ML17207A646

Forwards Evaluation of Design & Performance of Ref Core Configuration Utilizing Higher Enrichment Fuel.Requests NRC Approval by 800114 of Proposed Amend to Revise Max Enrichment Permitted by Tech Specs 5.3.1

ML17207A646
Person / Time
Site:	Saint Lucie
Issue date:	12/12/1979
From:	Robert E. Uhrig FLORIDA POWER & LIGHT CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
L-79-345, NUDOCS 7912140290
Download: ML17207A646 (57)

Text

pt REGULATORY FORMATION DISTRIBUTION S M (RIDS)

ASCESSION NBR0 7912140290 DOC O DATE ~ 79/12/12 NOTARIZED ~ NO DOCKET FACIL:50 335 St. Lucie Planti Unf t 1i Florida Power 8 Light Co, 0'5000335 AUTH ~ NAME 'AUTHOR AFFILIATION UMRIGg R ~ E ~ Florida Power 8 Light Co, RECIP,NAME RECIPIENT AFFILIATION EISENHUTrD,G. Division of Operating Reactors SU8JECT! For wards evaluation of design 8 performance of ref cot e configuration utilizing highet enrichment fue'I,Requests NRC approval by 800114 of proposed emend to, revise max enrichment permitted by Tech Specs 5,F 1 COBE: AOQIS TITLE: General COPIES RECEIVES:LTR ~ ENCL' SIZE:~+

'ISTRIBUTION

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P.O. BOX 629100, MIAMI,F L 331S2 y<1lbg fi~~4xh FLORIDA POWER S LIGHT COMPANY December 12, 1979 L-79-345 Office of Nuclear Reactor Regulation Attention: Mr. Darrell G. Eisenhut Acting Director Division of, Operating Reactors U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Dear Mr. Eisenhut:

Re: St. Lucie Unit 1 Docket No. 50-335 Reference Extended 6 cle Submittal On October 4, 1979 (L-79-282), Florida Power & Light Company (FPL) submitted a proposed license amendment to revise the maximum enrichment permitted by Technical Specification 5.3.1 to permit greater flexibility in assigning future core design features and associated operating cycle lengths for St. Lucie Unit l. In order to complete the review of that proposed amend-ment, the staff requested that FPL submit an evaluation of the design and performance of a reference core 'confi gurati on utilizing higher enrichment fuel. In accordance with that request, FPL here-with submits such an evaluation.

The fourth fuel cycle for St. Lucie Unit is the first cycle for 1

whi ch the higher enrichment fuel can be utilized, consequently the attached evaluation is based on a higher enrichment. core configur-ation which could be utilized for cycle 4. The higher enrichment core characteristics have been examined with respect to the safety analyses for St. Lucie Unit 1, Cycle 3 and, in all cases, the Cycle 3 safety analyses envelope the new conditions.

FPL must receive NRC approval of'he proposed amendment by January 14, 1980 in order to proceed with the higher enrichment core design.

A decision affecting manufacture of the high enrichment fuel for Cycle 4 must be made at that time.

The attachment has been reviewed and approved by the Florida Power &

Light Company Nuclear Review Board and the St. Lucie Facility Review Group.

Ver tr y yours, Rober E. ri Vice President Advanced Systems & Technology REU/MAS/rel Attachment )b cc: Mr. James P. O'Reilly, Region II Harold F. Reis, Esquire ') 918140 QQ(g PEOPLE... SERVING PEOPLE

ST. LUCIE UNIT 1 0

REFERENCE EXTENDED CYCLE SUBMITTAL I. INTRODUCTION AND

SUMMARY

This repor't provides an evaluation of the design and performance for the operation of St. Lucie I during its four th fuel cycle at the full rated power of 2560 t1WT. Operating conditions remain the same as those t

f'r Cycle 3. The core will consist of'resently C

operating Batch C, D, and E assemblies together with fresh Batch F =

assemblies.

System requirements have created a need for flexibility ib the Cycle 3 burnup length ranging from-7250 to 8250 tND/T. The Cycle 4 loading pattern described in this report has been designed to accommodate this range of shutdown points. In performing analyses of postulated accidents, determining limiting safety system settings and establishing limiting condi'tions for operations, values of key parameters were chosen to assure that expected. conditions are enveloped within the above Cycle 3 burnup range.

The sleeving of CEA guide tubes caused by wear of the CEA fingers follows the same procedure as reported for Cycle 3 in Reference 1 .

For Cycle 4 operation, only sleeved assemblies will be placed under CEAs and all 88 Batch F ass'emblies will be sleeved.

The evaluations of the reload core characteristics have been examined

.-with respect to the safety analyses describing Cycle 3, (Reference 2) hereafter referred to as the "reference cycle". In all cases, it has been concluded that the reference cycle. safety analyses properly envelope 1

the new conditions. The result of this evaluation is that'the operation of Cycle 4 requires only one Technical Specification change entailing an increase in allowed enrichment from 3.1 w/o to 3.7 w/o U-235.

2. OPERATING HISTORY OF THE REFERENCE CYCLE St. t,ucie Unit I is presently operating in its third fuel cycle utilizing Batch B, C, D, and E fuel assemblies at a licensed core power level of 2560 t'lAT. Operation of Cycle 3 has continued at or near

'icensed power.

It is presently estimated that Cycle 3 will terminate during March 1980.

To allow for flexibility in the Cycle 3 termination date, a range of burnups between 7250 and 8250 tQD/T has been anticipated. Operation of Cycle 4 is scheduled to commence in tray or June 1980.

3. GENERAL DESCRIPTION The Cycle 4 core will consist of the numbers and types of assemblies from the various fuel batches as described in Table 3-1. The primary change to the core for Cycle 4 is the removal of the remaining 21 Batch B assemblies and 67 of the 68 Batch C assemblies. These assemblies will be replaced by 40 Batch F (3.65 w/o enrichment) and 48 Batch F>> (3.03 w/o enri'chment) "assemblies. The 48 low enrichment Batch F* assemblies contain burnable poison pins with 12 pins per assembly. The location of poison pins within the lattice is the same as that for poison pin assemblies present in the reference cycle. The fuel management pattern developed for Cycle 4 allows for flexibility in Cycle 3 burnup length between 7250 and 8250 HWD/T. "

The loading pattern is shown in Figure 3-1.

The Cycle 4 core loading pattern is 90 degrees rotationally symnetric.

That is, if one quadrant of the core were rotated 90 degrees into its neighboring quadrant, each assembly would overlay a similar assembly.

This similarity includes batch tyne, number of fuel rods, initial enrichment and beginning of cycle burnup.

Figure 3-2 shows the beginning of Cycle 4 assembly burnup distribution for a Cycle 3 burnup length of 7750 t<WD/T. The initial enrichment of each assembly is also shown.

Tabl e '3-1 St. Lucie Unit 1 C cle 4 Core Loadin Beginning of Cycle 4 Batch Average Burnup HHD/HTU Initial Initial Number Shim Total Assembly Number of Enrichment (EOC 3 = of Loading Total Fuel Designation Assemblies w/o U-235 7750 HWD/T) Shims w/0 B4C Shims Rods 1 2.82 24,800 0 0 176 40 3.03 15,700 0 0 7,040 20 2.73 17,800 0 0 3,520

40. 3.03 6300 0 0 7,040 28

'.73 9300 0 0 4,928 f 40 3.65 0 0 0 7,040 f* 48 3.03 0 12 3. 03 576 7,872 217 576 37,616

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4.0 FUEL DESIGN

4. 1 Mechanical Design The fuel assembly complement for Cycle 4 is given in Table 3-1.

The mechanical design of the reload fuel assemblies, Batch F is identical to St. Lucie-1 Batch E fuel.

C-E has performed analytical predictions of cladding creep collapse, time for all St. Lucie-1 fuel batches that will be irradiated during Cycle 4 and has concluded that the collapse resistance of'll fuel rods is sufficient to preclude collapse during their design lifetime.

This lifetime will not be exceeded by the Cycle 4 duration. The results of this evaluation are shown in Table 4-1.

The analyses utilized the CEPAN computer code (Reference 3) and included as input conservative values of internal pressure, cladding dimensions, cladding temperature and neutron flux.

4.2 Hardware Modifications to Mitigate Guide Tube Hear.

II Batch C, D, E, and F fuel assemblies to be installed in CEA locations in Cycle 4 will have stainless steel sleeves installed in the guide tubes in order to mitigate guide tube wear.

A detailed discussion of the design of the sleeves and its effects on reactor operation is contained in Reference 4.

4.3 Thermal Design Using the FATES model (Reference 5), the thermal performance of the various types of fuel assemblies has been evaluated with respect to their 'Cycles 1, 2, and 3 burnups, proposed burnups'uring Cycle 4, their respective fuel geometries, and expected flux levels during Cycle 4. The Batch E fuel h'as been determi'ned to be the limiting fuel batch with respect to stored enerqy.

4.4 Chemical Design The metallurgical requirements of the fuel cladding and the fuel assembly structural members for the Batch F fuel have not been changed from the original Cycles 1, 2, and 3 designs. Therefore,

'he chemical or metallurgical performance of the Batch F fuel >>ill be unchanged from that of the original core fuel and discussions in the FSAR, Reference 6 are still valid.

4.5 Operating Experience Fuel assemblies incorporating the same design features as the St. Lucie Unit 1, Batch F fuel assemblies have had op rating "experiences at Calvert .Cliffs 1 and 2, Fort Calhoun 1, Hillstone II, Maine-Yankee and previous reload cycles for St. Lucie-l. The operating experience has been successful except for the CEA guide tube wear prohlen which has been addressed in Section 4.2.

Tamil 4-i ~ Predicted gad Col)apse Tiw> Coivpa~ik to I'rebec".ed Operating Tir.~

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5.0 NUCLEAR DESIGN 5.1 Physics Characteristics 5.1.1 Fuel Management The Cycle 4 fuel management employs a mixed central region as described'n Section 3, Figure 3-1. The fresh Batch F is comprised of two sets of assemblies, each having a unique enrichment in order to minimize radial power peaking. There are 40 assemblies with an enrichment of 3.65 wt/ U-235 and 48 assemblies with an enrichment of 3.03 wt/ U-235 and 12 poison shims per assembly. With this loading, the Cycle 4 burnup capacity for full power l

operation is expected to be between 14,300 MWD/T and 14,900 MWD/T, depending on the final Cycle 3 termination point. The Cycle 4 core characteristics have been examined for Cycle 3 terminations between 7250 and 8250 MWD/T and limiting values established for the safety analyses.'he loading pattern (see Section 3) is applicable to any Cycle 3 termina-tion point between the stated extremes.

Physics characteristics including reactivity coefficients for Cycle 4 are listed in Table 5-1 along with the corresponding values from the reference cycle. Please note that the values of parameters actually employed in safety analyses are different than those displayed in Table 5-1 and are typically chosen to conservatively bound. predicted values with accommodation for appropriate uncertainties and allowances.

Table 5-2 presents a summary of CEA shutdown worths and reactivity allowances for Cycle 4 with a comparison to reference cycle data. Table 5-2 generally characterizes the changes in reactivity that occur during a trip from full power with a corresponding change in core parameters to the zero power state. It is not inte'nded to represent any particular limiting A00 or accident, although the quantity shown as "Shutdown Margin" represents the numerical value of the worth which is applied to the hot zero power steam line break accident. For the analysis of any specific accident or AOO,

conservative or "m limiting" values are used. a result of previously established conservative limits, the scram worths calculated for Cycle 4 are bounded by the values used in the Cycle 3 safety analysis. The power dependent insertion limit (PDIL) curve and CEA group identification are unchanged from the reference cycle (Reference 2). ,Table 5-3 shows the reactivity worths of'arious CEA groups calculated at full power conditions for Cycle 4.

5.1.2 Power Distribution Figures 5-1 through 5-3 illustrate the all rods out (ARO) planar radial power distributions at BOC 4, MOC 4 and EOC 4 that are characteristic of the high burnup end of the Cycle 3 shutdown window. These .planar radial power peaks are characteristic of the .major portion of the active core length between about 20 and 80 percent of the fuel height.

Figure 5-4 illustrates the planar radial power distribution within the uooer 15 to 20 oercent of the core produced with the insertion of the first CPA regulating group, Bank 7. This power distribution characteristic of near BUC 4 is basea upon the low burnup end of the Cycle 3 shutdown window, providing an illustration of maximum power peaking expected for this configuration. Higher burnup Cycle 3 shutdown points tend to reduce power peaking in this upper region of the core with Bank 7 inserted. It is a characteristic of both ARO and Bank 7 inserted conditions that the Cycle 4 peaks are highest at BOC.

The radial power distributions described in thi s section are calculated data without uncer tai ntes or other allowances. However, single rod power peaki ng values do include the increased peaking that is characteristic of fuel rods adjoining the water holes in the fuel assembly lattice. For both DNB and kw/ft safety and setpoint analyses in either rodded or unrodded configurations, the power peaking values actually used are higher than those expected to occur at any time during Cycle 4. These conservative values, which are used in Section 7 of this document, . establish the allowable limits for power peaking to be observed during operation.

The range of allowable axial peaking is defined by the limiting conditions for operation of the axial shape index (ASI). Mithin these ASI limits, the necessary DNBR and kw/ft margins are maintained for a wide range of possible axial shapes. The maximum three-dimensional or total peaking factor anticipated in Cycle 4 during normal base load, all rods out operation at full power is 1.85 not including uncertainty allowances and augmentation factors. This is well within the operating limits established for Cycle 3.

5. 1.3 Safety Related Data 5.1.3.1 Ejected CEA The maximum reactivity worths and planar radial power peaks associated with an ejected CEA event are shown in Table 5-4 for both BOC and EOC.

These values- encompass the worst conditions. anticipated during Cycle 4 for the planned range of Cycle 3 termination. points and are bounded by the values used in the safety analysis for the reference cycle.

5.1.3.2 Dropped CEA The limiting parameters of dropped CEA reactivity worth and maximum increase in radial peaking factor have been calculated for Cycle 4.

The results indicate that the values uSed in the Cycle 3 analysis are still bounding. A comparison of these parameters for Cycles 3 and 4 is found in Table 5-5.

5.l.4 Augmentation Factors Augmentation factors have been calculated for the Cycle 4 core k

using the calculational model described in Reference 5. The input information required for the calculation of augmentation factors that is specific to the core under consideration includes the fuel densification characteristics, the radial pin power distribution and the single gap peaking factors. Augmentation factors for the Cycle 4 core have been conservatively calculated by combining for input the largest single gap peaking factors (calculated near end of cycle) with the most conservative (flattest) radial pin power distribution. The calculations yield non-collapsed clad augmentation factors showing a maximum value of l. 048 at the-top of the core. As shown in Table 5-6, the augmentation factors for Cycle 3 are more limiting than the values calculated for Cycle 4. The Cycle 3 results were used for this cycle.

8

hy 5.2 PHYSICS ANALYSIS tlETHODS 5.2. l -Uncertainties in treasured Power Distributions The power distribution measurement uncertainties which are applied to Cycle.4 are:

Fq 7.0 percent

, where Fq = Fxy' Fz, local power density Fr = 6.0 percent.

These values are to be used for monitoring power distribution parameters during operation.

5.2.P. Nuclear Design t'jethodology

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The analyses have been performed in the same manner and with the same methodologies used for the reference cycle analyses.

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TABLE 5-1 St. Lucie Unit 1 Cycle 4 Physics Characteristics Refer ence Units ~Cele 4 Dissolved Boron ~Cc1 e Dissolved Boron Content for Criticalit , CEAs Mithdrawn Hot full power, equilibrium PPH '50 1077 xenon, BOC Boron i<orth Hot Full Power BOC PPN/%ap 90 104 Hot Full Power EOC PPH/%ap 80 83 Reactivity Coefficients CEAs Mithdrawn Moderator Temperature Coeffi-cients, Hot Full Power Beginning of Cycle (Equi librium Xe) 10-4 ap/'F -0.2 0.0 End of Cycle 10-4 ap/'F -1.8 -1. 9 Do ler Coefficient Hot BOC Zero Power 10 5 ap/'F -1. 44 -1 . 64 Hot BOC Full'ower 10 5 l4p/'F -1 .1.3 -1. 26 Hot EOC Full Power 10 5 ap/'F 1 ~ 22 -1. 39 Total Delayed Neutron Fraction, geff Beginning of Cycle . 0060 .0063 End of Cycle . 0051 .0051 Neutron Generation Time, a*

BOC 10-6 sec 28 24 EOC 10-6 sec 33 29

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TABLE 5-2 St. Lucie Unit 1 Limiting Values of Cycle 4 CEA REACTIVITY VORTHS AND ALLOWANCES,

/.dp BOC EOC Reference Cycle Reload Cycle Peference Cycle Reload Cycl ti'orth Availabl e*

Worth of all CEAs inserted 10.5 9.7 11. 3 Stuck CEA allowance 2 ' 2.4 3.1 2.9 Worth of all CEAs less, highest worth 7.8 7.3 8.3 8,4 CEA stuck out i<orth Re uired Allowances)

Power defect, HFP to HZP {Doppler, Tavg, 1.7 1.9 2.2 2.5 redistribution)

Hoderator voids 0.0 0.0 0.1 0.1 CEA bite, boron deadband and maneuvering 0.6 0.5 0.6 0.6 band Required shutdown margin (Xdp) 3.3 3~3 .3. 3 3' Total reactivity required 5.6 5.7 6.2 6:5 Available i<orth Less Allowances Mar gin ava-ilable 2.2 . 1.6 2.1 1 ~ 9 For every accident or A00 considered in the safety analysis, a calculational uncertainty of 10Ã is

. deducted from the worth available..

TABLE 5-3 ST. LUCIE UNIT I CYCLE 4 REACTIYITY k'ORTH OF CEA REGULATING GROUPS AT HOT FULL POHER,

%%dDP Regulating CEAs Beginning of Cycle End of Cycle Group 7 0.57 0.80 Group 6 0 '1 0. 60 Group 5 0.32 0. 44 Note Yalues shown assume sequential group insertion.

TABLE 5-4 ST. LUCIE UNIT I CYCLE 4 CEA EJECTION DATA Limiting. Value Reference Cycl e Cycle 4 Haximum Radial Power Peak Safet Anal sis Value Calculated Value Full power with Bank 7 inserted; worst CEA ejected 3. 60 3. 02 Lero power with Banks 7+6+5 ins'erted; worst CEA ejected 8.34 6.61 Maximum E'ected CEA Worth Khp)

Full power with Bank 7 inserted; worst CEA ejected .29 .20 Zero power with Banks 746+5 inserted; worst CEA ejected .65 .50 Notes: Uncertainties and allowances are included in the above data.

Reference cycle results were those used in transient analysis.

TABLE 5-5 St. Lucie-1 Cycle 4 Full Length CEA Drop Data Limiting Values Reference C cle ~Cele 4 Minimum Worth %lNp .04 .10 Maximum Percent Increase in Radial Peaking Factor 17 Notes: (1) Ho uncertainties are included in above data.

(2) CEAs are either fully withdrawn or fully inserted for radial calculations.

(3) Reference cycle results were those used in transient analysis.

TA8LE 5-6 St. Lucie Unit 1 Augmentation Factors and Gap Sizes for Cycle 4 and Reference Cycle Reference C cle Reload C cle Core Core Noncollapsed Gap Noncollapsed Gap Height Height Clad Augmen- Size Clad Augmen- Size

~Percent) ~Inches) tation Factor ~inches) tation Factor ~Inches)

98. 5 134,7 1 . 058 2.04 1. 048 1.74

86. 8 118.6 1. 053 1.80 1 . 044 1.54 77.9 106.5 1 . 050 1.62 1. 041 1.38

66. 2 90. 5 1 .044 1. 38 1. 036 1.18 54.4 74.4 1. 038 1.14 1, 031 0.97 45,6 62. 3 1.033 0.96 1. 027 0.82 33.8 46,2 1. 026 0.72 1. 021 0. 62 22.1 30. 2 1. 018 0.48 1.015 0.41

13. 2 18.1 1. 013 0.30 1.010 0. 26 1.5 2.0 1. 003 0.06 1. 001 0. 05 Note: Values are based on approved model described in Reference 5.

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6. THERiLAL-HYDRAULIC DESI Gll 6.1 Dt(BR Analyses Steady state D[BR analyses of Cycle 4 at the rated power level of 2560 tie(t have been performed using the same design codes as described in the FSAR, Reference 6. Appropriate adjustments were made to the input of these codes to reflect the Cycle 4 power distribution.

Table 6-1 contains a list of pertinent thermal-hydraulic design

'arameters used for both safety analyses and for generating reactor protective system setpoint information.

The analyses were performed in the same manner as for the reference cycl e.

Investigations have been made to ascertain the effect of the CEA guide tube wear problem and the sleeving repair on D:IOR margins as by this type of analysis. The findings were reported 'stablished to the llRC in Ref rence 4 which conclude that the wear problem and the sleeving repair do not adversely affect DtSR margin.

6.2 Effects of Fuel Rod Bowing on Di(BR tlargin Effects of fuel rod bowing on D'<DR rergin have been incorporated in the safety .and setpoint analyses in the same manner as discussed in Reference p'. This reference contains penalties on minimum D/SR due to. fuel rod bowing as a function of burnup generated using f(RC guidelines contained in Reference 8 .

4 Reference General Characteristics Unit ~Cele 3 ~Cele 4 Total Heat Output(core only) t'lg 2560 2560 10 BTU/hr 8737 8737 .

Fraction of Heat Generated in .975 , .975 Fuel Rod Primary System Pressure Nominal PSIA 2250 2250 Minimum in steady state PSIA 2200 2200 Maximum in steady state PSIA 2300 2300 Design Inlet Temperature 'F 544 544

, Total Reactor Coolant Flow GPN 370,000 370,000 (minimum steady state) 1061b/hr 140.2* 140.2*

Coolant Flow Through Core 1061b/hr 135.0* 135.0*

Hydraulic Diameter ft 0.044 0.044 (nominal channel)

Average Mass Velocity 106lb/hr-ft2 2.53* 2.53*

Pressure Drop Across Core PSI 10.3 10.3 (minimum steady state flow irreversible aP over ent',re fuel assembly)

Total Pressure Drop Across Yessel PSI 33. 5 33.5 (based on nominal dimensions and minimum steady state flow)

Core Average Heat Flux (accounts for BTU/h'r-ft2,'F 174,400 174,310.

above fraction of heat generated in fuel rod and axial densification factor)

Total Heat Transfer Area (accounts for 48,860 48,872 axial densification factor)

Film Coefficient at Average Conditions BTU/kr-ft2 'F 5820 5820 Maximum Clad Surface Temperature oF 657 657 Average Film Temperature Difference oF 31 31 Average Linear Heat Rate of Undensified kw/ft 5.83 5.82 Fuel Rod (accounts for above fraction of heat generated in fuel rod)

.Average Core Enthalpy Rise BTU/lb 65* 65*

• Calculated at design inlet temperature, nominal primary system pressure.

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TABLE 6-1 (continued)

Reference Calculational Factors ~C'cle 3 ~Cele 4 Engineering Heat Flux Factor 1.03

, Engineering Factor on Hot Channel Heat Input 1.03 1.03 Inlet Plenum Nonuniform Distribution 1.05 1.05 Rod Pitch, Bowing and Clad Diameter 1.065 1.065 Fuel Densification Factor (axial) 1.01 1.01 Fuel Rod Bowing Augmentation Factor on Fr 1.018 1.018 Statistical Component of Fr 9 95/95 Confidence Level 1.06 .1.06

7.0 ACCIDENT T AND TRAtiS NT ANALYSIS OTHER THAr< LOCA The purpose of this section is to present the results of the safety analysis (other than LOCA) for St. Lucie Unit 1, Cycle 4 at 2560 t~iHT containing fuel assemblies with 3.65 w/o enrichment. The events considered for this analysis are listed in Table 7.1. These are the design basis events for the plant.

These events can be categorized into the following groups:

l. Anticipated Operational Occurrences for which the Reactor Protection System prevents the Specified Acceptable Fuel Design Limits (SAFDLs) from being exceeded;

2. Anticipated Operational Occu) rences for which the initial steady state overpower margin must be maintained in order to prevent the SAFDLs from being exceeded;

3. Postulated Accidents.

Each of the events listed in Table 7-1 has been reviewed for Cycle 4 to determine if an explicit reanalysis was required'. Table 7-1 indicates the analysis status of each event. Table 7-2 presents the core parameters used.

in the Cycle 4 analysis and compares them to the reference cycle. The review of each design basis event (DBE) entailed a comparison between all the current and reference cycle key transient paramet'ers that significantly impact the results of an event. The reference analysis for each event is the analysis upon which'he licensing of St. Lucie Unit 1, Cycle 3 was based. llhen the current cycle values of key parameters for a particular event are bounded by (conservative with respect to) the reference cycle, no reanalysis is required or performed.

The results of the review are that the key parameters for all the DBEs for Cycle 4 operation are the same as, or no worse than, the specified reference cycle input parameters, except for the following:

1. Higher critical boron concentration

2. Seized Rotor pin census

3. CEA Ejection pin census A reanalysis"of the Boron Dilution event was performed to determine the effects of the more adverse boron parameters for Cycle 4. The seized rotor event and CEA ejection event were reanalyzed to evaluate'the, impact of more adverse pin. census for, these postulated events.

For all DBEs other than those reanalyzed, 'the St. Lucie Unit 1 safety analyses-for previous relopd cycle license submittals bouncL the results that would b'e obtained for Unit" 1, Cycle 4 and demonstrate safe operation of St. Lucie Unit 1 Cycle 4 at 2560 llWT with the higher enrichment.

fuel.

In summary, the results of the reanalysis demonstrate that the conclusions reached in the reference cycle analysis for each event remain valid for Cycle 4.

TABLE 7-1 St. Lucie Unit 1, Cycle 4 Events Considered in Transient and Accident Analysis I'll i S<<

Anticipated Operational Occurrences for which the RPS Assures no Violation of SAFDLs:

Control Element Assembly Withdrawal '-

Not Reanalyzed Boron Dilution Reanalyzed St~tr )p of an Inactive Reactor Coolant Pump Not Reanalyzed Excess Load Hot Reanalyzed Loss of Load Not Reanalyzed Loss of Feedwater Flow Hot Reanalyzed Excess Heat Removal due to Feedwater tlalfunction Not Reanalyzed Rea'ctor Coolant System Depressurization Not Reanalyzed Loss of Coolant Flow Hot Reanalyzed Loss of AC Power Not Reanalyzed Anticipated Operational Occurrences which are Dependent on Initial Overpower Margin for Protection Against Violation of SAFDLs:

Loss of Coolant Flow Hot Reanalyzed Loss of AC Power Not Reanalyzed Full Length CEA Drop Not Reanalyzed Part Length CEA Drop Hot Reanalyzed Part Length CEA Nalpositioning Not Reanalyzed Transients Resulting from Malfunction of One Not Reanalyzed Steam Generator Postulated Accidents:

CEA Ejection Reanalyzed Steam Line Rupture Not Reanalyzed Steam Generator Tube Rupture Not Reanalyzed Seized Rotor Reanalyzed 1

Requires Low Flow Trip.

TABLE 7-2 St. Lucie 1 Core Parameters Input to Safety Analyses'h Reference Cycle 4 sics Parameters Uni ts ~C1 I 1 V Planar Radial Peaking Factors For DNB Margin Analyses (Fr)

Unrodded Region 1. 59 1. 59 Bank 7 Inserted 1.80 1. 80 For Planar Radial Component of 3-D Peak (Fx )

(kw/ft Limit Analyses)

Unrodded Region 1.58 1.58 Bank 7 Inserted 1.82 1.82 Peak Augmentation Factor 1. 071 1. 071.

Moderator Temperature Coefficient 10 bp/ F -2.5 ~ +.5 -2.5 ~ +.

Shutdown Margin (Yalue used in Zero Power) -4.1 / -3.3 -4.1 / -3.

(SLB) (1 loop/2 loop)

Safet Parameters Power Level NHt 2611 2611 Maximum Steady State Core Inlet Temperature oF '44 544 Minimum Steady State RCS Pressure psia 2200 2200.

Reactor Coolant Core Flow 10 lb/hr 134. 9 134.9 Full Power Axial Shape Index Limit Ip "023 ~ 23 Maximum CEA Insertion at Full Power  % Insertion of Group 7 25 Minimum Allowable Initial Peak Linear Heat Rate for transients other than LOCA kw/ft 16.0 16.0 Steady State Linear Heat Rate to Fuel Centerline kw/ft 21,0 21.0 Melt CEA Drop Time from Removal of Power Holding. Coils to 90% Insertion Sec 3.1 3.1 Three Pump Plenum Factor 1.09 1.09

J TABLE 7.1-1 Assumed Input Parameters for Boron Dilution Analysis Ref. Cycle Parameter ~21 2 ~Cele 4 Critical Boron Concentration, PPt1 (All Rods Out, Zero Xenon)

Power Operation 1200 1330 Startup 1300 1420 Hot Standby 1300 1420 Hot Shutdown 1300 1420 Cold Shutdown 1300 1420 Refueling 1200 1280 Inverse Boron Worth, PPt</% ap Power Operation 70 95 Startup 65 90 Hot Standby 70 Hot Shutdown 70 Cold Shutdown 55 70 Refuel ing 55 70

7.1 BORON 'DILUTION EVENT The Boron Dilution event has,been reanalyzed for Cycle 4 due to increases in the critical boron concentrations (See Table 7.1-1 for comparison between Cycle 2 and Cycle 4 boron parameters. This is the same reference cycle that was cited in the Cycle 3 license submittal). This increase in critical boron concentration is offset by a corresponding increase in the minimum inverse boron worth. Thus, the time to dilute to criticality for Cycle 4 is no less than the time calculated for the reference cycle.

The Boron Dilution event at power produces a slow power and temperature increase which causes an approach to both the DNBR and kw/ft SAFDLs. Since the Ttl/LP trip system monitors the transient behavior of core power level and core inlet temperature, the Tt1/LP trip assures that the DNBR SAFDL is not exceeded for power increases within the setting of the Variable High Power Level trip; for power excursions in excess of the Variable High Power Level trip, a reactor trip is actuated. The approach to the kb/ft SAFDL is terminated by either the Local Power Density-High trip, Variable High Power Level trip or the DNBR required trip discussed above.

For boron dilution initiated from hot zero power, critical, the power transient resulting from the slow reactivity insertion rate characterizing the boron dilution transient is terminated by the Variable High Power Level trip prior to approaching the SAFDLs. The re-evaluation shows the time to criticality is greater than 15 minutes for boron dilutions initiated from the Startup, Hot Standby, Hot Shutdown, and Cold Shutdown operational modes. For the re-fueling mode, the time to criticality is greater than 30 minutes. Consequently, the conclusions reached for Cycle 2 remain valid for Cycle 4.

7.2 SEIZED ROTOR EVENi The Seized Rotor Event was reanalyzed for Cycle 4 to evaluate the 'number of fuel pins predicted to experience DflB due to a slightly more adverse pin census distribution for Cycle 4 than for the reference cycle. (Reference cycle for this event is Cycle 3.)

The transient behavior of this event is the same as for the reference cycle since all'he transient related parameters are the same as, or conservative with respect to, the reference cycle. Therefore, only a recalculation of the number of fuel pins predicted to experience DNB was performed using the cycle 4 pin census.

The results show that, for Cycle 4, the number of fuel pins predicted to experience DNB is 1.05/, as compared to the 0.99~ reported for Cycle 3.

Therefore, the conclusion reached in the reference cycle that only a very small number of the fuel pins would experience DNB'emains valid for Cycle 4.

0 7.3 CEA EJECTION EVENT The CEA Ejection Event, was reanalyzed for Cycle 4 to evaluate the number of pins. predicted to experience incipient centerline 'elt due to a slightly more ad'verse pin census distribution for Cycle 4 than for the reference cycle.

'uel (Reference cycle for this event is Cycle 3.) In the reference cycle, no pin was predicted to exceed the criterion for clad damage (i.e., average deposited energy of 200 cal/gm).

The transient behavior of this event is the .same as for the reference cycle since all the transient related parameters are the same as, or conservative with respect to, the reference cycle. Therefore, onl'y a recalculation of the number, of fuel pins predicted to experience incipient centerline melting was performed using the cycle 4 pin census.

The results show that, for Cycle 4, the predicted fraction of fuel pins expected to experience incipient centerline melting for the transient initiated at full power is 0.045. For the reference cycle analysis, a calculated fractional value of 0.028 of the fuel pins were predicted to expel ience incioi nt centerline melting at full power. However, since no fuel pin is predicted to experience clad damage, the conclusion reached in the reference cycl e remains vali d.

References (Sections I through 7)

1. CEN-79-P, "Reactor Operation With Guide Tube Hear", February 3, 1978

2. Letter, Robert E. Uhrig (FPSL) to Victor Stello (NRC), dated February 22, 1979, "St. Lucie Unit 1 Docket No. 50-335 Proposed Amendment to Facility Operating License-.DPR-67"

3. CENPD-187, "CEPAN method of Analyzing Creep Collapse of Oval Cladding",

June 1975 CEN-80(N)-P; "Millstone Unit 2. Reactor Operation With ttodified CEA Guide Tubes", February 8, 1978

5. CENPD-139, "C-E Fuel Evaluation Hodel Topical Report", July 1, 1974

6. St. Lucie Nuclear Power Plant (Formerly Hutchinson Island) Unit One, Final Safety Analysis Report, in support of Docket No. 50-335 7, Supplement 3-P (Proprietary) to CENPD 225P, "Fuel and Poison Rod Bowing",

June 1979 Letter from D. B. Vassallo (NRC) to A. E. Scherer (C-E) dated June 12, 1978.

St. Lucie I Cycle 4 ECCS Performance Results I

INTRODUCTION ANO SUt<HARY The ECCS perfora>ance evaluation for St. Lucie I Cycle 4, presented herein, appropriate conformance with the Acceptance Criteria for 'emonstrates Light-Water-Cooled Reactors as presented in 10CFR50.46 (1) . The evaluation demonstrates acceptable ECCS performance at a peak linear heat generation rate (PLHGR) of 14.8 kw/ft and a power level of 2611 that (102Ã of 2560 t'lwt).

The method of analy is and results are presented in the following sections.

HETHOD OF ANALYSIS This analysis was performed using the approved C-E Large Break Evaluation tlodel (2) . The model was used to re-evaluate the limiting large break LOCA ECCS performance. The blowdown and refi ll-reflood parameters of the previous cycle remain unchanged. Therefore, only STRIPIN II( )

calculations were necessary to account for the different pin conditions.

Burnup dependent calculations were performed using the FATES (5) and STRIKIN-II codes to determine the limiting condition for the ECCS performance analysis. The break size and type analyzed, 0.8 DES/PD*,

is the same as was analyzed in previous cycles.

For conservatism, the PARCH code was not utilized in the Cycle 4 evaluation ~ The late ref lood heat transfer benefit from the use of the PARCH steam cooling heat transfer would have reduced the peak clad temperature reported herein.

• 0.9 x Double Ended Slot Break in the Reactor Coolant Pump Discharge Leg

O 8.2 RESULTS AND CONCLUSIONS Table 1 presents the analysis results for the limiting 0.8 DES/PD break. A list of the significant parameters displayed graphically is presented in Table 2. A summary of the fuel and system parameters is shown in Table 3.

As can be seen from the results, the worst break analysis results in a peak clad temperature of 1986'F which is well below the ci i teria limit. The local and core wid(

zirconium oxidation percentages are 10.49% and 0.60'i,, respectively. Hence, opera-s tion at a peak linear heat generation rate of 14.8 kw/ft and at a power level of 2611 Hwt (102K of 2560 Hwt) will result in acceptable ECCS performance.

3 COHPUTER CODE VERSION IDEHTIFICATION The following NRC-approved version of Combustion Engineering ECCS Evaluation I'1odel computer code was used in this analysis:

STRIKIH-II: Version No. 77036

REFERENCES (Secti on 8)

l. Acceptance Criteria for Emergency Core Cooling Systems for Light-Mater Cooled Nuclear Power Reactors, Federal Register, Vol. 39, No. 3 - Friday, January 4, 1974.

2. CENPD-132, "Calculative Nethods for the CE Large Break LOCA Evaluation Model", August 1974 (Proprietary).

CENPD-132, Supplement 1, "Calculational Methods for the CE Large Break LOCA Evaluation Model", December 1974 (Proprietary).

CENPD-132, Supplement 2, "Calculational methods for the CE Large Break LOCA Evaluation Model", July 1975 (Proprietary).

3. Letter from FPSL to llRC transmitting St. Lucie I Cycle 3 ECCS performance results (February 22, 1979; L-79-45),

4. CENPD-135, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program", August 1974 (Proprietary).

CENPD-135, Supplement 2, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program (Modifications), February 1975 (Proprietary).

CENPD-135, Supplement 4, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program", August 1976 (Proprietary).

CENPD-135, Supplement 5, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program", April 1977 (Proprietary).

5. CENPD-139, "CE Fuel Evaluation l1odel", July 1974 (Proprietary).

6. CENPD-138, "PARCH A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup", August 1974 (Proprietary).

CENPD-138, Supplement 2, "PARCH - A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant fleatup", January 1977 (Proprietary).

TABLE 1 Summary of Results for St. Lvcie I Cycle 4 ECCS Performance Results Oxidation /

Break ~Tt Peak Clad

~tt Time T

of Time of Clad Ru ture Local Core Bi-de 0.8 DES/PD 1986'F 249.Z sec 55.32 sec 10.49- (. 60.

St. Lucie I Cycle 4 Variables Plotted as a Function of Time Variables ~Fi ure D~esi nation Peak Clad Temperature ~

1 Hot Spot Gap Conductance Peak Local Clad. Oxidation 3 Clad Temperature, Centerline Fuel Temperature, Average Fuel Temperature and Coolant Temperature for Hottest Node Hot Spot Heat Transfer Coefficient Hot Rod Internal Gas Pressure

Va TABLE 3 St. Lucie I Cycle 4 General System Parameters Q~uanti t Value Reactor Power Level (102% of Nominal) 2611 Hwt Average Linear Heat Generation Rate (102K of Nominal) 6.0932 kw/ft Peak Linear Heat Generation Rate 14.8 kw/ft Gap Conductance at PLHGR 1527 BTU/hr ft 'F Fuel Centerline Temperature at PLHGR 3510.3 oF Fuel Average Temperature at PLHGR 2195.6 oF Hot Rod Gas Pressure 1035.8 psia Hot Rod Burnup 1488 t@lD/t )TU

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