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| issue date = 05/21/1979
| issue date = 05/21/1979
| title = Submits Analysis Supporting Amend to App a of Licenses DPR-31 & DPR-41.DNBR Safety Limit of 1.24 Is Consistent W/Cobra Iiic Computer Code.Forwards DNBR Safety Limit for Cobra Iiic Analysis.
| title = Submits Analysis Supporting Amend to App a of Licenses DPR-31 & DPR-41.DNBR Safety Limit of 1.24 Is Consistent W/Cobra Iiic Computer Code.Forwards DNBR Safety Limit for Cobra Iiic Analysis.
| author name = URHIG R E
| author name = Urhig R
| author affiliation = FLORIDA POWER & LIGHT CO.
| author affiliation = FLORIDA POWER & LIGHT CO.
| addressee name = STELLO V
| addressee name = Stello V
| addressee affiliation = NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR)
| addressee affiliation = NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR)
| docket = 05000250, 05000251
| docket = 05000250, 05000251
Line 17: Line 17:


=Text=
=Text=
{{#Wiki_filter:REGUI.ATTY INFORNATION DISTRIBUTIO SYSTEN (RIBS), Af'CESSION NBRi7905250426~
{{#Wiki_filter:REGUI.ATTY INFORNATION DISTRIBUTIO               SYSTEN (RIBS)
DOC~DATE;79/05/21 NOT RIZEDt NO DOCK FACIL!50"250 TURKEY POLNT PLANTE UNIT 3g'FLORIDA POWER AND LIGHT C 0500 50 251 TURKEY POINT PLANTg UNIT 4i FLORIDA PO'WER AND"LIGHT C 05000251 AUTH'o NAME AUTHOR AFFILIATION URIGr R~E"~FLORIDA PUWER 8, LIGHT CO.REC IP~NAME RECIPIENT AFF IL'IATION STELLOiV~DI VISION OF OPERATING REACTORS SUB JECT: SUBMITS ANALYSIS SUPPORTING AMEND TO APP A'F OL DPR 31 8 PPR 41~'DETERMINES DNBR SAFETY LIMIT OF 1G24 CONSISTENT W/COBRA IIIC COMPUTER CODK~FORWA'RDS"DNBR SAFETY LIMIT FOR COBRA III:C'NALYSIS DISTRIBUTION CODE!A001S:COP IES RECK I VED t LTR+~ENCL+~SIZE~T'ITLK: GENERAL DISTRIBUTION FOR AFTER ISSUANCE OF OPERATING LIC N ,E-C~4 OT'E S~~~~~~++~<~FR~eirer~~<ree>~~et~~~z+~>~~~~~~>~FR>~reeeeareyere~reeeyeytqe>RN~>we~><yy<yeyyRRNF~<~~~we RECIPIENT ID CODE/NAME'CTION:
, Af'CESSION NBRi7905250426~                                   DOC ~ DATE; 79/05/21     NOT RIZEDt NO         DOCK FACIL!50"250 TURKEY POLNT PLANTE UNIT 3g 'FLORIDA POWER AND LIGHT                                             C 0500 50 251 TURKEY POINT PLANTg UNIT 4i FLORIDA PO'WER AND "LIGHT                                     C 05000251 AUTH'o NAME                                 AUTHOR     AFFILIATION URIGr     R ~ E "~                           FLORIDA PUWER       8, LIGHT CO.
05 BC dR8&/INTERNAI.:
REC   IP ~ NAME                             RECIPIENT AFF IL'IATION STELLOiV ~                                   DI VISION OF OPERATING REACTORS SUB JECT:         SUBMITS ANALYSIS SUPPORTING AMEND TO APP                               A'F OL DPR 31 8 PPR 41 ~ 'DETERMINES DNBR SAFETY LIMIT OF 1G24 CONSISTENT W/COBRA IIIC COMPUTER CODK ~ FORWA'RDS "DNBR SAFETY LIMIT FOR COBRA                 III:C'NALYSIS DISTRIBUTION CODE! A001S:COP IES RECK I VED t LTR +~ ENCL +~ SIZE ~
OI RES"~12 ILE 15 CORE PERF BR 17 ENGR BR 19'PLANT SYS BR'1 EFLT TRT SYS EXTERNAL!g3 LPDR 23 ACRS COPIES LTTR ENCL 7 7 1 1 2 2 1 1 1 1 i''1 1 1 16 16 REC IPIFNT ID COOK/NAME 02 NRC PDR 10 TA/EDO 16 AD SYS/PROJ 18 REAC SFTY SR 20 KEB 22 BR INKMAN'04 NS IC COPIES LTTR ENCL'1 1 1 1 1 1 1 1 1 1 1 1 NN 29'@7>TOTAL NUMBER OF COPIES REQUIRED!LTTR 38 ENCL 38 Qi I~I'J E jw f<k I a 1 IP l I l f~i ea e 4 W C I't 2'%H 2~~'I lk P.o.BOX 529100, MIAMI, FL 33152 Q@ig L..i%FLORIDA POWER 8 LIGHT COMPANY May 21, 1979'L-79-127 Director of Nuclear Reactor Regulation Attention:
T'ITLK: GENERAL DISTRIBUTION FOR AFTER ISSUANCE OF OPERATING LIC
Mr.Victor Stello,,Jr., Director Division of Operating Reactors'U.S.Nuclear Regulatory COIImIission Washington, D.C.20555
      ,E- C ~                                                                                               4 N OT'E S ~ ~~~~~++~<~FR~eirer~~<ree>~~et~~~z+~>~~~~~~>~FR>~reeeeareyere~reeeyeytqe>RN~>we~><yy<yeyyRRNF~<~~~we RECIPIENT                             COPIES              REC IPIFNT          COPIES ID                                      LTTR ENCL        ID COOK/NAME          LTTR ENCL' dR8 & /
                                            "~
CODE/NAME'CTION:
05 BC                                       7      7 INTERNAI.: OI                                                                02 NRC PDR i''
RES                                   1      1                                      1 12 ILE                                     2      2      10 TA/EDO                1      1 15 CORE PERF BR                             1      1      16 AD SYS/PROJ          1      1 17 ENGR BR                                 1      1      18 REAC SFTY SR          1      1 19 'PLANT SYS                   BR'1 20 KEB                          1 EFLT TRT SYS                                   1      22 BR INKMAN            1      1 EXTERNAL! g3 LPDR                                             1       1     '04 NS IC                 1     1 23 ACRS                                    16      16 NN 29'@7>
TOTAL NUMBER OF COPIES                                 REQUIRED! LTTR       38   ENCL     38


==Dear Mr.Stello:==
                                                        'J Qi                I
Re: Turkey Point Units 3&4 Docket Nos.50-250 and 50-251 Proposed Amendment to Facility 0 eratin Licenses DPR-31 and DPR-41 In accordance wi.th, 10 CFR 50.30, Florida Power'&Light Company (FPL')submits herewith three (3)signed Originals and forty (40)copies of a request to amend Appendix A of Faci-lity Operating License DPR-31 and DPR-41.The request has been prepared in response to a'September 15,.1978 letter.from Mr.A.Schwenc'er of your staff..As requested in the September 15,letter"(.Reference 1);, FPL has,,performed an analysis to derive a DNBR safety limit consistent with the COBRA.IIIC computer code.This.code was used'n analysis submitted previously (References 2,3,and 4)to demonstrate the thermal margin available at Turkey Point Units'3 and,4 to compensate for a reduction in'DNBR due to the effects of fuel rod bowing.The NSSS vendor had obtained'DNBR limit of 1.24 from an analysis of critical.heat flux tests, using the THING computer code with the W-3 critical heat flux correlation and the L-grid correction.
                                        ~ I E    jw I
However, this data base is.proprietary and is not available to'FPL.Instead, a corresponding DNBR limit was derived for the COBRA IIIC code by com-paring, the COBRA IIIC results with THING'esults.
f
for a somewhat different data base (Reference 5)which was proposed'n your letter of September 15.The, analysis (attached) yielded a DNBR safety limit of 1.273 for the COBRA IIIC code.Revised"Reactor Core Thermal and Hydraulic Safety Limit" curves for tube plugging levels of<15'A,>15%to<195, and>19/to<25/, based on the derived DNBR limit of 1.273, are attached.Revised Bases pages are also attached.vgpp2 gpss%A"~po v PEOPLE...SERVING PEOPLE 4 P 4 Mr.Victor Stello, Jr., Director Division of Operating Reactors U'.S.Nuclear Regulatory
                            <k IP a   1 l
'Commission Page 2 The information supplied herewith supplements our previous submittals concerning the effects of rod bow on DNBR at Turkey Point Units 3 and 4.It is proposed that the attached Reactor Core Safety Limits be substituted for the present Figures 2.1-1,, 2.l-la and 2.1-1b in the Technical Specifications, and that with these new and more restrictive limits, the units can be operated safely without any other penalties for rod bow.The proposed amendment has been reviewed by the Turkey Point Plant Nuclear Safety Committee and the Florida Power 8 Light Company Nuclear Review Board.They have determined that it does not adversely affect the heal,th and safety of the public.Very truly yours, Robert E.Uhrig Vice President Advanced Systems 5 Technology REU:MAS:ncr Attachment cc: 'Mr.James P.O'Reilly, Region II Robert Lowenstein, Esquire 14 REFERENCES 1.Letter, A.Schwencer to R.E.Uhrig, September 15, 1978, (Docket Nos.50-250 and 50-25,1).2.Margins.in Turkey Point Uni:ts 3 and 4'afety Ana'lysis to Offset the Effects fo Fuel Rod Bowing, FPL Report NAD-gR-25, submitted with letter L-77-106, April 4, 1977.3.Letter L-78-217, R.E.Uhrig to V.Stel,lo, June 22, 1978.4.Letter L-78-230, R.E.Uhrig to V.Stello, July 10, 1978.5.Rosal, E.R.et al., High Pressure Rod.Bundle DNB Data ,with Axially Non-Uniform Heat Flux,.Nuclear Design Vol 31, 1974.
I l    f
0 I 650 640 2400 psia 630-2250 psia 620 2100 psia 610 600 1900 psia 580 570 Note: These curves are applicable with steam generator tube plugging<15 percent, 560 550 540 0 20 40 60 80 100 120 140 RATED PONER (PERCENT)I"igure 2.1-1 Reactor Core Thermal and Hydraulic Safety Limits, 3 Loop Operation 5/21/79
              '%H
                        ~i ea e 4
W C I 't 2                                         2
                                              ~ ~'
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650 640 2400 psia 630 2250 psia 620 2100 psia 610 600 1900 psia 590 580 570 Note-These curves are applicable with steam generator t.ube plugging>15%and<194.560 550 540 20 Figure 2.l-la.40 60 80 100 120 RATED PONER (PERCENT)Reactor Core Thermal,and Hydraulic Safety Limits, 3 Loop Operation 140.5/21/79 Qi 650 640 2400 psia 630 2250'sia 620 2100 psia 610 600 1900 psia 590 580 570 Note: These curves are applicable for steam generator tube plugging>19~and<258 560 550 0 20 Figure 2.l-lb 120 40 60 80 100 RATED POWER (PERCENT)Reactor Core Thermal and Hydraulic Safety, Limits, 3 Loop Operation 140 5/21/79 Qi ig B2.1 BASES FOR SAFETY LIMIT, REACTOR CORE To maintain the integrity of the fuel cladding and,prevent fis-sion product release, it is necessary to prevent overheating of the cladding under all operating conditions.
P.o. BOX 529100, MIAMI, FL 33152 Q@ig L ..i%
This is accomplish-ed by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is very large-and the clad surface temperature is only a few degrees Fahrenheit above the coolant, saturation temperature.
FLORIDA POWER 8 LIGHT COMPANY May 21, 1979
The upper boundary of the nucleate boiling regime is termed departure from nucleate boiling (DNB)and at this point thereis a sharp reduction of the heat transfer coefficient, which would result in high clad temperatures and the possibility of clad failure.DNB is not, however, an observable parameter during reactor operation.
                                                            'L-79-127 Director of Nuclear Reactor Regulation Attention: Mr. Victor Stello,,Jr., Director Division of Operating Reactors
Therefore, the observable parameters; thermal power, reactor coolant temperature and pressure;have been related to DNB through the W-3 DNB correlation.
'U. S. Nuclear Regulatory COIImIission Washington, D.C.      20555
The M-3 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.
 
The local DNB heat flux ratio, (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is'ndicative of the margin to DNB.The minimum value of the,DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.273.This corresponds
==Dear Mr. Stello:==
'to a 95%probability at a 95%con-fidence level that DNB will not occur and is chosen as an ap-priate margin to DNB for all operating conditions.
 
(1,2)The curves in the Specification represent the loci of points of, thermal power, coolant system pressure and'verage temperature for which the DNBR is no less than 1.273.The area of safe opera-.ti.'on is below these lines.The curves are based on the following nuclear hot channel factors: F"=2.~1 q N AH=1.55.B2.1-1 5/21/79 ig  
Re:  Turkey Point Units 3 & 4 Docket Nos. 50-250 and 50-251 Proposed Amendment    to Facility 0  eratin  Licenses  DPR-31 and DPR-41 In accordance wi.th, 10 CFR 50.30, Florida Power '& Light Company (FPL')
~~These limiting hot channel factors are higher than those cal-culated at full power for the range from all contxol rods fully withdxawn to maximum allowable control rod insertion.
submits herewith three (3) signed Originals and forty (40) copies of a request to amend Appendix A of Faci-lity Operating License DPR-31 and DPR-41. The request has been prepared in response to a 'September 15,. 1978 letter .from Mr. A. Schwenc'er of your staff..
The con-trol rod insertion limits are covered by Specification 3~2.Slightly higher hot channel factors could occur at lower power levels because additional control rods are in the core.$Iow-ever, the control rod insertion limits dictated by Figure 3.2-1 ensure that the DNBR is always greater at partial power than at full power.The hot channel factors are also sufficently large to account for the degree of malpositioning of part-length
As  requested in the September 15,letter"(.Reference 1);, FPL has,,performed an  analysis to derive a DNBR safety limit consistent with the COBRA .IIIC computer code. This. code was used'n analysis submitted previously (References 2,3,and 4) to demonstrate the thermal margin available at Turkey Point Units '3 and,4 to compensate for a reduction in 'DNBR due to the effects of fuel rod bowing.
>'rods that is allowed before the reactor trip set points are reduced and rod withdrawal block and load runback may be required.(3)Rod withdrawal block and load runback occur befoxe reactor trip setpoints are reached.The Reactor Control and Protection System is designed to prevent any'nticipated combination of transient.
The NSSS vendor had obtained'      DNBR  limit of 1.24 from an analysis of critical. heat flux tests, using    the  THING computer code with the W-3 critical heat flux correlation      and the L-grid correction.      However, this data  base is. proprietary and is not available to 'FPL. Instead, a corresponding DNBR limit was derived for the COBRA IIIC code by com-paring, the COBRA IIIC results with THING'esults. for a somewhat different data base (Reference 5) which was proposed'n your letter of September
conditions that would result in exceeding DNBR design limits, including the effects of fuel xod bowing.(1)(2)References (1)NAD-2199 (2)NAD-QR'25 (3)FSAR 3.2.2.Any reference to part-length rods'o longer applies after the part-length rods are removed from the reactor.B2.1-2 5/21/79 0 ig I I}}
: 15. The, analysis (attached) yielded a DNBR safety limit of 1.273 for the COBRA IIIC code. Revised "Reactor Core Thermal and Hydraulic Safety Limit " curves for tube plugging levels of <15'A, >15% to <195, and >19/ to <25/, based on the derived DNBR limit of 1.273, are attached.
Revised Bases pages are also attached.
vgpp2 gpss%
A"~ po            v PEOPLE... SERVING PEOPLE
 
4 P Mr.
4 Victor Stello, Jr., Director Division of Operating Reactors U'. S. Nuclear Regulatory 'Commission Page  2 The  information supplied herewith supplements our previous submittals concerning the effects of rod bow on DNBR at Turkey Point Units 3 and
: 4. It is proposed that the attached Reactor Core Safety Limits be substituted for the present Figures 2.1-1,, 2.l-la and 2.1-1b in the Technical Specifications, and that with these new and more restrictive limits, the units can be operated safely without any other penalties for rod bow.
The proposed    amendment has been reviewed by the Turkey Point Plant Nuclear Safety Committee and the Florida Power 8 Light Company Nuclear Review Board. They have determined that    it does not adversely affect the heal,th and safety of the public.
Very  truly yours, Robert E. Uhrig Vice President Advanced Systems 5 Technology REU:MAS:ncr Attachment cc:  'Mr. James P. O'Reilly, Region  II Robert Lowenstein, Esquire
 
14 REFERENCES
: 1. Letter,  A. Schwencer to R. E. Uhrig, September  15, 1978, (Docket Nos. 50-250 and 50-25,1).
: 2. Margins .in Turkey Point Uni:ts 3 and 4'afety  Ana'lysis to Offset the Effects fo Fuel Rod Bowing, FPL Report    NAD-gR-25, submitted with letter L-77-106, April 4, 1977.
: 3. Letter L-78-217,  R. E. Uhrig to V. Stel,lo, June 22, 1978.
: 4. Letter L-78-230,  R. E. Uhrig to V. Stello, July 10, 1978.
: 5. Rosal, E. R. et al., High Pressure Rod .Bundle DNB Data
  ,with Axially Non-Uniform Heat Flux,. Nuclear Design Vol 31, 1974.
 
0 I 650 2400  psia 640
          -2250  psia 630 620        2100  psia 610 1900 psia 600 580 Note:    These curves are 570                      applicable with steam generator tube plugging
                          <15  percent, 560 550 540 0        20      40        60        80      100 120 140 RATED PONER    (PERCENT)
I"igure 2.1-1    Reactor Core Thermal and Hydraulic Safety Limits, 3 Loop Operation    5/21/79
 
650 2400 psia 640 630 2250 psia 620   2100 psia 610 600 1900 psia 590 580 Note-   These curves are 570                  applicable with steam generator t.ube plugging
                    >15% and < 194.
560 550 540 20     40       60     80       100   120   140 RATED PONER (PERCENT)
Figure  2.l-la. Reactor Core Thermal,and Hydraulic Safety Limits, 3 Loop Operation
                                                      . 5/21/79
 
Qi 650 2400   psia 640 630     2250'sia 620     2100   psia 610 600     1900   psia 590 580 Note:     These curves are 570                applicable for steam generator tube plugging
                  >19~ and   < 258 560 550 0       20       40       60     80     100   120  140 RATED POWER   (PERCENT)
Figure  2.l-lb    Reactor Core Thermal and Hydraulic Safety, Limits, 3 Loop Operation 5/21/79
 
Qi ig B2. 1   BASES FOR SAFETY     LIMIT, REACTOR CORE To   maintain the integrity of the fuel cladding and,prevent fis-sion product release, all it is necessary to prevent overheating of operating     conditions. This is accomplish-the cladding    under ed by operating     within the   nucleate   boiling regime of heat transfer, wherein the heat transfer coefficient is very large and the clad surface temperature is only a few degrees Fahrenheit above the coolant, saturation temperature.           The upper boundary of the nucleate boiling regime is termed departure from nucleate boiling (DNB) and at this point thereis a sharp reduction of the heat transfer coefficient, which would result in high clad temperatures and the possibility of clad failure. DNB is not, however, an observable parameter during reactor operation.
Therefore, the observable parameters; thermal power, reactor coolant temperature and pressure; have been related to DNB through the W-3 DNB correlation. The M-3 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.
The local DNB heat flux ratio, (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is'ndicative of the margin to DNB. The minimum value of the,DNBR during steady state operation, normal operational transients,       and   anticipated transients is limited to 1.273. This corresponds 'to     a 95% probability at a 95% con-fidence level that DNB will not occur and is chosen as an ap-priate margin to DNB for all operating conditions. (1,2)
The curves     in the Specification represent the loci of points of, thermal power, coolant system pressure and'verage temperature for which the DNBR is no less than 1.273. The area of safe opera-.
ti.'on is below these lines.
The curves are based on the following nuclear hot channel factors:
F" = 2.~1 q
N AH =   1.55.
B2.1-1                           5/21/79
 
ig
~ ~
These   limiting hot channel factors are higher than those cal-culated at   full power for the range from all contxol rods fully withdxawn to maximum allowable control rod insertion. The con-trol rod insertion limits are covered by Specification 3 2. ~
Slightly higher hot channel factors could occur at lower power levels because additional control rods are in the core. $ Iow-ever, the control rod insertion limits dictated by Figure 3.2-1 ensure that the DNBR is always greater at partial power than at full power.
The hot channel   factors are also sufficently large to account for the   degree of malpositioning of part-length >'rods that is allowed before the reactor trip set points are reduced and rod withdrawal block and load runback may be required. (3)     Rod withdrawal block and load runback occur befoxe reactor trip setpoints are reached.
The Reactor   Control and Protection System is designed to prevent any'nticipated combination of transient. conditions that would result in exceeding DNBR design limits, including the effects of fuel xod bowing. (1) (2)
References (1)   NAD 2199 (2)   NAD QR'   25 (3)   FSAR 3.2.2.
Any reference to part-length rods'o longer applies after the part-length rods are removed from the reactor.
B2.1-2                         5/21/79
 
0 I
ig I}}

Latest revision as of 00:24, 4 February 2020

Submits Analysis Supporting Amend to App a of Licenses DPR-31 & DPR-41.DNBR Safety Limit of 1.24 Is Consistent W/Cobra Iiic Computer Code.Forwards DNBR Safety Limit for Cobra Iiic Analysis.
ML17338A716
Person / Time
Site: Turkey Point  NextEra Energy icon.png
Issue date: 05/21/1979
From: Urhig R
FLORIDA POWER & LIGHT CO.
To: Stello V
Office of Nuclear Reactor Regulation
Shared Package
ML17338A717 List:
References
L-79-127, NUDOCS 7905250426
Download: ML17338A716 (18)
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Text

REGUI.ATTY INFORNATION DISTRIBUTIO SYSTEN (RIBS)

, Af'CESSION NBRi7905250426~ DOC ~ DATE; 79/05/21 NOT RIZEDt NO DOCK FACIL!50"250 TURKEY POLNT PLANTE UNIT 3g 'FLORIDA POWER AND LIGHT C 0500 50 251 TURKEY POINT PLANTg UNIT 4i FLORIDA PO'WER AND "LIGHT C 05000251 AUTH'o NAME AUTHOR AFFILIATION URIGr R ~ E "~ FLORIDA PUWER 8, LIGHT CO.

REC IP ~ NAME RECIPIENT AFF IL'IATION STELLOiV ~ DI VISION OF OPERATING REACTORS SUB JECT: SUBMITS ANALYSIS SUPPORTING AMEND TO APP A'F OL DPR 31 8 PPR 41 ~ 'DETERMINES DNBR SAFETY LIMIT OF 1G24 CONSISTENT W/COBRA IIIC COMPUTER CODK ~ FORWA'RDS "DNBR SAFETY LIMIT FOR COBRA III:C'NALYSIS DISTRIBUTION CODE! A001S:COP IES RECK I VED t LTR +~ ENCL +~ SIZE ~

T'ITLK: GENERAL DISTRIBUTION FOR AFTER ISSUANCE OF OPERATING LIC

,E- C ~ 4 N OT'E S ~ ~~~~~++~<~FR~eirer~~<ree>~~et~~~z+~>~~~~~~>~FR>~reeeeareyere~reeeyeytqe>RN~>we~><yy<yeyyRRNF~<~~~we RECIPIENT COPIES REC IPIFNT COPIES ID LTTR ENCL ID COOK/NAME LTTR ENCL' dR8 & /

"~

CODE/NAME'CTION:

05 BC 7 7 INTERNAI.: OI 02 NRC PDR i

RES 1 1 1 12 ILE 2 2 10 TA/EDO 1 1 15 CORE PERF BR 1 1 16 AD SYS/PROJ 1 1 17 ENGR BR 1 1 18 REAC SFTY SR 1 1 19 'PLANT SYS BR'1 20 KEB 1 EFLT TRT SYS 1 22 BR INKMAN 1 1 EXTERNAL! g3 LPDR 1 1 '04 NS IC 1 1 23 ACRS 16 16 NN 29'@7>

TOTAL NUMBER OF COPIES REQUIRED! LTTR 38 ENCL 38

'J Qi I

~ I E jw I

f

<k IP a 1 l

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W C I 't 2 2

~ ~'

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P.o. BOX 529100, MIAMI, FL 33152 Q@ig L ..i%

FLORIDA POWER 8 LIGHT COMPANY May 21, 1979

'L-79-127 Director of Nuclear Reactor Regulation Attention: Mr. Victor Stello,,Jr., Director Division of Operating Reactors

'U. S. Nuclear Regulatory COIImIission Washington, D.C. 20555

Dear Mr. Stello:

Re: Turkey Point Units 3 & 4 Docket Nos. 50-250 and 50-251 Proposed Amendment to Facility 0 eratin Licenses DPR-31 and DPR-41 In accordance wi.th, 10 CFR 50.30, Florida Power '& Light Company (FPL')

submits herewith three (3) signed Originals and forty (40) copies of a request to amend Appendix A of Faci-lity Operating License DPR-31 and DPR-41. The request has been prepared in response to a 'September 15,. 1978 letter .from Mr. A. Schwenc'er of your staff..

As requested in the September 15,letter"(.Reference 1);, FPL has,,performed an analysis to derive a DNBR safety limit consistent with the COBRA .IIIC computer code. This. code was used'n analysis submitted previously (References 2,3,and 4) to demonstrate the thermal margin available at Turkey Point Units '3 and,4 to compensate for a reduction in 'DNBR due to the effects of fuel rod bowing.

The NSSS vendor had obtained' DNBR limit of 1.24 from an analysis of critical. heat flux tests, using the THING computer code with the W-3 critical heat flux correlation and the L-grid correction. However, this data base is. proprietary and is not available to 'FPL. Instead, a corresponding DNBR limit was derived for the COBRA IIIC code by com-paring, the COBRA IIIC results with THING'esults. for a somewhat different data base (Reference 5) which was proposed'n your letter of September

15. The, analysis (attached) yielded a DNBR safety limit of 1.273 for the COBRA IIIC code. Revised "Reactor Core Thermal and Hydraulic Safety Limit " curves for tube plugging levels of <15'A, >15% to <195, and >19/ to <25/, based on the derived DNBR limit of 1.273, are attached.

Revised Bases pages are also attached.

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4 Victor Stello, Jr., Director Division of Operating Reactors U'. S. Nuclear Regulatory 'Commission Page 2 The information supplied herewith supplements our previous submittals concerning the effects of rod bow on DNBR at Turkey Point Units 3 and

4. It is proposed that the attached Reactor Core Safety Limits be substituted for the present Figures 2.1-1,, 2.l-la and 2.1-1b in the Technical Specifications, and that with these new and more restrictive limits, the units can be operated safely without any other penalties for rod bow.

The proposed amendment has been reviewed by the Turkey Point Plant Nuclear Safety Committee and the Florida Power 8 Light Company Nuclear Review Board. They have determined that it does not adversely affect the heal,th and safety of the public.

Very truly yours, Robert E. Uhrig Vice President Advanced Systems 5 Technology REU:MAS:ncr Attachment cc: 'Mr. James P. O'Reilly, Region II Robert Lowenstein, Esquire

14 REFERENCES

1. Letter, A. Schwencer to R. E. Uhrig, September 15, 1978, (Docket Nos. 50-250 and 50-25,1).
2. Margins .in Turkey Point Uni:ts 3 and 4'afety Ana'lysis to Offset the Effects fo Fuel Rod Bowing, FPL Report NAD-gR-25, submitted with letter L-77-106, April 4, 1977.
3. Letter L-78-217, R. E. Uhrig to V. Stel,lo, June 22, 1978.
4. Letter L-78-230, R. E. Uhrig to V. Stello, July 10, 1978.
5. Rosal, E. R. et al., High Pressure Rod .Bundle DNB Data

,with Axially Non-Uniform Heat Flux,. Nuclear Design Vol 31, 1974.

0 I 650 2400 psia 640

-2250 psia 630 620 2100 psia 610 1900 psia 600 580 Note: These curves are 570 applicable with steam generator tube plugging

<15 percent, 560 550 540 0 20 40 60 80 100 120 140 RATED PONER (PERCENT)

I"igure 2.1-1 Reactor Core Thermal and Hydraulic Safety Limits, 3 Loop Operation 5/21/79

650 2400 psia 640 630 2250 psia 620 2100 psia 610 600 1900 psia 590 580 Note- These curves are 570 applicable with steam generator t.ube plugging

>15% and < 194.

560 550 540 20 40 60 80 100 120 140 RATED PONER (PERCENT)

Figure 2.l-la. Reactor Core Thermal,and Hydraulic Safety Limits, 3 Loop Operation

. 5/21/79

Qi 650 2400 psia 640 630 2250'sia 620 2100 psia 610 600 1900 psia 590 580 Note: These curves are 570 applicable for steam generator tube plugging

>19~ and < 258 560 550 0 20 40 60 80 100 120 140 RATED POWER (PERCENT)

Figure 2.l-lb Reactor Core Thermal and Hydraulic Safety, Limits, 3 Loop Operation 5/21/79

Qi ig B2. 1 BASES FOR SAFETY LIMIT, REACTOR CORE To maintain the integrity of the fuel cladding and,prevent fis-sion product release, all it is necessary to prevent overheating of operating conditions. This is accomplish-the cladding under ed by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is very large and the clad surface temperature is only a few degrees Fahrenheit above the coolant, saturation temperature. The upper boundary of the nucleate boiling regime is termed departure from nucleate boiling (DNB) and at this point thereis a sharp reduction of the heat transfer coefficient, which would result in high clad temperatures and the possibility of clad failure. DNB is not, however, an observable parameter during reactor operation.

Therefore, the observable parameters; thermal power, reactor coolant temperature and pressure; have been related to DNB through the W-3 DNB correlation. The M-3 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.

The local DNB heat flux ratio, (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is'ndicative of the margin to DNB. The minimum value of the,DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.273. This corresponds 'to a 95% probability at a 95% con-fidence level that DNB will not occur and is chosen as an ap-priate margin to DNB for all operating conditions. (1,2)

The curves in the Specification represent the loci of points of, thermal power, coolant system pressure and'verage temperature for which the DNBR is no less than 1.273. The area of safe opera-.

ti.'on is below these lines.

The curves are based on the following nuclear hot channel factors:

F" = 2.~1 q

N AH = 1.55.

B2.1-1 5/21/79

ig

~ ~

These limiting hot channel factors are higher than those cal-culated at full power for the range from all contxol rods fully withdxawn to maximum allowable control rod insertion. The con-trol rod insertion limits are covered by Specification 3 2. ~

Slightly higher hot channel factors could occur at lower power levels because additional control rods are in the core. $ Iow-ever, the control rod insertion limits dictated by Figure 3.2-1 ensure that the DNBR is always greater at partial power than at full power.

The hot channel factors are also sufficently large to account for the degree of malpositioning of part-length >'rods that is allowed before the reactor trip set points are reduced and rod withdrawal block and load runback may be required. (3) Rod withdrawal block and load runback occur befoxe reactor trip setpoints are reached.

The Reactor Control and Protection System is designed to prevent any'nticipated combination of transient. conditions that would result in exceeding DNBR design limits, including the effects of fuel xod bowing. (1) (2)

References (1) NAD 2199 (2) NAD QR' 25 (3) FSAR 3.2.2.

Any reference to part-length rods'o longer applies after the part-length rods are removed from the reactor.

B2.1-2 5/21/79

0 I

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