ML18227D145

Submits Description of Fuel Which Will Be Reloaded Into Unit 4, Cycle 3 Operation. Proposed Technical Specifications Relating to Refueling Are Being Submitted Under Separate Cover Letter

ML18227D145
Person / Time
Site:	Turkey Point
Issue date:	02/25/1976
From:	Robert E. Uhrig Florida Power & Light Co
To:	Stello V Office of Nuclear Reactor Regulation
References
L-76-68
Download: ML18227D145 (44)

Text

NRC'FORM 1S5 Q-'761 NRC DISTRIBUTION FDR PART 50 DOCKET MATERIAL DOCKET NUMBER go. ps(

FILE NUMBER TO:

// MR VICTOR STELLO FROID:

FLORIDA POWER 6 LIGHT CO MIAMI, FLA R E UHRIG DATE OF DOCUMENT S-DATE RECEIVED 3-1-74, Q LETTER

~ORIGINAL OCOPV DESCRIPTION Q NOTO R 12 E D

~ UNC LASS I FI ED PROP INPUT FORM ENCLOSURE NUMBER OF COPIES RECEIVED 8 iSflKi3 l/

P LTR TRANS THE FOLLOWING......

NOTE:

TECH SPEC RELATING TO THE REFUELING ARE BEING SUBMITTED UNDER SEPARATE COVER LTR.

RELOAD SAFETY EVALUATION TURKEY POINT PLANT UNIT 4, CYCLE 83m OPERATION......

TURKEY POINT N3 SAFETY ASSIGNED AD:

BRANCH CHIEF.

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RELOAD SAFETY EVALUATION TURKEY POINT PLANT UNIT 4, CYCLE 3 FLORIDA PONER AND LIGHT COMPANY

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TABLE OF CONTENTS Title

~PR e

1.0 XNTRODUCTION AND

SUMMARY

2.0 REACTOR DESIGN 2.1 Mechanical Design 2.2 Nuclear Design 2.3 Thermal and Hydraulic Design 3.0 POWER CAPABXLXTY AND ACCIDENT EVALUATION 3.1 Power Capability.

3.2 Accident Evaluation 3.3 Xncidents Reanalyzed 4.e REFERENCES

'f

LIST OF TABLES Table Title' I

2

'3 Fuel Assembly Design Parameters Kinetics Characteristics Shutdown Requirements and Margins Results of Rod Ejection Analysis

'E

LIST OF FIGURES Title Loading Pattern Source and Burnable Poison Locations Control Group Insertion Limits for Unit 4 Three Loop Operation Control Group Insertion Limits for Unit 4 Two Loop Operation

t

1. 0 INTRODUCTION AND

SUMMARY

Turkey Point Unit 4 achieved initial criticality in June, 1973 and is now in its second cycle of operation.

The unit is scheduled for a refueling shutdown on April 1'9, 1976 with Cycle 3 startup planned for late May.

The Turkey Point 4 Cycle 3 core loading pattern is shown as Figure l.

Forty of the Region 2 assemblies will be removed and replaced by forty Region 5 assemblies (see Table 1).

The five Region 1 assemblies used in the core during Cycle 2 will be replaced by five Region 1 assemblies stored in the spent fuel pit during Cycle 2.

Depleted borosilicate burnable poison rods will be used in Cycle 3.

The location of these rods is shown in Figure 2.

This report presents an evaluation for Cycle 3 which demonstrates that the core reload will not adversely affect the safety of the plant; It is not the purpose of this report to present a reanalysis of all potential incidents.

Those incidents analyzed and reported in the FSAR which could potentially be affected by fuel reload have been reviewed for the Cycle 3 design described herein.

The results of new analyses have been included and the justification for the applicability of previous results from the remaining analyses is presented.

It has been concluded that the Cycle 3 design does not cause the previously acceptable safety limits for any incident to be exceeded.

This conclusion is based on the 'assumption th'at:

(1) Cycle 2 operation is terminated after 8700

+ 1000.MWD/MTU and (2) there is adherence to plant operating limitations as clarified ip proposed modifications to the Technical Specifications.

Nominal design parameters for Cycle 3 are 2200 Mwt rated

power, 2250 psia system pressure, 546'F core inlet temper-

ature, and 5.56 kw/ft average linear fuel power density (based on 144" active fuel length).

1 l

',2.0 REACTOR DESIGN l

2.1 Mechanical Design The mechanical design of Region 5 fuel is dimensionally the same as Region 4 fuel.

Region 5 fuel has different enrichment as noted in Table l.

Other physical design aspects of Region 5 are the same as Region 4, except that the initial prepressurization level of the'fuel rods has been decreased by 65 psi.

Clad flattening time is predicted to be >30,000 EFPH for the limiting region (Region

3) using the current Westinghouse evaluation model Therefore, Region (1) 3 has a nominal Cycle 3 allowed residence time of i2,000 ~EFFH (assumes a Cycle 2 lifetime of 6700

'EFPH and a Cycle 1 lifetime of l0,300 IEFPH).

Clad '.

flattening will not occur during Cycle 3 since this e

I'ycle will nominally operate for 6700 EFPH.

Westinghouse has had considerabl'e experience with Rircaloy clad fuel.

This experienc is extensively described in WCAP 8183, "Operational Experience with Westinghouse Core's" This report is updated about every six months.

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2.2 Nuclear Design The Cycle 3 loading pa+em. results in a maximum F of 2.32 under normal o crating conditions.

Table 2

provides a comparison f the Cycle 3 core kinetics characteristics

'with the current limit based on pzevicusly submitted kccident analyses.

Xt can be seen from the table that most of the Cycle 3 values fall sgithin the current limits.

These parameters are evaluated in Sect).on 3.0.

Table 3 provides the control rod worths and requirements.

The required shutdown margin is based on a previously submitted ac'cident 'analysis The available shutdown margin exceeds the minimum required.

2.3 Thermal and H draulic Design No significant variations in thermal margins will result from the Cycle 3 reload.

The present DNB core limits have been found to be conservative.

While fuel properties were such that the original ECCS analysis for Turkey Point.

was not, applicable (7) to Turkey Point. Unit, 4 for the early part of Cycle 2, the Turkey Point Unit 4 fuel has had sufficient.

exposure (see FPL letter L-75-380, July 29, 1975) (8) so that the basic ECCS analysis again represents a

conservative calculation for Cycle 3.

An ECCS anal-ysis which takes into account rod bowing was sub-mitted to the Regulatory Staff on November 26, 1975 (9) as part, of the Turkey Point. Unit 3 reload submittal.

That analysis showed that Chere is sufficient margin to operate Unit 3 safely under. the present Technical Specifications.

As the stored energy in the fuel for Cycle 3 of Turkey Point Unit 4 is no greater than that of the fuel in Unit 3, and the propensity for rod bowing is no greater, the ECCS analysis of No-vember 26, 1975 is also conservative with respect to Unit 4 and no changes in the present Technical Specification core power limits are required.

l

Likewise, as part of the Unit 3, Cycle 3 reload submittal, a discussion of the rod bow effect on DNB margin was presented on November 21, 1975 (10)

It was concluded that for Turkey Point 3 the various design margins in the calculational model exceeded the DNB penalty for rod bow effects and that therefore no additional restrictions are required.

Similar reasoning is applicable to Turkey Point

Unit, 4 because it has the same type of fuel assemblies and operates with the same power density as Unit 3.

Therefore, it is concluded that the present DNB core limits are conservative.

~

~

3.0 POWER CAPABILITY AND ACCXDENT EVALUATION 3.1 Power Ca ability This section reviews the plant power capab'lity con-sidering the consequences of those incidents examineQ in the FSAR using the previously accepted design. bases.

It is concluded that the core reload will not aQversely affect the ability to safely operate at 100% rated power during Cycle 3.

For the overpower transient, a maximum local rod power limit of 21.8 kw/ft corresponds to the fuel centerline temperature limit of 4700'P for Region 3 fuel.

This can be -accomodated with margin in the Cycle 3 core.

The time dependent densification model was used for this. evaluation.

The LOCA limit is met by main-taining P at or below 2.32.

3.2 Accident Evaluation The effects of the reloaQ on the design basis and postulated incidents analyzed in the PSAR hav'e been examined.

In most cases,'t was found that the effects can beaccommodated within the conservatism of the initial assumptions used in the previous applicable safety analysis.

Por that incident which was reanalyzed, it was determined that the applicable h

design basis limits are not exceeded, and therefore

/

the conclusions presented in the FSAR are still valid.

This reload can typically affect. accident analyses input parameters in three major areas'-:

kinetics characteristics, control rod yqqpths, and 'core peaking ter'actors.

Cycle 3 parameters i'ach of these three r'reas were examined as discussed 'below to ascertain whether new accident analyses are required.

'I j>

~

Kinetics Parameters.

A comparison of Cycle 3 kinetics parameters with current limits is given in Table 2.

Most. of the Cycle 3 coefficients remain within the bounds of current limits.

The small changes in core physics parameters have a negligible effect on transient analysis.

Therefore, no additional accident analysis is required due to changes in these parameters.

Control Rod Worths Changes in control rod worths may affect shutdown margin, differential rod worths, and ejected rod worths.

Table 3 shows that Cycle 3

shutdown margin is adequate.

Table 2 shows that the reactivity insertion rate due to control rod withdrawal is not greater than was previously analyzed.

Cycle 3

ejected rod worths are less than the current limits.

Core Peakin Factors Evaluation of 'peaking factors for the rod out of position and dropped RCCA incidents shows that DNBR is maintained above l.

For t..e dropped bank incident, the turbine runback setpoint is sufficient to prevent a

DNBR less than 1.3.

Peaking factors following control rod ejection were less for Cycle 3 than the. current limits 3.3 Xncidents Regnal zed The end of cycle full power rod ejection inciden" was re-evaluated since the average fuel.temperature conservatively assumed at the initial hot spot linear power density exceeded that previously used in this incident* by approximately 260 F.

The effect of the initially higher fuel temperature is to increase the peak transient fuel and clad temperature following rod ejection, but. by less than 260'F.

Since

• This fuel temperature was used only for the end of life ejected x'od analysis.

Fuel temperatures previously used in other incidents are unaffected.

l

the peak fuel and clad temperatures obtained for the previous analysis were well below the limiting criteria

, the results with the 260 I'ncrease (shown in Table

4) still do not cause the criteria to be exceeded.

The limiting case for the rod ejection incident actually occurs at the beginning of life for which the parameters are as described in the previous analysis (S)

4.0.

REFERENCES

George, R. A., et al "Revised Clad Flattening Model",

RCAP 8377 (Propriet.ary) and NCAP 8381 (Non Proprietary),

July 1974.

2.

- Hellman, J.

M. (Ed.), "Fuel Densification Experimental Results and Model for Reactor Operation",

NCAP 8218-P-A, March 1975 (Proprietary) and NCAP 8219-A, March 1975 (Non Proprietary).

3.

. Plocido, V. J.

and Schreiber, R. E., "Operational.

Experience with Nestinghouse Cores",

NCAP 8183, Revision 3, -May 1975.

4.

Final Safety Analysis Report.,

Turkey Point Units Number 3 and 4.

5.

"Fuel Densification, Turkey Point Plant Unit Number 3",

NCAP 8074 '(Proprietary) and NCAP 8075 (Non Proprietary),

February 1973.

6.

Turkey Point Plant Unit 3, Docket. Number 50-250; Cycle 3 Reload Fuel Submittal and Proposed Amendment to Facility Operating License DPR-31, September 9,

1975 (Attachment to Letter from Robert E. Uhrig to Roger S. Boyd).

Turkey Point Plant Units 3 and 4, Docket Numbers 50-250 and 50-251, FSAR ECCS Final Acceptance Criter'a, March 10, 1975 (attachment to letter from Robert E.

Uhrig to George Lear).

8.'urkey Point Plant Units 3 and 4, Docket Numbers 50-250 and 50-251, Proposed Amendment to Facility Operating Licenses DPR-31 and DPR-41, July 29, 197,5 (attachment to letter from Robert E. Uhrig to Angelo Giambusso)

9.

0 Turkey Point Plant Unit 3, Docket Number 50-250, Cycle 3 Reload Fuel Submittal Add'tional Information, November 26,

1975, (Letter from Robert E. Uhrig to Roger S. Boyd).

10.

Turkey Point Plant Unit 3, Docket Number 50-250, Cycle 3 Reload Fuel Submittal Additional Information, Nov'ember 21, 1975 (Letter from Robert E. Uhrig to Roger S. Boyd).

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Table 1

Turkey Point 4 - Cy.cle 3

1

Fuel Assembly Design 'Parameters ReqloA 2

3 4

5 Enrichment (w/o U-235),

Density (5 Theoretical)+

l.85

2. 56 3.11

2. 55 3- 00

93. 8

92. 9

92. 8

94. 3

95. 0 Number of Assemblies Approximate Burnup at

Beg.inning of Cycle 3

(f"'U/f'TU) l2 48 52

~

40 14700 23900 21400 7400 0

~ All regions except Region 5 are as-built values; Region 5 is the nominal

value, however, an average density of 94.5'heoretica1 was'sed in thermal evaluations.

~

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Table 2

Turkey Point Unit 4 Kinetics Characteristics Hoderator Temperature Coefficient,.(hp/'F)xl0

. Current Limit! 'Cycle 2

-3.5 to +0.3*

-3.5 to 0

~Cele 3

-3.5 to 0 Doppler Coefficient (hp/'F )xl0 Delayed Neutron Fraction, jeff (')

-1.6 to -1.0

.50. to.72

.50 to.59.

.50 to.'72

-2.6 to.-1.0 -2.6.to -1.0

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Prompt heutron Lifetime

(>sec) 0 Hax'imum Differential Rod cnorth of T!Io Banks lloving Together at HZP (pcm/in)**

14 to 18

.80.

20 80 20 (max. )

80

• The positive coefficient does not occur at operating conditions

• • pcm = 10 hp

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Table 3

'.- Turkey Point 4 - Cyc1e 2 and 3

'Shutdown Requirements and Hargins Control Rod worth Kh Cycle 2

'OC EOC Cycle 3 BOC EOC All Rods Inserted. Less horst Stuck Rod

{1) Less 10Ã

6. 97

6. 27

6. 52

5. 86 5.87

5. 27

5. 84

5. 25

. Control Rod Re uirements i.'ho Reactivity Defects (Doppler, Tavg,

Yoid, R distributior.)

l. 66 2.70 '.73

2. 7l Rod Insertion Allowance

. (2) Total Requirements

.70

2. 36

.70

3. 40

.50

2. 23

50

3. 21 Shutdown fear in 1)-

2

'AAo 3.9l ',47 3.04 Re uired Shutdown f1aroin

Ãho 1'. 00

l. 77

l. 00

l. 77

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Table. 4 RESULTS OF ROD EJECTION AlNLYSIS.

MAXXMUM.FUEL AND CLAD TEMPERATURES AT THE HOT SPOT

'ND OF LIFE, HOT FULL POtlER Previous Anal sos. (5)

~Cele 3

Fuel Average Temperature(

F)

Fuel Centerl ine Temperature('F)

Clad Average Temperature('F) 31 80 4275 1835 3440 4535 2095

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