ML17345B253

Significant Hazards Evaluation for Fh/Fq Proposed Amend to Licenses DPR-31 & DPR-41 Dtd 830819

ML17345B253
Person / Time
Site:	Turkey Point
Issue date:	09/09/1983
From:	Robert E. Uhrig FLORIDA POWER & LIGHT CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
L-83-477, NUDOCS 8309150128
Download: ML17345B253 (20)

Text

REGULATORQtNFORNATION DISTRIBUTION +TEN (RIDE)

ACCESSION NBR 8309150128 DOC DATE E 83/09/09, NOTARIZED: NO DOCKET 0 FACIL:50-250 Turkey Point Planti Unit 3F Florida Power and Ljght C 05000250 50-251 Turkey Point Planti Uni't PF Florida Power and L.ight C 05000251 AUTH AUTHOR AFFILIATION UHRIGP R E. Florida Power & Light Co.

REC IP NAME RECIPIENT AFFILIATION

.EISENHUTiD.G ~ Division of L,icensing

SUBJECT:

Significant hazards evaluation for FH/FQ proposed amend to Licenses DPR-31 8 DPR-41 dtd 830819 ~

DISTRIBUTION CODE: A001S COPIES RECEIVED:LTR g ENCL SIZE:

TITLE: OR Submittal: Gener al Distr ibution NOTES RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL NRR ORB 1 BC 01 7 7 INTERNAL: ELD/HOSE 0 NRR/DF/MTEB NRR/DL DIR 1 1 NRR'/DL/ORAB NRR/OS I/METB 1 1 NRR/DSI/RAB I I 04 1 1 RGN2, EXTERNAL: ACRS 09 6 6 LPDR 03 NRC PDR 02 1 1 NSIC 05 NTIS 1 1 TOTAL NUMBER OF COPIES REQUIRED: LTTR 25 ENCL 23

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O. BOX 14000, JUNO BEACH, FL 3340B FLORIDA POWER & LIGHT COMPANY September 9, I 983 L-83-477 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: Turkey Point Units 3 & 4 Docket Nos. 50-250 8 50-25 I Proposed License Amendment FgH/FQ Attached is the Significant Hazards Evaluation for the F~H/FQ Proposed License Amendment dated August I9, I983 (L-83-455).

Also attached is a list of corrections and corrected pages to Attachment A and B of the above referenced letter. Please replace the corrected pages in that submittal.

Very truly yours, Robert E. Uhrig Vice President Advanced Systems 8 Technology REU/SA V/cab Attachment cc: J. P. O'Reilly, Region II Harold F. Reis, Esquire Ulray Clark, Administrator Radiation Health Services Tallahassee, Florida 3230 I 8309150128 830909 PDR ADOCK 05000250 P PDR PEOPLE... SERVING PEOPLE

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Significant Hazards Evaluation FqH/ Fa Technical Specification Changes Turkey Point Units 3 8 4 Three primary changes are proposed:

I. The hot channel factor F<H limit is increased from 1.55 to 1.62.

2. The total peaking factor Fg limit is increased from 2.30 to 2.32.

3. The Overpower 5T setpoints and thermal-hydraulic limit curves are made more conservative.

l. F~H Limit increase 10 CFR 50.92 (c)(l): The increase in F<H limit from 1.55 to 1.62 does not significantly increase the probability or consequences of accidents previously analyzed for the following reasons:

a) The increase in the hot channel F~H limit entails no physical changes in plant equipment or operating procedure and therefore will not increase the probability of design basis accidents analyzed in the FSAR.

b) The concern with higher FgH limit is the possible occurrence of fuel failure. The safety analysis shows that the proposed increase in the FgH limit does not lead to departure from nucleate boiling (DNB) in the core and that, therefore, there would not be significant fuel failure. The consequences of previously analyzed accidents are, therefore, not significantly increased.

10 CFR 50.92 (c)(2): The increase in FgH limit from I'.55 to 1.62 does not create the possiblity of a new or different kind of accident from any accident previously evaluated for the following reasons:

The change involves no plant equipment or operating procedure changes; therefore there is no possibility of a new accident not previously analyzed.

10 CFR 50.92 (c)(3): The increase in F>H limit from 1.55 to 1.62 does not cause significant reduction in margin of safety explained as follows:

The safety evaluations show the fuel to be within the bounds of the same fuel failure criteria as before. Therefore, the proposed change does not cause a significant reduction in margin of safety.

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This proposed amendment compares closely to example (vi) as listed in the "Examples of Amendments that are not considered likely to involve Significant Hazards Considerations," 48FR I 4870~4/I 6/83) as shown below.

Identification of additional DNBR margin to accommodate the reduction in margin resulting from the increased FgH limit meets the FSAR design basis therefore it is:

(vi) A change which either may result in 'some 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 component specified in the Standard Review Plan: for example, a change resulting from the application of a small refinement of a previously used calculational model or design method.

2. Fg Limit Increase IO CFR 50.92 (c)(l): The increase in Fg limit from 2.30 to 2.32 does not-significantly increase the probability or consequences of accidents previously analyzed for the following reasons:

a) The limiting accidents for total peaking factor Fg are the small and large break loss of coolant accidents (LOCA). The increase in the Fg limit involves no changes in plant equipment or operating procedure and therefore will not affect the probability of a LOCA; therefore the probability of design basis accidents analyzed in the FSAR is not increased.

b) The consequences of a LOCA would be fuel failure and associated fission gas releases. Analyses of small and large break loss of coolant events for 2.32 Fg show peak clad temperatures and other core parameters indicative of fuel failure well below the acceptable limits of I 0 CFR 50.46.

Therefore, it is concluded that the proposed change will not significantly increase consequences of the previously analyzed accidents.

IO CFR 50.91 (c)(2): The increase in Fg limit from 2.30 to 2.32 does not create the possibility of a new or different kind of accident from any accident previously evaluated because it does not involve equipment or procedure changes that could cause a new accident.

IO CFR 50.9I (c)(3): The increase in Fg limit from 2.30 to 2.32 does not cause significant reduction in margin of safety.

The acceptance criteria of IO CFR 5046 include adequate safety margins. The LOCA analyses for 2.32 Fg predict results well below the acceptance limits. It is therefore concluded that there is no significant reduction in margin of safety.

This proposed amendment compares closely to example (vi) as listed in the "Examples of Amendments that are not considered likely to involve Significant Hazards Considerations", 48FR I 4870747l 6/83) as shown below.

,x The proposed change to increase Fg is shown in the ECCS analysis to remain in compliance with 10 CFR 50.46 therefore it is:

(vi) A change which either may result in some 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, component specified in the Standard Review Plan; for example, a change resulting from the application of a small refinement of a previously used calculational model or design method.

3. Chan e in Over. wer bT Set pints and Thermal-H draulic'Limit Curves The change in the Overpower LT setpoints and thermal-hydraulic limit curves is in the conservative direction. Therefore 10 CFR 50.92 (c)(l),.(2) and (3) are negative.

This proposed amendment compares closely to example (ii) as listed in the "Examples of Amendments, that are not considered likely to involve Significant Hazards Considerations," 48FR I 4870~4/ l 6/83) as shown below.

The Overpower 5T setpoints and thermal-hydraulic limit. curves changes are in the conservative direction therefore they are:

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

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.List of Corrections of L-87,-455 Attachment A Instructions List of Figures Replace page Page 2.3-3 Replace page Figure 3.2-3 Replace page Page 82.1-2 Replace page Page 10 'Section 4.2, 2nd paragraph, delete third sentence "An evaluation has ..."

0 0 LIST OF FIGURES F iciure Title

2. I- I Reactor Core Thermal and Hydraulic Safety Limits, Three Loop Operation

2. I- I a Deleted

2. I- lb Deleted

2. I-2 Reactor Core Thermal and Hydraulic Safety Limits, Two Loop Operation

3. I- I DOSE EQUIVALENT I- l3 I Primary Coolant Specific Activity Limit Versus Percent of RATED POWER with the Primary Coolant Specific Activity I.O Ci/gram Dose Equivalent I-l3I

3. I- I a Reactor Coolant System Heatup and Cooldown Pressure Limits 3.I. I b Reactor Coolant System Heatup and Cooldown Pressure Limits

3. I- Ic Reactor Coolant System Heatup and Cooldown Pressure Limits

3. I- Id Reactor Coolant System Heatup and Cooldown Pressure Limits

3. I -2 Radiation Induced Increase in Transi'tion Temperature for A302-B Steel

3. I -2c Radiation Induced Increase in Transition Temperature for A302-B Steel

3. I -2d Radiation Induced Increase in Transition Temperature for A302-8 Steel 3.2- I Control Group Insertion Limits for Unit 4, Three Loop Operation 3.2- I a Control Group Insertion Limits for Unit 4, Two Loop Operation 3.2-lb Control Group Insertion Limits for Unit 3, Three Looper Operation 3.2- I c Control Group Insertion Limits for Unit 3, Two Loop Operation 3.2-2 Required Shutdown Margin 3.2-3 K(Z) vs. Core Height 3.2-3a Deleted 3.2-4 Maximum Allowable Local KW/FT

4. I 2- I Sampling Locations 6.2- I Offsite Organization Chart 6.2-2 Plant Organization Chart

83. I I Effect of Fluence and Copper Content on Shift on RTNDT for Reactor Vessel Steel Exposed to 550 F Temperature B3. I -2 Fast Neutron Flunece (E IMEV) as a function of Effective Full Power Years B3.2- I Target Band on Indicated Flux Difference as a Function of Operating Power Level B.3.2-2 Permissible Operating Band on Indicated Flux Difference as a Function of Burnup (Typical)

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1. 5000 1.2500

1. 0000 0.1500 0.5000 TOTAL I 0 2.320 COBE llf:1GIIT 0.000 1.000 0.2500 C.OOn 1.000

10. 000 O.Oio 12.000 0.611 0.0 Cl C3 CI C) O O

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C3 AI ED AJ CO C OII L IIL 1G IIT I I' I Figure 3.2-3

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Ov e . c vc r Q T I 1 .Pa -Kl dT K2 (T ) f (f')

p dr, Indicated' zt rz=e" p--ar, r T Average t~~erature, r T I"d catM ave.age re~erat ra at no...."'.nal eood'lo"s and rated pox r, 0 for d'ecreas' average teweratu e; 0.2 sec-/2 For

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'than T for T equal to or =ore than T; 0 for T less Pate of cha"ge of ta= -'rat re, =/sec Ct f (zqs d " '"M above.

Lo= ress ri-er pressure e" ~ to cr greater tha" 1835 'ps'g-ssur'"er press 'a eq ='-"o or "ess than 2385 pWg-essa:ze- ' er -' circa to' r ess ha a 9Q of sc~~e-cs~+A'r

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The curves of Figure 2.1-1 show the loci of points of T.".=">:. ~

PO'. ER, Reactor Coolant System pressure and average temperature for which tne cal culated D!lBR is no less than the design Df'3R value or the average enthalpy

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a ..he vessel exi t i s l ess than the enthalpy of saturated liquid.

The curves are conservative for an enthalpy not channel factor, F(H, of >.62 and a reference cosine with a peak of 1.55 for axial power shape. An allo>>ance is included for an increase in F,H at reduced power based on tne expression:

FcH < 1.62 Ll + 0-3 (1-P)3 where P is the fraction of RATED THER)'iAL POl"-"R.

These limiting heat flux conditions are higher than those cal culated for tne range of al 1 control rods fully wi thdra>>n io the maximum al owable control red 1

inseri1on 1 imit assuming the axial power imba'.ance is wi thin the limits of

(< c) f unct '.on of the Overiemperature I

LT irip. linen the axial power

-...ba ance i s no. ""'thin the io erance, . he axial power imbalance effeci on Over:e2pera t ure ' ir i s wi l recvee tne seip" in.=< to prov i""" "roiecti oli 1

consistent >>~ ih core safety limits.

Fuel rod bowing reduces the values of OfiB ra.io (Ot3R). The amount of tne DI'=R reduciion is 4.7<.'or LOPAR fuel with the L-grid Dhd correlaiion and 5.='or ine GFA fuel wi=n th ilRB- DNB correla ion. The penalties are ca'.c.'ia-.e-pursuant to "Fuel Rod Bow Evaluation," h'CAP-8691-P-A, Rev. 1 (proprietary) and t'CAP-8692 Rev. 1 (non-proprietary). The restrictions of the Core Thermal Hydraulic Safety Limits assure that an amount of ONER margin greater tnan or equal to the above penalties is retained to offset the rod bow DHBR penalty.

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