group, dependinn en their predicted worth.
The individual measurement group worths will be verified to be within +b- 15%
or + /- 0.1% AK/ K twhichever. is larger) of predicted values. In addition, the total worth of all the measurement groups will be verified to be within
+/- 10% of the predicted total worth.
11.3 Power Ascension Testino Following completion of tne Low Power Physics Test sequence, reactor power will be increased in accordance with norn4al .operat.Eng procedures. The power ascension will be monitored by an off-line NSSS performance and data-processing computer algcrithm. This computer code will be continuously-e.xecuted in parallel with the power ascension to monitor CPC and COLSS performance relative to the processed plant dcta against which they.are normally calibrated.
If necessary, the power ascension will be su1 pended while necessary data reduction and equipment. calibrations are performed. Thus -
the monitoring algorithm continuously ensures conservative CPC and COLSS operation while optimizing overall efficiency of the test program.
11.3.1 Reactor coolant Flow Reactor coolant flow will be measured ay calorimetric methods at steady state conditions in accordance with the Technical Specifications. Accaptance criteria will require that the COLSS indicated flow be conservative with respect to the measured flow and that the CPC indicated flow be conservative with respect to the COLSS indicated flow.
11.3.2 Fuel Symmetry Verification Fixed incore detector data will be examined at lower power to verify.that no detecta'le o fuel misloadinmi exist. Individual instrumented fuel assembly powers will be verified to be within +6-10% of the cymmetric group average.
11.3.3 Core Power Distribution ,
Core Power distribution data using fixed incere neutron-detectors will be used to further verify proper-fuel loading and to verify consistency between the l-L as-built core and the engineering design models. This is accomplished using measurement data f rom two power plateaus.
L l
Compliance with the acceptance criteria at the intermediate-power plateau tbetween 40 and 70% reactor power) gives reasonable assurance that the power distribution will remain within the design limits while reactot power is increased to 100%.
The final power-distribution comparison is performed at full power.-Axial'and radial power distributions are' compared to design. predictions as a final ,
verification that the core is operating in a manner consistent ~with its design within the associated design-uncertainties. !
The measused results are compared to the predicted values in the following manner for both the ' inte rmediate and full ' power analyses :
~
11-2
, - , ~L. , , , ,rmo r- e cy y, -
4 A. The root-mean-square (RMS) error between the measured and predicted radial power distribution for each of the 217 fuel assemblics will be verified to be less than or equal to 0.05.
B. The RMS error between the measu.ed and predicted axial power distribution for each of the 51 anial nodes will be verirled to be less than or equal to 0.05 C.
The measured values et planar radial peaking factor ( r'y) ,
integrated radial peak factor (F r), core average axial peak (F 2),
and the 3-D power peak q(F ) wi'l be verified to be within +/- 10%
of their predicted values.
11.3.4 Shape Annealing Matrix (SAM) and Boundarv Point Power Correlation Coefficients (BPPCC'S) Verification The S AM and BPPCC's are determined f rom a linear regression analysis of the measured excore detector readings and corresponding core power distribution determinea from the incore detector signals. Since these values must be representative for a rodded and unrodded core throughout the cycle, it is g\ '
desirable to use as wide a range of axial shapes as are available to estabIfsh their values. The spectrum of axial shapes encountered during the powcr ascension has been demonstrated to be adequate for the calculation of the matrix elements.
The incore, excore, and related data are compiled and analyzed throughout the power ascension algorithm.
by the off-line NSSS performance and data processing The results of the analysis are used to modify the appropriate CPC constants if necessary.
1 11.3.5 Radial Peakino Factor (RPF) and CEA Shadowing Factor (CSF) l Verification The RPF's and the CSF's are calculated using fixed incore detector and excore detector data from the following CEA configurations:
- All Rods Out
- Group 6 fully inserted
- Group 6 fully inserted & PLCEA's @ 37.5 inches withdrawn
- PLCEA's G 37.5 inches 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.
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 of f-line NSSS performance code to verify the adequacy of the CPC Temperatere Shadowing Factor constants.
11-3
. . _ _ . - -_ .- - . - . . - . ~ - ~ .- ..~. - .-... .
11.3.7- -. Re a c t i vit y Coefficients ~at Power The isothernal . temperature coef ficient ;-(ITC) is-measured at approximately full power by swinging turbine . lcad to alternately increase and._ decrease core :iulet temperature.
The ' swings in ; ore temperature-and power are-used along with the predicted _ power coefficient to calculate the.ITC, The predicted fuel temperature coefficient is.then subtracted from the ITC.to-obtain the MTC.
The measured MTC is then used to 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 test are not met, an evaluation will be performed by the plant's-reactor engineering group with' assistance from~the fuel vendori as needed. The resultsiof this evaluation will be presented to the Plant Cperations-Review Committee. Resolution will be required prior to continued power. escalation. If an unreviewed sar'ety '
question is involved, the NRC will be notified.
I L-i l-i 9
11-4
. ;-- . . . _ . - . ~ . . . _ . . _ . . .. , . , _ , _ . . . , . , . . - . _ . , , , . , . . , . , , . . , _ - , - _ . . _ , . .
. .. - _ ~ . _ . - . . . ~ . - . .. . . _ . - -. -- -- -. -. ,
12.0 REFERENCES
12.1 Section 1~.0 References (1-1) "Waterford Steam Electric Station, Unit No. 3 Final Safety Analysts Report," Docket No. S h382,-
(1-2) Letter from K.'d. Cook (LP&L) to George W. Knighton (NRC), "Waterford 3 SES,- Docket No. 50-382, Reload Analysis Report (RAR)," August 29, 1986.
(1-3) Letter from K. W. Cook (LP& L) to George W. Knighton'
-(NRC), "Waterford 3 SES,-Docket No. 50-382. Reload-Cycle 2 Reports," October 1, 1986.
(1-4) CEN-386-P, " Verification of. the Acceptability of a f1-Pin Burnup Limit of 60 MWD /Kg for Combustion Engineering 16x16 Pwr Fuel," June 1989.
(1-5) Letter, A. C. Thadani 'NRC) to A. E. Scherer (CE),
" Generic Approval of C-E Fuel Performance Code FATES 3B (CEN-161 (B ) -P , Supplement 1-P (TAC NO. - Ms17 6 9) , "
November 6, 1991.
(1-6) CEN-35 6 (V) -P-A, Revision 01-P-A, " Modified Statistical .
Combination of Uncertaintios," May, 1988 12.2 Section 2.0 References None 12.3 Section 3.0 References None
.12.4 Section 4.0 References (4-1) . D. R. Earles, ."WSES-3 Cycle 5 Reload Analysis.Paport",
L-90-040, December 12, 1990.
(4-2) CEN-386-P, " Verification of the Acceptability of.'a.1-Pin Burnup Limit of 60 MWD /Kg for' Combustion Engineering'16x16 PWR fuel," June 1989.
(4-3) CEN-161(B)-P Supplement 'l-P, " Improvements to Fuel Evaluation MMel', " Combustion Engineerit;g Inc . ,-.
August 1986.
(4-4) CENPD-139-P-A, "C-E Fuel. Evaluation Model Topical-Report," Combustion Engineering Inc., July 1974.
12-1 92B.NRCRAR6. doc l
-. , . . , . - - ._.,.-,,_.__._,.m .., ...._ . _ . . . _ . -
-- - . . . . .- . . _. . . . - . . . -.~ - -. .. . - . . . . .
t 4
4 (4-5)- CEN-161 (B) -P- A, "ImprovementSLto-Fuel Evaluation Model," C;tbustion Engineering Inc., August 1989.
(4-6) Lv; t e r . A. O, Thadani - (10U0) to A. E.-Scherer (CE),
"Ger.eric Approval of.C-E Fuel Performance Codo FATES 3B (CEN-161(B)-P,- Supplement 1-P (TAC NO. MS17 6 9) , "
November 6, 1991.
12.5 Section 5.0 References (5-1) MSS-NA3-P, " Verification of CECOR.Coefticient Methodology for' Application Pressurized Water' Reactors .t of the Middle South Utilities System," August 1, 1984.
(5-2) CENPD-266-P-A, "The-ROCS-and DIT Ccmputer-Codes for Nuclear.Oesign," April, 1983.
(5-3) CEN-386-P, " Verification of the - Acceptability fof a 1-Pin Burnup Limit of.60. MWD /Kg for combustion Engineering-16x16 PWR Fuel," June 1989.
(5-4) FIESTA One Dimensional Two Group Space Time-~ Kinetics Code for Calculating-PWR Scram Reactivities," CEN-122 -
November 1979.
(5-5) CENPD-275-P'-A, "C-E Methodology for! Core.0esigns' Containing Gadolinia,Urania Burnable Absorbers," - ,
May 1988, 12.6 Section 6.0 References (6-1) CENPD-161-P-A, " TORC Code,?A' Computer'.Codeifor:
Determining the Thermal-Margin ofta 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,"cSeptember.1976.
(6-3)
CENPD-206-P-A, " TORC Code, Verification and Simplified -
Modeling Methods,"-June-1981. :
(6-4) CEN-160 (S) -P', ~ nev. 1-P, "CETOPfcode_Strudture and-
!! -Modeling Methods.for San Onofre Nuclear-Generating
\< Station Units 2 and 3," September 1981.
l_ (6-5) CEN-283 (S)-P,: " Statistical Combination ofi Uncertainties, Part 1, Combination of System Parameter
. Uncertainties in Thermal. Margin _ Analysis for San Onof re Nuclear Generating : Station Units ' 2 ' and 3, ": June-1954.
12-2 *
.. . -. . . = - . -
(6-6) CEN-155-(3)-P, "CE-1 Applicability to San Onofre Units l
2 and 3 HID-2 Grids, Response to NRC Questions," March-1981.
(6-7) CEN-16 5 (S ) -P, " Responses to NRC Concerns en l.
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," Dccket No. 50-382, October 1991.
(6-9) CEUPD-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).
$, 1 12.7 Section 7.0 References I'c i (7-1) "Waterford Steam Electric Station Unit'No. 3, Final Safety Analysis Report," Louisiana Power and Light Co., Docket No. 50-382.
l (7-2) Letter from K.W. Cook (LPL) to George W. Knighton (NRC), "Waterford 3 SES, Docket No. 50-382, Reload Cycle 2 Reports," October 1,' 1986.
l i
(7-3) l "CESEC, Digital Simulation of a Combustion Engineering Nuclear Steam Supply System," Decerber 1981, Enclosure 1-P to LD-82-001, January 6,11982.
(7-4) R. V. Macbeth, "An Appraisal of Forced Convection Burnout Date," Proc. Instr. Mech. Engrs., Vol. 180, Pt. 3C, PP 37-50, 1965-1966.
(7-5) D. H. Lee, "An Experimental Investigation of Forced Convection Burnout in High Pressure Water - Part IV, ,
l Large Diameter Tubes at about 1600 psia," A.E.E.W.
Report R479, 1966.
(7-6) CEN-160 (S) -P, Rev. 1-P, "CETOP Code Structure and l -Modeling Methods for San Onofre Nuclear Generating l Station Units 2 and 3," September 1981.
l l- (7-7) CENPD-1BB-A, "HERMITE Space-Time Kinetics," July-1975.
(7-8) CENPD-161-P, " TORC Code --A Camputer Code for Determining the Thermal Margin of a Reactor Core,"
July 1975.
12-3
. . . . . - . . - - - - _ . - --.- - .-;- -. .. , . . ~ _
(7-9) CENPD-206-P, " TORC Code Verification and Simplified-Modeling Methods," January 1977 (7-10) CEMPD-lS3, " Loss of Flow - C-E Methods for Loss of Flow Analysis," July 1975.
(7-11) D. R. Earles, "WSES-3 Cycle 5 Reload Analysis Report",
L-90-040, December 12, 1990.
.' 2 . 9 Section B.0 References (B-1) Acceptance Criteria for Emergency Core Cooling Systems for Light. Water Cooled Nuclear Power Reactors, Federal Register, Volume 39, No. 3, January 4, 1974, 18-2) Letter from K. W. Cook (LP& L) to NRC, "WSES Unit 3, Docket NO. 50-382, Reanalysis of the Large Break LOCA-for Cycle 2", May-18, 198' (8-3) "Waterford Steam Electric Station Unit-No. 3,. Final Safety-Analysis Report," Louisiana Poweroand L'.ght-Co., Docket No. 50-382.
I (B-4) CENPD-139-P-A, "CE Fuel Evaluation Model Topical
. Report," July 1974. .
(8-5) CEN-161(B)-P-A, " Improvements to Fue1~ Evaluation Model," August 1389.
CEN-161 (B) -P , Supplement 1-P, " Improvements'to Fuel Evaluation Model," April.1986.
(8-6) Letter. A. C..Thadani (NRC)hto'A..E. Scherer - (CE) ,
" Gene-ic Approval. of' C-E Fuel Performance Code FATES 3B (CEN-161(B)-P, Supplement 1-P (TAC NO; M21769) . "
November: 6, 1991.
! (8-7) CENPD-135-P,."STRIKIN-II,'A Cylindrical Geometry Fuel-Rod Heat Transfer Program," April 1974.
CENPD-135, supplement 2-P, "STRIKlN-II,: A Cylindrical Gecmetry Fuel _ Rod Haat Transfer Program -
-(Modifications),": February 1975.
CENPD-135-P, Supplement.4-P,. "STPIKIN-if, A .
Cylindrical Geometry Fuel Rod Heat Transfer Program,"
August 1976.
CENPD-135-P,' Supplement 5-P,. "STRIKIN-II, A:
Cylindrical-Geometrv Fuel Rod Heat Transfer Program,"~
April 1977~ .
l 12-4.
_ _ _ _ - ~ ~
12,9 Section 3.0 Refeterces (9-1)
CEN-304-P, Rev. 01-P, " Functional Design Requirements for a Control Element Assembly Calculator," May 1986.
(9-2) CEN-30$-P, Rev. 01-P, " Functional Design Requirement f or a Core Protection Calculator," May 1986.
(9-3)
CEN-330-P-A. "CPC/CEAC Software Modifications for the CPC Improvement Program Reload Data Block,"
October 1987 (9-4) CEN-323-P-A, " Reload Data Block Constant Installation Guidelines," September 1986.
(9-5) ,CEN-312-P, Rev. 01-P, " Overview Description of the Core'Opsirating Limit Supervisory' System (COLSS ) , "
a November 1986, 1
12.10 Section ~10.0 References None 12.11 Section 11.0 References None 1
'12 r- -- -.----___=..L_. --_.
- . I I
APPENDIX A ,s l
.TO WATERFORD 3 CYCLE'6 RELOAD ANALYSIS REPORT '
EVALUATION OF CHANGE TO NUCLEAR DESIGN ME".7 ,DS (1) Descriptian of Chance The original methods and computer codes used ,o analyze the nuclear design of the core are described in Chapter.4 of the Waterford 3 Safety Analysis Report.
A licensee'is allowed to make changes to these methods and codes provided that.
the change does not involve either a' Technical Specification change or an Unreviewea safety Question.
The nuclear design nethods and computer codes provide calculated values for the f ollowing nuclear design parameters :
Reactivity .
Reactivity Coefficients a
Control Rod Worths Peaking Factors Power Distribution Related Factors l
Several changes have been made to these methods and computer-codes to L
! (1) . simplify their use, -(2) improve their computational efficiency (e.g.,
' the exchange of data between : codes), and- (3) ; enhance their calculational' accuracy.
Of the- three types of changes,' only. the -latter, enhancing their calculational 1 accuracy, is most likely to significantly affect the' numerical results. Since the results of nuclear design analysis are-used:as input'to the transient.
safety analysis that considers accidents and malfunction cf equipment important to safety, . these changes must be evaluated to determitte whether or notlan unreviewed safety gaestion is created.
The original nuclear design methods and. computer codes are described in CE's proprietary Topical Report CENPD-266-P-A, "The ROCS & DIT Computer--Codes for l
Nuclear Design," dated April 1993. This Topical Report was generically reviewed and approved by the NRC1, Subsequent to the NRC's approval, changes A-1
were made to the methods and codes that could ef fect the calculational accuracy of the nuclear design computer codes. Tbsse changes are as.follows:
Implementation of Modal Expansion Method to ROCS Improved Accounting of Anisotropic Scattering and Higner Order Interface Current Angular Distributions in DIT Use of Assembly Discontinuity Factor? between ROCS and DIT Update of Biases and Uncertainties Applied to Calculated Parameters A description of each change is provided below. The descriptions provide suf ficient detail to perf orm a safety evaluation.
Nodal Expansion Method The Nodal Expansion Method (NEM) was added to the-ROCS codo as an alternative to the original Higher Order Difference (HOD) formulation. The ROCS code 4.
providesreactorpowerdistributionsandeffectiveneutronmultiplication[,\
factnrs.
This data is-then used to derive control rod worths, depletion, 6'-
react.vity significant coefficients and reactivity differentials, Use of the NEM achieves reduction in computer running times and also improves agreement-with fuel managemont measurement data 2, Although the NEM had not i yet been fully integrated into the ROCS code, the use of the NEM was fully described in CE Topical Report CENPD-266 that was approved by the NRC.
Specifically,-Topical Report CENPD-266 explained that UEM had been incorporated into s version of C-E's coarse-mesh kinetics code.
HERMI*E. Furthermore, Topical Report CENPD-266 presented numerical comparisons of the NEM.and HOD methods for solving the neutron diffusion equations. The results showed that the substitution of NEM for the-HOD method in ROCS would not have a significant impact on calculational results and uncertainties.
In recognition of thu expected future implementation of the NEM into ROCS, the NRC stated the'following in the Safety Evaluation Report (SER) that approved C-E Topical Report CENPD-266:
"We have reviewed the ROCS and DIT compater codes as described in CENPD-266-P and CENPD-266-NP and find them to be acceptable for nuclear core design and safety-related neutronics: calculations made by CC in licensing actions for power distributions, ' control rod worths, i
differential.
depletion, : reactivity coefficients and reacti vity
- We also conclude' that the ROCS code, including the I fine-mesh module MC, is of sufficient accuracy for the generation of coefficient libraries for the in-core instrumentation.
The -staff, however, recommends that CE perform .further verifica tion - when the NEM is incorporat! , into the _ ROCS _ code in order ' to be assured that equivalent salculational biases and-uncertainties are obtained with ROCS-NEM as compared to ROCS-HOD. "
A-2 m .
.~ --- ~ -- -_... - -_- .. . -. - __ - ~ . -- -- -
Before using ROCS-NEM for nuclear design analysis for Waterford 3, CE perf ormed f-ther verification to confirm that the calculational biases and uncertainties obtained with ROCS-NEM are equivalent to ROCG-HOD. The SER did-not requira CE to resubmit the ROCS-NEM version of the code tolthe NRC for ,
approval. It is important to note,Lhowever, that the NRC did recommend that the biases and uncertainties obtained_when NEM was incorporated into ROCS be-equivalent-when comptred to ROCS-HOD. 3y equivalent, it is understood that the results between the two methods need not be numerically identical, but rather that the two methods be equal to the degree that the same conservative relationship is maintained between calculated.and measured data (i.e., a 95/95 tolerance limit). ,
i CE has confirmed that the ROCS-NEM nuclear core design and safety-related neutronics calculations of power distributions, control rod worths, depletion, reactivity coefficients and reactivity differentials maintain the same conservative relationship between calculated and measured data. In particular, the tolerance limits applied to the calculated results from ROCS-HOD and ROCS-NEM are identically defined as "the value that must.be added to the calculated results to assure that 95% of the alculated values will be greater than the "true" value with 95% confidence." Thus, che change which adds NEM to ROCS has been demonstrated to.be equivalent to the ROCS-HOD version, which was~ approved by the NRC.
Anisotropic scattering and Higher Order Interface Current' Angular Distributions j
- In order to maintain tne calculational accuracy in CE Topical Report CENPD-286 when evaluating fuel containing gadolinium as a burnable pmi on,'OE had to improve the way the nuclear design computer code acccunted for the ef fects of l
i anisotropic scattering and higher order interface current _ angular distributions in the DIT code. The DIT code is a transport theory-bar 1 code-which performs spectral and spatial calculations'in fuel cell and fuel assembly geometries.
The DIT calculations provide few group neutron cross sections for use by-the ROCS code.
The improved method for accounting for anisotropic scattering and higher order interface current angular distributions was. submitted by CELin a generic .
Topical Report which was reviewed and approved by the' NRC3 Thesel approved methods and computer codes are described in CE Topical Report ~CENPD-275-P 1
Revision 1-P-A, "CE Methodology for Core Designs Containing Gadolinia-Urania Burnable Absorbers," dated May 1988. Although these changes were motivated by -
the need to obtain additional calculational accuracy to analyze' gadolinium as a-burnable poison, the method itself is independentcof the' burnable absorber-used in the core.
-Topical Report CENPD-275.was not submitted on a plant specificEdocket. It was reviewed by the NRC for generic implementation on PWR cores. In recognition of the' generic applicability'of'the improvements made to the BIT codes the NRC stated the following in the SER that approved CE: Topical Report CENPD-275:
\ 1 "We have reviewed the changes 'made to the DIT - and ROCS /MC codes -
and methodology to accommodate ihe use of'the integral burnable )
absorber gadolinium in PWR cores. These changes are ; typical; of the types made by the induscr.y for computing gadolinia ' cores. The A ,
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numerical results that were provided show that. acceptable agreement has been obtained between detailed calculations "and design calculations.
We conc 1Jde therefore'that the changes made to the DIT and ROCS /MC coaes and methodology are acceptable. "
"We also conclude that the neutronics methods described in the
- eport (DIT, ROCS /MC, and FDQ), as modifiede ato ~ acceptable for calculating the neutronic characteristics of PHR cores containing up to 8 weight percent gaiolinia bearing fuel rods."
It is also important to note that benchmark analysis provided in Tupical Report CENPD-275 validated the changes made in the DIT code with B qC poison that contained no gadolinium. The NRC SER, thus, concluded that the methods described in Topical Report CENPD-275 are acceptable for calculating the neutronic characteristics of PMt cores containing up to 8 weight percent gadolinia bearing fuel rods. This includes the case where the PWR core contains zero weight percent gadolinia by virtue of the fact that many of the assemblies used for benchmarking purposes did not contain any gadolinium bearing fuel rods.
Indeed,'the NRC also noted in the SER the following; "The results obtained for the Lead Test Assemblics (LTA) are consistent with those obtained for the non gadolinium bearing- fuel assemblies.
- results provideTheadditional staff concurs valida with CE's tion of conclusion that these the DIT code and-methodology." .
\
Assembly Discontinuity Factors Assembly discontinuity. factors (ADF's) aro used in the nuclear industry 4 as a method to eliminate homogenization error ~in nuclear design analysis where the global heterogeneous solution is known. The-use of ADF's improves the internal agreement between the DIT and ROCS-codes. The ADF's are derived-from the very assembly calculations required by the conventional homogenization methods and, therefore, they do not. add any new information to the.overall calculational methodology. Thus, the use of the ADF'sLis, expected.to improve i
the accuracy of results obtained from ROCS when compared ~te DIT. CE has confirmed that the assembly discontinuity factors improve the accuracy of the nuclear design analysis method and computer codes.
Biases and Uncertainties _
l In view of the above changes that have ueen maw to the methods and nuclear
~
design cceputer codes,J the biases-and uncertainties applied to the nuclear design parameters were F::;. ally reevaluated by CE. .For nuclear design parameters, .the biar represents either the average of measured value minus the. j calculated valuc, or the average ratio of the-differenceLbetween the measured .
l value'and che calculation value compa' red to.the calculated value. .The !
L !
l-uncertain.y-value-represents the 95/95 tolerance range for the parameter ,of '
interest.
l The reevaluation: produced revised bias and uncertainty values that are !
equivalent to:those reported in CE Topical Report CENPD-266. By equivalent, .'
it is-meant that-the results are not numerically ~ identical, but.rather-that i
their appilcation preserves the same. conservative statistical. relationship-
- 4 a
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between calcu13ted and measured data (i.e., the 95/95 probability / confidenen level).
The methods used to generate the new biases and uncertainties are the same as that described in Topical Report CENPD-266, with the exceptions of the method used to determine the bias and uncertainty f or the net _(N-1) rod worth. In the Topical Report, the bias and uncertainty associated with net (N-1)' rod worth were calculated by evaluating individual bank worth measurements.
When CE reevaluated the bias and uncertainty for the (N-1) configuration, CE used the bias and uncertainty associated with the sum of the bank worths (i . e . , " total" worth) in lieu of that for-individual banks. The use of the total rod worth uncertainty is considered more appropriate than the individual bank worth since the total rod worth configuration is more representative of the higher control rod density of the (N-1) configuration.
This change in the bias and uncertainty used for the (N-1) case remains conservative because actual (N-1) measurements demonstrate that the uncertainty of the (N-1) rod worth is lower than the uncertainty of the total worth.
This is expected since the (N-1) configuration is strongly influenced by the reactivity of the unrodded region of the core. Thus, the (N-1) configuration is less sensitive to the precision of the calculated effective control rod cross sections than are either the total or individual bank configurations.
The change in method to calculate the (N-1) rod worth produces equivalent set of bias and uncertainty,'wherein the same conservative relationship is maintained between calculated and measured data (i.e., a 95/95 tolerance limit).
(2) Unreviewed Saf ety Ouestion Determination i
The changes to the nuclear design analysis methods and computer codes described above do not require changes to the Technical Specifications. No-unreviewed safety question exists, (3) Safety Evaluation The determination that the changes to the nuclear design analysis methods and computer codes described above do not create an unreviewed safety question is demonstrated by the following:
- 1. The probability of occurrencelor the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety = analysis report will not be increased by the changes to the nuclear design analysis methods and computer codes described above.
The results of nuclear design analyses are usta as -inputs to the analysis of accidents or malfunction of equipment important to safety that are e*raluated in the safety analysis report. These inputs do not j
alter the physical characteristics'of any component-involved in'the-initiation of an accident or any subsequent equipment malfunction.
Thus, there is no increase ~in the probability of occurrence of an A-5
accident or malfunction of equipment important to safety previously evaluated in the safety analysis report as a result of this change.
The consequences of an accident or malfunction of equipment important to safety evaluated in the safety analysis report is aftected by the value of inputs to the transient safety analysis. There is always the potential for the value of_the nuclear design parameters to change solely as a result of the new reload fuel core loading pattern.
Rugardless of the source of a change, an assessment is always made of changes to the nuclear design parameters with respect to their effects on the consequences of accidents and equipment malfunctions previously evaluated in the safety analysis.
If increased consequences are anticipated, compensatory actions are implemented to neutralize any expected increase-in consequences. These compensatory actions include, but are not limited to, crediting any existing margins in the analysis or redefining the operating envelope to avoid increase consequences. Thus, the nuclear design parameters are intermediate results and by themselves will not result in a increase in the conssyuence of accident or malfunction of equipment important.t94 safety evaluated in the safety analysis report. '
{6es Therefore, the changes to the nuclear design analysis. methods and computer codes described above do not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important analysis report.
to safety previously evaluated in the safety ii. The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report will not be created by the changes to the nuclear design analysis methods and computer codes described above.
As noted abovo, the results of nuclear design analysis are used as inputs to the transient safety analysis of accidents or malfunction of equipment important
! report.
to safety that are evaluated in the safety analysis component These inputa do not alter the physical characteristics of any involved-in equipment malfunction the initiation of an. accident or any subsequent i Thus, there is no increase in the possibility of an accident or malfunction of equipment important to safety previously
- evaluated in the safety analysis' report as a result _of this change.
Thus, the changes to the nuclear design analysis methods and computer codes dcocribed above will not create the possibility for an accident or malfunction of a dif ferent type than any evaluated previously in the safety analysis report.
iii.
The margin of safety.as defined in the basis for any Technical Specification will.not be reduced by the changes to the nuclear design l
analysis methods and computer codes described above.
Benchmarking of the new nuclear design methods and computer codes _has.
demonstrated analysis are not that the values of those parameters used in the safety significantly changed relative to the values obtained A-6
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e * ,
l using the previous methods.and computer codes. For any changes in the
\
calculated values that do occur, the reevaluation of-the biases and-neertainties ensures that the current margin of-safety is maintained.
Specifically, use of these revised biases and uncertainties in safety evaluations continues to provide the ' sane statistical assurance that the.
values of the nuclear parameters used in the safety analysis do not exceed the actual values on-at least.a 95/95 probability / confidence basis.
The changes to the nuclear design analysis . methods and computer codes described above, therefore, doLnot reduce the margin of safety as defined in the basis for any technical specification.
In conclusion, the changes to the nuclear design analysis methods and computer codes described above do not involve an unreviewed safety question and does not require a change to the Technical Specifications.
References
" Acceptance for Referencing of Licensing Topical Report CENPD-266-P, CENPD-266-NP "The ROCS and DIT Computer Codes for Nuclear Design",
April 4, 1983
- 2. R. A. Loretz and S. G. Wagner,-"Recent Enhancements to the ROCS /MC Reactor Analysis System," Technical Paper presented at the International Conterence on the Physics of Reactor Operation, Design and Computation in Marseille, France,. April 23-27, 1990 3.
USNRC Letter from Ashok C. Thadani (NRC) to A. E. Scherer (CE),
" Acceptance f or Peterencing of Licensing Topical Report CENPD-275-P, Revision 1-P, "CE Methodology for Core Designs Containing Gadolinia-Urania Burnable Absorbers," May 16, 1988.
L 4 K.
E S. Smith, " Assembly Homogenization Techniques for-Light. Water Reactor Analysis,* Progress in Nuclear Energy, Volume 17, 1986, i
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J
' APPENDIX B TO - '
WATERFORD 3 CYCLE 6 RELOAD ANALYSIS. REPORT DEBRIS RESISTANT TUEL DESIGN-b b
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i 4
A debris resistant feature has buen incorporated-into Batch G -(the Cycle 5 .
reload Batch) = and follow Batch designs. The basic concept- of this feature is to move the fuel and poison column axially upward so that the debris' that is caught by the Inconel grid interacts with a lengthened solid end cap instead of the fuel or poison rod cladding.
The differences between the debris resistant design and the previous design are as follows:
1)
Lengthening the lower end cap on each fuel and poison rod so that the bottom-of the clad-is within the Inconel spacer grid assembly.
2)
Removing one fuel pellet from each fue. rod. This r. educes the' active fuel length from 150.-inches to_149.6 inches. Also, the plenum spring was redesigned to minimize its material volume. /.s a result of these two changes the interi>r void volume of Batch G and follow Batch fuel reds is within 0.5% of the value for Batch F and pre.vious Batches 3)
Removing one burnable poison pellet from each poison rod. This reduces the poison stacr',
length from 136 inches to 135 inches. Also, the upper spacer 0.7 inch, tube was shortened and the overall rod length was increased by 4)
Shortening the height of.the lower end fitting by 0.700 inch. The design of the lower end_ fitting has been-changed to improve the ease of installing the fuel bundle assemoly in the core and te standardize the debris-resistant fuel design. These changes to the lower end fitting design will not cause the calculated stress intens/ ties in the various load conditions to exceed the design limits.
5)
Changing the guide tube to lower end fitting connection to accommodate the 0.700 inch reduction in lower end fitting height.
The design of the Zircaloy and Incone' spacer grid assemblies remains unchanged.
The locations of Zircaloy grids:with respect to those in previous-fuel Batches remains unchanged. The InconeJ'7 rid in Batch G anc follow Batches is located 0.700 inch lower than it was for eviousl Batches.
Other minor refinements to the fuel assembly mechanical design made sir.ce Cycle 2 include:
L Thi perimeter strips for the HID-1 spacer grid-assemblies have the location of the corner impact arch lowered with respect to its previous location. This change improves fabrication of the rpacer grid assemblies.
l-The poison rod assembly design has been modified by replacing the solid Zircalcy-4 spacers with hollow Zircaloy tubes. By using hollow spacer tubes, :the poison rod internal . volume is increased,1 allcwing higher.
burnup poison rods.
Starting with Batch r (Cycle 4), poison-rod overall length'has been increased so th;t it is the same length s the fuel rod. This design:
B-2
- t -
I I 1
- e.
- i change will not have any effect on fuel assembly performance since \
poison rod growtn is bounded by fuel rod growth.
I The locking discs used in the lower end fitting conriection to the guide tube have been redesigned, enabling the use of one design in all CE fuel bundle assemblies. The redesign does not alter any design interface requirements between the fuel bundle assembly and the four alignment pins located on the core support plate, and does not affect the structural integrity of the guide tube connection.
l l
t a
B-3