ML17301A108

Application for Amend to License NPF-16,revising Tech Specs to Reflect Changes Required to Commence Cycle 2 Operation. Affidavit,Reload Safety Evaluation Summary & Reload Safety Rept Encl.App I to Rept Withheld (Ref 10CFR2.790)

ML17301A108
Person / Time
Site:	Saint Lucie
Issue date:	06/04/1984
From:	Williams J FLORIDA POWER & LIGHT CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
Shared Package
ML17215A430	List: ML17215A429 ML17215A431 ML17301A108
References
L-84-148, NUDOCS 8406110288
Download: ML17301A108 (40)

Text

REGULATO INFORMATION DISTRIBUTION TEM (RIDS)

ACCESSION NBR)8406110288 OUTDATE'4/06/04 NOTARIZED: YES DOCKET 389 St, Lucie Plant~ Unit 2< Florida Power 8 Light Co. 'ACILs50 05000389 AUTH,NAME AUTHOR AFFILIATION WILLIAMS~J>>H, Fl or ida Power 8 Light Co, REC IP NAME

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RKC IP IKNT AFFILIATION EISENHUT~D>>G ~ Division of Licensing

SUBJECT:

Application for amend to License NPF 16grevising operations Tech Specs to reflect changes required to commence Cycle- 2 Affidavitrreload safety evaluation summa y 8 "Reload Safety Rept" encl,App I to rept withheld (ref 10CFR2" 790) ~

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~glib'LORIDA POWER & LIGHT COMPANY June 4, 1984 L-84-148 Office of Nuclear Reactor Regulation Attention: Mr. Darrell G. Eisenhut, Director Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Eisenhut:

Re: St. Lucie Unit No. 2 Docket No. 50-389 Proposed License Amendment

~C*I 2RI 2 ln accordance with IO CFR 50.90, Florida Power & Light Company submits herewith three signed originals and forty copies of a request to amend Appendix A of Facility Operating License NPF-I6.

This amendment is submitted to reflect changes required to commence operation of Cycle 2, which is currently scheduled for November 25, l984.

Therefore, NRC approval is requested on or before November 25, l 984.

The proposed changes are summarized in the attached St. Lucie Unit 2 Cycle 2 Reload Safety Evaluation Summary, and are shown on the accompanying marked-up Technical Specification pages. A detailed Reload Safety Report is attached.

It should be noted that the proposed changes permit operation of St. Lucie Unit

. <NA 2 gyes Cycle 2 at the licensed power level of 2560 MWt. However, the analyses OPJO incorporate and bound operation for core power levels up to 2700 MWt.

OO Authorization for operation up to 2700 MWt. will be requested in a future Ieo license amendment application.

I'

,2 IK No In accordance with IO CFR 50.9I(a)(l), it has been determined that the proposed OQ amendment does not involve any significant hazards considerations pursuant to

'Cf 2 rig IO CFR 50.92. The No Significant Hazards Considerations determination is

=-

attached.

,'O.'OA RL,fL fn accordance with IO CFR 50.9I(b)(l), a copy of the proposed amendment is being forwarded to the State Designee for the State of Florida.

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22 yI(S PEOPLE... SERVING PEOPLE

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-'Page 2 .

Office of Nuclear Reactor Regulation Mr. Darrell G. Eisenhut, Director Division of Licensing Appendix I to the attached 'Reload Safety Evaluation Report is proprietary information, and therefore, exempt from public disclosure in accordance with IO CFR 2.790.

The proposed amendment has been reviewed by the St. Lucie Facility Review Group and the Florida Power & Light Company Nuclear Review Board.

The proposed amendment has been determined to be a Class IV amendment. A check for $ I2,300.00 is-attached in accordance with IO CFR l70.22.

Very truly yours, J. W. Williams, Jr.

Group Vice President Nuclear Energy, JWW/R JS/db Attachment cc: J. P. O'Reilly Regional Administrator, Region II U.S. Nuclear Regulatory Commission IOI Marietta Street, N.W., Suite 2900 Atlanta, GA 30303 Lyle Jerrett, Ph.D., Director Office of Radiation Control Dept. Health 8 Rehabilitative Services I 3 I 7 Winewood Boulevard Tallahassee, FL 3230 I

AFFIDAYIT PURSUANT TO 10 CFR 2.790 Combustion Engineering, Inc.

State of Connecticut County of Hartford I, A. E. Scherer, depose and say that I am the Director, Nuclear Licensing, of Combustion Engineering, Inc., duly authorized to make this affidavit, and have revi ewed or caused to have reviewed the information which is identified as proprietary and referenced in the paragraph immediately below. I am submitting this affidavit in conformance with the provisions of 10 CFR 2.790 of the Commission's regulations and in conjunction with the application of Florida=-

Power and Light Company for withholding this information.

The information for which proprietary treatment is sought is contained in the following document:

Statistical Combination of Uncertainties - FPP~L Unit 2, Cycle 2 Reload Report Appendix I .

r This document has been appropriately designated as proprietary.

I have personal knowledge of the criteria and procedures utilized by Combustion Engineering in designating information as a trade secret, privileged or as confidential commercial or financial information.*

~

Pursuant to the provisions of paragraph (b) (4) of Section 2.790 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure, included in the above referenced document, should be wi thhel d.

1. The information sought to be withheld from public disclosure are the methodology related to the determination of the probability distributions for specific uncertainties and the combination of uncertainties to be used in determining plant setpoints and related technical. specifications, which is owned and has been held in confidence by Combustion Engineering.

2. The information consists of test data or other similar data concerning a process, method or component, the application of which results in a substantial competitive advantage to Combustion Engineering.

3. The .information is of a type customarily held in confidence by Combustion Engineering and not customarily disclosed to the public. Combustion Engineering has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The details of the aforementioned system were provided to the Nuclear Regulatory Commission via letter OP-537 from F.H. Stern to Frank Schroeder dated December 2, 1974. This system was applied in determining that the subject document herein are proprietary.

4. The information is being transmitted to the Commission in confidence

,under the provisions of 10 CFR 2.790 with the understanding that it is to be received in confidence by the Commission.

5. The information, to the best of my knowledge and belief, is not available .in public sources, and any disclosure to third parties has been made pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence.

6. Public disclosure of the information is likely to cause substantial harm to the competitive position of Combustion Engineering because:

a. A similar product is manufactured and sold by major pressurized water reactor competitors of Combustion Engineering.

b. Development of this information by C-E required tens of thousands of man-hours and hundreds of thousands of dollars. To the best of my knowledge and belief a competitor would have to undergo similar expense in generating equi valent information.

c. In order to acquire such information, a competitor would also require considerable time and inconvenience related to the development of methods to statistically combine uncertainties and determine uncertainty probability distributions for specific uncertainties.

d. The information required significant effort and expense to obtain the licensing approvals necessary for application of the information.

Avoidance of this expense would decrease a competitor's cost in applying the information and narketing the product to which the information is applicable.

e. The information consi'sts of methods and statistical models used to combine uncertainties and the resultant net uncertainty to be applied in determining plant setpoints and technical specifications, the application of which provides a competitive economic advantage. The availability of such information to competitors would enable them to modify their product to better compete with Combustion Engineering, take marketing or other actions to improve their product 's position or impair the position of Combustion Engineering 's product, and avoid developing similar data and analyses in" support of their processes, methods or apparatus.

f. In pricing Combustion Engineering 's products and services, significant research, development, engineering, analytical, manufacturing, licensing, quality assurance and other costs and expenses must be included.

The ability of Combustion Engineering 's competitors to utilize such information

without similar expenditure of resources may enable them to sell at prices reflecting significantly lower costs.

g.- Use of the information by competitors in the international marketplace would increase their ability to market nuclear steam supply systems by reducing the costs associated with their technology development. In addition, disclosure would have an adverse economic impact on Combustion Engineering's potential for obtaining or maintaining foreign licensees.

Further the deponent sayeth not.

A. E. cherer Director Nuclear Licensing Sworn .to before me et%

this /o day of May d'

ary Public

~;DTA'i 'S~il II; YOTAB'i'.UDLIC 1

STATF. OF CO'Hi iP:T(C!3": <>'.c. u~i42

'COiMMiSilON E)li')RES i'3,'(RCh Jl, 19'9

STATE OF FLORIDA COUNTY OF DADE J. W. Willians, Jr., being duly sworn, deposes and says:

That he is a Group Vice President of Florida Power 5 Light Company, the Licensee herein; That he has executed the foregoing document; that the statements made in this document are true and correct to the best of his knowledge, information, and belief, and that he is authorized to execute the document on behalf of said Licensee.

Appendix to the Re oa Safety Report is proprietary, and therefore, exempt from public disclosure in accordance with Section'2.790 of the NRC "Rules of Practice", Title 10, Code of Federal Regulations.

J. W. Willians, Jr.

Subscribed and sworn to before me this 4~~ dey of c/one, 19 SA NOTARY PUB 4C, in and for th County of Dade, St ates of Florida.

h Ny commission expires c/w/ ZP AS

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St. Lucie Unit 2 Cycle 2 Reload Safety Evaluation Summar This report provides a safety evaluation for the operation of St. Lucie Unit 2 Cycle 2 at 2560 MWt. Technical Specifi-cation changes are r'equired to enable operation with 18 month cycles and low,leakage fuel management configurations.

The report provides the necessary analysis to support these Technical Specification changes. The analysis incorporates a change to the CEA (control element assembly) configuration to obtain greater flexibility in operational control and a reduction in the minimum required reactor coolant (RCS) flow to gain sufficient margin between measurable flow and required flow. In addition, the analysis incorporates and bounds operation with a core power level of up to 2700 MWt (although a request to increase the rated core power to 2700 MWt is not included). The required analysis for 2700 MWt also includes a recalculation of containment pressure and temperature during transients. A request for authorization for operation up to 2700 MWt will be submitted in a future license amendment application.

The safety evaluation makes use of the Statistical Combinat'ion of Uncertainties (SCU) methodology in the analysis to provide for a more realistic assessment of system instrumentation uncertainties, system processing uncertainties, manufacturing tolerances and modeling uncertainties. This methodology together with several propose'd Technical Specification changes, which are more restrictive than Cycle 1, provide the extra margin to accommodate more economical fuel management designs, a reduced required minimum RCS flow, and a core power level of up to 2700 MWt without a significant increase in the consequences of potential accidents or any reduction in safety margin (i.e., all required safety criteria are met).

The safety evaluation generally follows the NRC Standard Review Plan (SRP) guidelines in the performance of the safety analyses, with any deviations sufficiently justified.

The analyses do not employ any new or unreviewed methodology (the SCU methodology was previously used and NRC approved for St. Lucie Unit 1, Calvert Cliffs Unit 2 and Arkansas Nuclear One Unit 2).

The proposed Technical Specification changes are summarized in the attached table. These proposed Technical Specification changes and the supporting safety evaluation has been reviewed by the St. Lucie Facility Review Group and the Florida Power Er, Light Company Nuclear Review Board and found to be necessary and meet all the required safety criteria.

" ..8406110288

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Page 1 TABLE 4-1 ST LUCIE UlfIT 2 TEHPiICAL SPECIFICATIOH AND BASES CHAHGES S cification Action Remarks 2-1 2.1.1.2 Change peak linear heat to Peak linear heat to centerline melt centerline melt limit from limit is raised to the calculated 21.0 kw/ft to 22.0 kw/ft limit for Cycle 2, as described in Section 2.2.**

2-3 Figure 2.1-1 Replace this figure with Thermal limit lines are being changed revised figure. to reflect analysis at 2700 MWT, Technical Specification radial peaking factors and the implementation of margin recovery programs.

2-4 Table 2.2-1 The Containment Pressure This change is being made so that the High Trip: Allowable value trip setpoint is consistent with the is being reduced from 5.0 assumptions made to the containment psig to 4.1 psig. pressure High High trip setpoint in the LOCA containment pressure and the pre-trip steam line inside containment analyses, Section 3.3.4.

2-5 Table 2.2-1 Change design reactor coolant All analyses sensitive to minimum flow flow from 370,000 gpm to requirements were performed assuming 363,000 gpm on Footnote (*). a 363,000 gpm minimum guaranteed flow rate.

2"9 Figure 2.2-3 Replace figure with The TM/LP LSSS is being changed to revised figure. reflect analysis at 2700 MWT, Technical Specification radial peaking factors, and the implementation of margin recovery'rograms.

2-10 Figure 2.2-4 Replace figure with The TM/LP LSSS is being changed to revised figure. reflect analysis at 2700 MWT, Technical Specification radial peaking factors, and the implementation of margin recovery programs.

~Refers to sections contained in Reload Safety Report.

Page 2 of 7 S cification Action Remarks B2-1 B2.1.1 Change minimum DNBR limit The value of DNBR, which corresponds to the from "1.20" to read "an 95/95 criteria, changes slightly from acceptable limit". cycle to cycle due to the application of statistical uncertainty analysis; specific values of the DNBR limit are being deleted to avoid the necessity of cycle-by-cycle Tech. Spec. Revisions.

B2-1 B2.1.1 Change statement on DNBR The value of DNBR, which corresponds to the B2-4 B2.2.1 from "1.20" to "the acceptable 95/95 criteria, changes slightly from minimum DNBR limit". cycle to cycle due to the application of statistical uncertainty analysis; specific values of the DNBR limit are being deleted to avoid the necessity of cycle-by-cycle Tech. Spec. revisions.

B2-2 Figure B2.1-1 Replace figure with Figure is being changed to reflect higher revised figure. radial peaking.

f 3/4 1-3 3/4.1.1.2. Change shutdown margin for The shutdown margin is being increased Mode 5 from 2.0% delta k/k to reflect the assumptions used in the to 3.0% delta k/k. boron dilution event, Section 3.2.4.4.

3/4 1"8 3.1.2.2 Change shutdown margin from To be consistent with Technical 3/4 1-10 3.1.2.4 2.0% delta k/k to Specification 3/4.1.1.2.

3/4 1-12 3.1.2.6 3.0% delta k/k.

3/4 1-14 3.1.2.8 Change shutdown margin at To be consistent with Technical 200'rom 2.0% delta k/k Specification 3/4.1.1.2.

to 3.0% delta k/k.

Page 3 8 'cification Reaarks 3/4 1-18 3.1.3.1 Reduce number of CEA regulating Due to change in number of CEA 3/4 1-19 groups from 6 to 5 in Items b;2 regulating banks.

3/4 1-19a and h. of the Action Statement.

Reword Item d. to reflect use of figure showing dropped CEA recovery time vs. measured Pr.T Remove Footnote (N) which Change to reflect higher radial peaks showed the time constraints used in analysis to support increased on a single CEA drop. This is dropped CEA recovery time flexibility.

now contained in Item d. which includes a figure showing dropped CEA recovery time vs. measured Fr.

Resequence Items e. through g.

to reflect addition of new Item e.

3/4 1-24 3.1.3.4 Change CEA drop time from This reduced time is consistent with 3.0 seconds to 2.7 seconds. plant measurements.

3/4 1-28 Figure 3.1-2 Replace figure with The PDIL is being changed to accommodate revised figure. the new CEA rod pattern.

3/4 2-4 Pigure 3.2-2 Replace figure with LHR Ex-core LCO is being revised to reflect revised figure. analysis at 2700 MMt, Technical Specification radial peaking factors, and the implementation of margin recovery programs.

3/4 2-5 Figure 3.2-3 Replace figure with Allowable combinations of thermal power and revised figure. Pr f FxyT are be ing revised to ref lect analysis at 2700 MWt and the implementation

.of margin recovery programs.

Page Action Remarks 3/4 2-7 3.3 2 Change the FTx limit The value for FxyT limit is raised to from 1.60 to .75. reflect the value used in the safety analysis.

3/4 2;9 3.2.3 Change the Fr limit The value for FrT limit is raised to reflect from 1.60 to 1.70. the value used in the safety analysis.

3/4 2-9 4.2.3.2 Delete all references to Rod bow penalties have been accommodated in 3/4 2-11 Table 3.2-1 rod bow penalty. the revised DNBR limit of 1.28.

3/4 2-12 Figure 3.2-4 Replace figure with The DNB-LCO is being changed to reflect revised figure. analysis at 2700 MWt, Reactor Coolant Flow of 363,000 gpm, Technical Specification radial peaking factors, and the implementation of margin recovery programs.

3/4 2-15 Table 3.2-2 Increase upper bound of Upper bound cold leg temperature change cold leg temperature reflects safety analysis assumptions from 548'F to 549"F. performed for Cycle 2.

Decrease reactor coolant All analyses sensitive to minimum flow flow rate from 370,000 requirements were performed assuming gpm to 363,000 gpm. a 363,000 gpm minimum guaranteed flow rate.

3/4 3-6 Table 3.3-2 Change Containment Pressure This reduced time is consistent with High response time from 1.55 plant measurements.

seconds to 1.15 seconds.

3/4 3-17 Table 3.3-4 Changed containment spray This change was made to be consistent on Containment Pressure with assumptions in the High - High Trip Setpoint from containment pressure analysis.

9.30 psig to 5.40 peig and the allowable value from 9.40 psig to 5.50 psig.

Change the Containment Pressure This change was necessary because of the High Trip Setpoint from 5.0 change made to the Containment Pressure psig to 4.7 psig and the High - High Trip Set Point.

allowable value from 5.10 psig.

to 4.80 psig.

Page of 7 S cification Action Remarks 3/4 3-20 Table 3.3-5 Change Feedwater Isolation This change is being made to incorporate Response Time from <5.35/5.35 the specified valve closing time and to to <5.15/5.15 for both eliminate the 0.25 second additional Containment Pressure - High conservatism that was assumed in and Steam Generator Pressure- Cycle l.

Low.

3/4 4-9 3.4.3 Change minimum and maximum This change is being made to be consistent pressurizer indicated level with a new pressurizer level program from 65% to 68.0%. and assumptions made in the excess charging event, Section 3.2.5.1.

3/4 7-1 3/4.7.1 Replace these pages with Changes made to allowable power values 3/4 7-2 Table 3.7-1 revised pages. reflect analysis at 2700 MWt. Format of 3/4 7-3 Table 3.7-2 specification has been changed to improve clarity.

3/7 7-10 3.7.1.6 Change full closure times of These changes reflect appropriate closure 5.6 'seconds and 5.35 seconds times for the main feedwater isolation valve both to 5.15 seconds. (5.15 seconds was assumed in peak containment pressure analysis.)

B3/4 1-1 B3/4.1.1.1 Change the required shutdown The shutdown margin is being increased B3/4.1.1.2 margin with Tavg < 200'rom to reflect the assumptions used in the B3/4 1-2 B3/4.1.2 2.0% delta k/k to 3.0% boron dilution event, Section 3.2.4.4.

delta k/k.

Page 7 Chan Ho. Pacae S cification Action Remarks B3/4 1-4 B3/4.1.3 Remove wording indicating at Change wording, since power levels at what power levels a DNBR SAFDL which a DNBR SAFDL violation may occur violation could occur, and could vary slightly from cycle to cycle.

clarify the wording on how this potential violation is eliminated.

Increase steady-state radial peak from FrT ~ 1.60 to FTr 1 70 B3/4 1-4 B3/4.1.3 Change additional margin from Tr 1.50 to FrT<1 70~

'hese actual radial peak for changes reflect the assumptions utilized in the single drop CEA analysis found in Section 3.2.4.3.

Change Item 5 from a 30 minute misalignment time for an FT < 1.50T to 60 minutes for an FTr< 1.55.

B3/4 2-2 B3/4.2.2, Delete last paragraph which Rod bow penalties have been accommodated B3/4.2.3 6 discusses rod bow penalties, in revised DNB limit of 1.28.

B3/4.2.4 and delete table on rod B3/4 2-3 Table B3/4.2-1 bow penalties.

B3/4 2-2 B3/4 2 ' Change "minimum DNBR limit of The value of DNBR, which corresponds to the

> 1.20" to "an acceptable 95/95 criteria, changes slightly from minimum DNBR". cycle to cycle due to the application of statistical uncertainty analysis; specific values of the DNBR limit are being deleted to avoid the necessity of cycle-by-cycle Tech. Spec. revisions.

Page Chan e Ho. Pacae S cification Action Remarks B3/4 7-1 B3/4.7.1.1 Replace page with revised page. Changes made to allowable power values reflect analysis at 2700 MWt. Format of specification has been charged to improve clarity.

5-3 5.3.1 Change "...236 fuel rods clad..." to "...236 fuel This new statement is appropriate if assemblies with poison rods are loaded and poison rod locations. into the core. Cycle 2 will contain such All fuel and poison rods assemblies.

are clad..."

Change "...a maximum total The weight of 1698.5 grams is a Cycle 1 weight of 1698.5 grams uranium" to "...approximately maximum weight. By wording it 1700 approximately 1700 grams, variations in grams uranium". loading wei.ghts can be tolerated.

5-1 5.2.1 Change containment net free Change in this value represents a more volume from 2.5 x 10 ft detailed analysis of the containment net to 2.506 x 10 ft free volume.

5-3 5.3.2 Increase the number of full- Eight full length CEAs are being added into length control element vacant part length CEA locations.

assemblies (CEAs) from 83 to 91;

0 NO SIGNIFICANT HAZARDS CONSIDERATIONS ST. LUCIE 2 CYCLE 2 OPERATION AT 2560 MWZH DOCKET: 50-389 LICENSE: NPF-16 MAY 1984

Page 1 INTRODUCTION The requested amendment to the St. Lucie Unit 2 operating license is being submitted in support of the upcoming Cycle 2 core reload. The reepested amendnent will incorporate technical specification changes as discussed in the evaluation. The reload will involve replacing approximately one-third of the reactor core and additional new Control Element. Assemblies will be installed in existing=equipped locations.

The Region D fresh fuel assemblies to be used in this reload are not significantly different from those previously found acceptable to the NRC for St. Lucie Unit 2 Cycle 1. The analytical methods used to deaanstrate conformance with the technical specifications and regulations have been previously approved"by the NRC staff. In addition, the proposed technical specification. changes do not change the applicable acceptance criteria previously approved by the NRC Staff. The evaluation performed in suppor't of this amendment has determined that, when measured against the standards in 10CFR50.92, no significant hazards consideration exists. It is also concluded that this amendm nt involves no unreviewed safety questions per 10CFR50.59.

T1".CHNICAL SUMKQK The St.,-Lucie Unit 2 nuclear pcwer plant is presently licensed to operate at a rated thermal pcwer of 2560 Mwth with a physical configuration as defined and described by the FSAR. This reload involves renoving depleted fuel assemblies from approximately one-third of the nuclear core and replacing them with fresh fuel of a similar type as previously loaded. The magnum nominal enrichment of the Region D fresh fuel will be 3.65 weight percent uranium 235 as conlpared to a nominal maximun enrichment in Cycle 1 of 2e73 w/o. The fresh fuel assemblies will also incorporate minor .dinensional changes as a result of design. changes recognized as desirable at other C-E plants.-, These changes create a larger space between the top of each fuel rod.and the fuel upper end fitting flow plate thus all(wing greater. space for fuel rod expansion. The, fuel assembly guide tubes will be changed fran cold worked zircaloy,to annealed zircaloy which will result in a 1cwer growth rate of the-fuel assembly. The increase in enrichment is incorporated in the Region D fuel assemblies to provide for an extended fuel cycle length., There has been no change to the, fuel design bases and as such the new fuel"continues to satisfy General Design Criteria 10 and ll and other design bases considered in the Staff review of the fuel for Cycle 1.

Page 2 BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION An evaluation of this request for amendment has been perforrred to denanstrate that no significant hazards consideration exists, based upon a canparison with the criteria of 10CFR50.92(c). The requested technical specification changes have been categorized into several subheadings for the purposes of this evaluation.

A. MANGES TO Kk'ETY LIMITS Refinem nts in calculational techniques have led to the follcwing two proposed changes.

1. The minimum 'value of the DNBR during steady-state operation, normal operational transients and anticipated transients is increased from 1.20 to 1.28.

2. The alliable limit on peak linear heat rate of the fuel. is increased frcm 21 kw/ft to 22 kw/ft.

There has been no change to the criteria used to establish these safety limits. The proposed M3R value still provides at least a 95% probability at a 95% confidence level that Departure free Nucleate Boiling (DNB) does not occur on a fuel rod having that minimum DNBR during steady state operation or during anticipated operational occurrences. The evaluation of'the various factors associated with DNB will now be based on the Statistical I

Combination of Uncertainties (SCU) methodology (Appendix of the Reload Safety Report). This rrethodology also incorporates adjustments for rod bow directly in the E5B limit, whereas in the reference cycle (Cycle 1) rod bow was acccunted for explicitly in the nanitoring of the radial peaking factor. The SCU methodology is described in C-E report CEN-123(F)-P, and has been previously reviewed and approved by the NRC. Application of the techniques to the plant specific param ters of St. Lucie Unit 2 is described in the accompanying Reload Safety Report.

The proposed new value for peak linear heat rate is still a value corresponding to centerline fuel melt as determined by the fuel evaluation model, FATED'. The pcver-to-centerline melt limit for Cycle 2 takes credit for decreased pcwer peaking which is characteristic of highly burned fuel. Also, since a decrease in fuel melt terrperature accompanies burnup, the rmst limiting pcver-to-centerline melt has been found to occur at an interm diate burnup range. Using conservative estimates of the burnup point at which the poorer peaking begins to decrease and the rate at which it decreases for Cycle 2, the rmst limiting power-to-centerline melt has been determined to be in excess of 22 kw/ft.

Page 3 These revised safety limits have been factored into the safety analyses performed for this reload application and all results are within previcvsly established criteria and design basis; hence, no reduction in safety margin has resulted from these changes.

These technical specifications provide a numerical value with which to judge and verify the acceptability of safety analyses that are performed. Therefore, these changes have no impact on accident probability and consequence, for either accidents previously analyzed or the potential for different accidents.

Therefore, these proposed changes may be considered similar to the example in 10CFR50.92 for anendments that are considered not likely to involve significant hazards considerations:

"(vi) A change which either may result in some increase to the probability or consequences of a previously-analyzed accident or may reduce in scme way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or ~nent specified in the Standard Review Plan; for example, a change resulting from the application of a small refinarent of a previously used calculational riedel or design method."

TlKSNICAL SPECIFICATION CHANGES TO EX%ANCE OPERATING MMGIN The allied plant operating space as defined and controlled by the technical specifications is revised in the follcwing areas.

1. The allcwable planar radial peaking factor (F ) has been increased frcm 1.60 to 1.75 and the allcwRle integrated radial peaking factor (F r) has been increased frcm 1.60 to 1.70.

2. The minimum required Reactor Coolant, System (RCS) flew has been reduced frcm 370,000 gpm to 363,000 gpm.

3. The maximum all(wed cold leg temperature has been increased frcm 548 F to 549 F.

4. Increased restrictions to the LSSS and LOOs are implem nted to offset the effects of the increased operating space produced by items 1, 2 and 3.

Page 4 Detailed calculations were perforned to evaluate the impact of these changes on Anticipated Operational Occurences and Postulated Accidents. The extent of these analyses can be characterized within the foll<wring six categories.

1. Increase in heat renaval by the secondary system.

(Section 3.2.1)

2. Decrease in heat rival by the secondary system.

(Section 3.2.2)

3. Decrease in reactor coolant fle+rate.

(Section 3.2.3) r

4. Reactivity and pcwer distribution ancnalies.

(Section 3.2.4)

5. Decrease in reactor coolant system inventory.

(Section 3.2.6)

6. Loss of Coolant events.

(Section 3.3)

HOTE: Section numbers refer to the sections in the Reload Safety Report.

The criteria for judging the acceptability of these events has not changed from the reference cycle (Cycle 1). The detailed results of these calculations are provided in the accmpanying Reload Safety Report along with comparisons with the appropriate limiting criteria. The follcving discussion provides a summary of various events analyzed with respect to the three basic criteria; i.e., offsite dose, reactor coolant system pressure, and fuel performance.

l. Offsite Dose Acceptance guidelines for offsite radiation dose continue to be based on 10CFR100 criteria. The nest limiting postulated accident with respect to offsite dose was determined to be a steamline break outside of containment (Section 3.2.1.5b).

The detailed analysis of this postulated accident includes assumptions such as concurrent loss of AC pcver and the rmst adverse values for the process parameters (RCS temperature, pressure, core MID, NSSS pcwer, etc.) that affect the outccme of this event. Even with the conservatism assurred, the results are well within the limits of 10CFR100. The consequences of a steamline break inside containment are

Page 5 even less severe with respect to offsite dose since the releases are confined within the containment building.

The limiting Anticipated Operational Occurence which is analyzed for impact on offsite dose is the Inadvertent Opening of a Steam Generator Safety Valve (Section 3.2.1.4). It is assumed that this event will result in a ccaplete blcwdcxm of. one steam generator and partial blnrdown of the other. Conservative assumptions to maximize the calculated doses include maxinam steam generator and RCS radionuclide concentrations. The results continue to be a small fraction of 10CFR100 limits.

2~ Reactor Coolant S stem Pressure Acceptance guidelines for RCS pressure are based on RCS design limits as defined by General Design Criteria 14 ard

15. The rmst limiting postulated accident with respect to KS pressure was found to be a feedwater system pipebreak (Section 3.2.2.6). This event is analyzed with conservative assumptions, such as loss of AC parer and the rrost adverse values for the process paraneters that affect the results.

Also, a parametric evaluation is performed to identify the exact break size that maximizes the RCS pressure peak.

These conservative calculations show that the pressure peak resulting from this event is still below the RCS upset pressure limit of 2750 psia.

The limiting Anticipated Operational Occurence which affects KS pressure is the Loss of Condenser Vacuum event (Section 3.2.2.3). The resulting loss of load causes an increase in steam generator pressure which is relieved by opening of the secondary safety valves. There is also an increase in RCS pressure which allcws protective systems to initiate a reactor trip at the high pressure setpoint to terminate the event. The peak RCS pressure attained is well below the upset pressure limit of 2750 psia.

3. Fuel Perforaance Criteria in this category recpire that a eoolable fuel geometry is maintained such that continued rerroval of decay heat is ensured. This condition is met by maintaining fuel temperatures below the Specified Acceptable Fuel Design Limit (SAFDL) and limiting the duration of DHB during postulated accidents. The nest limiting postulated accident with respect to fuel integrity was determined to be the Steamline Break Outside of Containment (Section 3.2.1.5b).

Page 6 Note that this event has been previously discussed as the limiting postulated accident with respect to offsite- 'mst dose. The result indicates that only a small nurser of fuel pins are predicted to fail and a eoolable geometry is maintained.

The limiting Anticipated Operational Occurrence that is considered in this category is the Total Loss of Forced KS.

Flew (Section 3.2.3.2). The conditions assum d in this analysis include the maximum allied cold leg temperature, maximum radial peaking factors and minima RCS flnrrate as-proposed. A parametric analysis is performed to determine the axial shape index within the allcwable range that provides the rmst severe results. This event is used to establish the minimum initial margin that must be maintained by the Limiting Conditions for Operation (LCDs) with respect to the DNBR limit. Hence, this event results in an acceptable minimum KM3R of 1.28.

Another set of criteria that is established to evaluate fuel performance is described by 10CFR50.46. Assurance that these criteria are satisfied is provided by the detailed analyses performed for small break IDCA, large break IDCA and post-IDCA long term cooling. The highest Peak Clad Temperature (PCT) calculated, resulted from a Double-Ended Guillotine Break at Pump Discharge (DZQ/PD) with a PCT of 2041'F as contpared to an allowable limit of 2200'F. A detailed description of these analyses and corresponding results is provided in the Reload Safety Report (Section 3.3.1). Xn all cases, the analytical results shor acceptability with respect to the 10CFR50.46 criteria.

These detailed calculations shcw that incorporation of the increased operating space, when offset by the rmre limiting restrictions imposed by changes to the LSSS and LCOs result, in limiting events which are still below the corresponding acceptance criteria. Therefore, no reduction in safety margin has occurred.

The ccabined results of these calculations when ccmpared to the reference cycle (Cycle 1) show that these proposed changes do not result in any increase in the probability of those events previously analyzed and no significant increase in the consequences of these events can be shcwn.

None of these proposed changes result in any nadifications to plant equipment; the minor variations in plant paraneters accounted for in the evaluations of AOOs and postulated 're

Page 7 accidents as described'bove. Therefore, this evaluation has further concluded that these changes do not provide a potential for accidents different from those previously considered.

Since these proposed changes yield results which are well within acceptance criteria, the changes can be considered similar to the example provided in 10CFR50.92 for amendm nts that are considered not likely to involve significant hazards considerations:

"(vi) A change which either may result in sane increase to the probability or consequences of a previously-ana1yzed accident or may reduce in scxne way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or ccmgenent specified in the Standard Review Plan; for example, a change resulting from the application of a small refinarant of a previously used calculational model or design method. "

The rmde 5 shutdcwn margin is increased from 2% to 3% delta k/k as a result of the fuel management program which will permit an increased cycle length.

The acccmpanying Reload Safety Report analyzes anticipated operational occurrences that are affected by the proposed changes in cycle length and Mode 5 shutdcwn margin. The limiting events with respect to radiological release and loss of shutdown margin are the follcwing.

l. Inadvertent opening of a steam generator safety valve (Section 3.2.1.4) .

2. Post trip analysis of a steam line break frcm Hot Full Pcurer (Section 3.2.1.5.c) .

3. Chemical Volurre and Control System (CVCS) malfunction (Section 3.2.4.4) .

These analyses were perforned with bounding values of shutdcwn margin, rod worth, and boron worth for the current fuel loading.

The results from an analysis of the inadvertent opening of a

Page 8 steam generator'afety valve shou that reliable control of reactivity is maintained and that radiological doses at the site boundary are a smll fraction of the 10CFR100 guidelines. The, steam line break analysis shcws that, with the same HZP shutdown requiratent as for the previcus cycle, there will be no significant return to pcwer. Analysis of the QlCS malfunction (boron dilution) shears that under all operating and refueling conditions the time from annunciation to criticality will meet or exceed the required minimum criteria. Thus, all criticality criteria are met. Increases in fuel temperatures and coolant pressures are regulated by the constraints imposed by the LSSS and LO3s.

From these analyses it in can be concluded teat there is no the probability and consequences of significant increase accidents previously analyzed. Nor do these changes create the possibility of a new or different kind of accident. The changes do not reduce the safety margin inasmuch as the safety analyses show that acceptable results are obtained with the same criteria pertaining to offsite dose rates, return to parer and tine fran annunciation to criticality.

This change can be considered as being similar to the example in 10CFR50.92 for amendments that are considered not likely to involve significant hazards considerations:

"(ii) A change that constitutes an additional limitation, restriction, or control not presently included in the technical specifications; for example, a rare stringent surveillance requirement."

Cycle 2 will incorporate several changes related to the Control Element Assemblies (CEAs), primarily to enhance operational characteristics such as control of axial shape index. A description of the physical configuration of these changes is provided here, along with a summuy of the affected technical specifications.

Eight additional CEAs will be installed in core locations which are presently unrodded. These locations are already provided with drive rmtors, position indicating instrum ntation and all associated hardware. This change will take advantage of these equipped locations to increase the total number of CEAs available for reactor control. These additional CEAs will also result in an increase in the available shutdcwn margin. The sequence in

Page 9 which these eight new CEAs and the existing 83 CEAs are maneuvered will be changed. The 83 CEAs in the reference cycle (Cycle 1) are suMivided into six regulating and two shutdcam banks. The 91 CEAs available for Cycle 2 will be suMivided into five regulating and two shutdcwn banks. This grouping change will increase the number of CEAs from four to twelve in the first sequentially inserted group during reactor control maneuvers.

Also, the CEA insertion limitation (Pnrer Dependent Insertion Limit, PDIL) will be revised. The changes to group configuration and PDIL will increase the anaunt of control available to plant operators and will alice for a narc even application of CEA.

worth, which will minunize the effects on core radial pcwer distribution.

Safety analyses have been performed to verify the acceptability of increasing the a@cunt of time allured to recover a dropped CEA. Plant experience with operational surveillance has sheen that the actual CEA, drop time associated with a reactor trip is conservatively faster than previously assumed for the reference cycle. Therefore, changes concerning CEA recovery time and CEA drop tine will be incorporated into the technical specifications.

Detailed analyses of Anticipated Operational Occurrences which were performed to confirm the acceptability of these changes include the following:

l. Uncontrolled CEA withdrawal from a subcritical or lcd pcwer condition (Section 3.2.4.1) .

2. Uncontrolled CEA withdrawal at power (Section 3.2.4.2) ~

3. CEA misoperation (rod drop) (Section 3.2.4.3).

Analysis of these events have shcwn that there is no significant increase in the consequences of these events resulting from the proposed changes. The postulated accident which is ~st significantly affected by these proposed changes is the CEA Ejection Event. This event wculd result from the highly unlikely failure of a pressure housing which retains a CEA. The analysis of this event is performed in accordance with the NRC approved C-E methodology described by CENPD-190A, (Section 3.2.4.6). The analysis shcvs that the rrast severe results, which occur at a zero pcver initial condition, predict that no fuel failures will occur. Therefore, acceptance criteria related to fuel performance and offsite dose are satisfied and no reduction in safety margin has resulted from these changes.

Page l0 These, changes are similar to the example in 10CFR50.92 for amendments that are considered not likely to involve significant hazards considerations:.

~ I

"(vi)-A change which either may result in same increase,to the probability or consequences of a previously-analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or canponent specified in the Standard Review Plan; for example, a change resulting from the application of a small refinement of a previously used calculational rmdel or design method."

The physical implementation of these changes will be accmplished by mx3ifications to the existing Control Element Drive Mechanism Control System (CEDMCS) . Originally defined functional requirements and specifications for this equipment will be retained and, hence, there will be no impact on the probability of previously analyzed events and no potential for new events.

To assure containment integrity, the follcwing changes are proposed:

1. Containment spray high-high trip setpoint is leered from f

9.30 psig to 5.40 psig and alliable values rem 9.40 psig to 5.50 psig.

2. The high containm nt pressure setpoint for Engineered Safety Features (ESF) functions is leered fran 5.0 psig to 4.7 psig. The allowable value is reduced from 5.1 psig to 4.8 psig. The high containment pressure setpoint of, 4.0 psig for reactor trip remains the sane as Cycle 1, hcwever, the alliable value is reduced fran 5.0 psig to 4.1 psig.

3. The- alliable response tine for high containment pressure instrumentation is reduced fran 1.55 seconds to 1.15 seconds.

These changes are similar to the example in 10CFR50.92 for amendments that are considered not likely to involve significant hazards considerations:

Page ll

"(ii) A change that constitutes an additional limitation, restriction, or control,not presently included in the technical" specifi-cations: for exanple a narc stringent surveillance requirement. "

The lever containm nt spray trip setpoint results in liower peak containamt pressure following mass and energy releases to the contairurent. The high containment pressure setpoints have been reduced as a result of the high-high containment pressure setpoint changes, to assure proper sequencing of autmatic safety system actions. The reduction in response time is justified based on in-plant experience with instrument performance. Section 3.3.4 of the Reload Safety Report sos that with the proposed changes, a higher core poorer (2700 Mph) can be accanrmdated without comprcmising containment integrity.

The report presents analyses that she@ peak containm nt pressures for a large break ZQCL or a main steam line break, the two limiting transients for containment pressure, will be belch the design pressure of 44 psig. Thus, the probability and consequences of previously analyzed events have not increased nor has the safety margin decreased. The probability for a new accident has not increased as no new failure mechansim has been introduced. The lower limit on initial containment pressure has not been changed, thereby assuring that the assurrptions used in the HCCS analysis remain valid.

PRESSURIZER WATER UWEE A change to the pressurizer water level control system is incorporated to raise the normal operating water level in the pressurizer. This level program improvement will provide greater margin between the pressurizer heater cutoff level setpoint and the projected minimum water level follcwing a reactor trip.

Consequently, to accaramdate this control system setpoint change, the maximum alliable indicated pressurizer water level is increased fran 65% to 6ES. This change has been accounted for in analysis of a CVCS malfunction (Section 3.2.5.1) which is the limiting event affected by this change. The analysis concludes that the operator has 20 minutes available to take corrective action follcving annunciation of the high pressurizer water level alarm to prevent filling the pressurizer. This is a sufficient and acceptable period of time for the operator to terminate the charging-letdcm flex imbalance and hence no reduction in safety margin has occurred. This change also has no affect on the probability or consequence of new or previously analyzed accidents.

Page 12 This increase in allowable pressurizer water level is similar to the example in 10CFR50.92 for amendments that are considered not likely to involve significant hazards considerations:

"(vi) A change which either may result in sane increase to the probability or consequences of a previously-analyzed accident or may reduce in scne way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or carponent specified in the Standard Review Plan; for exanple, a change resulting from the application of a small refinement of a previously used calculational model or design method."

G. SE005EARY SAPPY VALVE The main steamline safety valve operability requirenent is changed to incorporate revised maximum alliable power limits to be in effect when fewer than all safety valves are in service.

This specification will new be of the same format and technical content as the corresponding St. Lucie Unit 1 requiraaent. The same calculational methods used for the reference cycle (Cycle 1) are applied here and no increase or decrease in rated valve capacity is assumed. The analyses which support this change are rue based on steam flnnates which wmld be present with the plant operating at 27001%th. The revised specification continues to ccnply with the ASME Boiler and Pressure Vessel Section III code requirements to limit peak secondary system pressure to 110%

of design pressure. Therefore, no reduction in safety margin has occurred, and the probability/consequence of accidents is not f

af ected.

Since this change results in a reduction in the allied fractional pcwer level, the change may be considered similar to the example in 10CFR50.92 for amendm nts that are considered not likely to involve significant hazards considerations:

(ii) A change that constitutes an additional limitation, restriction, or control not presently included in the technical specifications; for example a rrore stringent surveillance requirement."

Page 13 H. CORRECTIONS AND ADMINISTRATIVECHANGES The follcwing two changes constitute editorial corrections in the existing technical specifications:

Section 5.3.1 change "fuel rods" to "fuel and poison rods" to include fuel assemblies containing poison rods.

2~ Section 5.3.1 change "1698.3 grams" to "approximately 1700 grams" to permit minor variations in core loading and weight.

These changes are of an administrative nature and follcw the example given in 10CFR50.92 for amendments that are considered not likley to involve significant hazards considerations:

"(i) A purely administrative change to technical specifications: for example, a change to achieve consistency throughout the technical specifica-tions, correction of an error, or a change in mnenclature. "

IV. CONCLUSION From the considerations detailed above it can be. concluded that the proposed amendments to the St. Lucie Unit 2 Technical Specifications do not a) increase the probability or consecpences of accidents previously analyzed b) increase the potential for accidents different from any accident previously considered c) reduce the safety margin.

Therefore of it istheconcluded that in accordance with the provisions 10CFR50.92 changes involve no significant hazards considerations.