L-80-129, Application for Amend to Licenses DPR-31 & DPR-41 Revising Tech Specs to Reflect Results of ECCS Analysis on Power Distribution Limits for Steam Generator Tube Plugging Level of 25%.Supplements 800213 Amend Request
| ML17339A995 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 04/29/1980 |
| From: | Robert E. Uhrig FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML17339A996 | List: |
| References | |
| L-80-129, NUDOCS 8005050279 | |
| Download: ML17339A995 (55) | |
Text
REGULATORY INFORMATION DISTRIBUTION SYSTEr'r (BIOS)
ACCESSION NRR:8005050279 DOC CHOATE; 80/04/29 NOTARIZED; NO DOCKET FACIL:50es 50 Turkey Point PlantP Unit 3r Florida Power and Light C
05000250 AUT
~ N4~
~
~
$~251 Turkey Point Pl antE Unit 4E Florida Power'nd'ight C
05000251 AUTHOR AFFILIATION OHRIGPR,ED Florida Power 8 Light Co, REC IP ~ NAME RECIPIENT AFFILIATION
SUBJECT:
Application for amend to License OPR"31 E
OPR"41 revising Tech Specs to reflect r esul ts of ECCS analysis on po~er distribution 1 imi ts for steam generator tube plugging level of 25/ ~ Supplements 800213 amend reouest
~
DISTRIBUTION CODE:
ASSIS COPIES RECEIVED:LTR + ENCL SIZE:
TIJLE: General Distrfoution for after Issuance of Operating Lic NOTES:
RECIPIENT IO CODE/NAME ACTION:
05 BC OR8 ~i INTERNAL~
Ol G FILE 12 I 17 ENGR RR 19 PLANT SYS BR'1 EFLT TRT SYS OELD EXTERNAL: 03 LPDR 23 ACRS COPIES LTTR ENCL 7
7 1
1 2
2 1
1 1
1 1
1 0
1 1
16 16 RECIPIENT IO COOE/NA~>E 02 NRC POR 15 CORE PERF BR 18 REAC SFTY BR 20 EEB EPB DOR STS GROUP LEADR 04 NSIC COPIES LTTR ENCL 1
1 1
1 1
1 t
1 1
TOTAL NUMBER OF COPIES REQUIRED:
LTTR 38 ENCL 37
t P.O, SOX 529100, MIAMI,FL 33152 8 OOgOg O FLORIDA POWER d, LIGHT COMPANY April 29, 1980 L-80-129 Director of Nuclear Reactor Regulation Attention:
Mr. Darrell G. Eisenhut, Acting Director Division of Operating Reactors U.
S.
Nuclear Regulatory Commission Washington, D.
C.
20555
Dear Mr. Eisenhut:
Re:
Turkey Point Units 3 and 4
Docket Nos.
50-250 and 50-251 Proposed Amendment to Facility 0 eratin Licenses DPR-31 and DPR-41 In accordance with 10 CFR 50.30, Florida Power 8
Light Company submits herewith three (3) signed originals and forty (40) copies of a
request to amend Appendix A of Facility Operating Licenses DPR-31 and DPR-41.
This request supplements our amendment request 'of February 13, 1980 (L-80-51).
Our NSSS vendor (Westinghouse) has completed a
revised ECCS analysis for Turkey Point Units 3 and 4.
- February, 1978 Appendix K evaluation models were used for the worst DECLG break (C>=0.4),
assuming-a steam generator plugging level of 25$,
a 5%%d reduction in thermal design flow, and removal of a 65'F fuel temperature conservatism.
The limiting break was reanalyzed at an Fq of 1.97 with a resulting peak clad temperature of 2136'F.
The results of the analysis are presented in Attachment 1.
Our NSSS vendor is currently investigating the impact on.LOCA evaluation results of fuel rod models proposed by the NRC in draft NUREG-0630.
Compensation for possible penalties from use of these fuel rod models has been demonstrated by available improvements in the ECCS evaluation model (See ).
Until final resolution of the overall fuel rod model
- concern, the procedure used to perform this analysis is believed to be suitably conservative and acceptable.
During our review of the ECCS analysis for'255 steam generator tube plugging, it was determined that recent in-containment structural modifications were not explicitly included in the structural heat sink data table (Table 3
of ).
We have evaluated the effect of the structural modifications (0.55 increase in the total amount of steel in containment) and determined that this will have an insignificant effect on the calculated containment backpressure and subsequently on the peak clad temperature, and that there is peak clad temperature margin available to cover this effect.
Accordingly, Florida Power
& Light concludes that the attached 25%%d analysis and the 22%%d analysis submitted on February 13, 1980 remain valid for the current heat sink configuration in both the Unit 3 and Unit 4 contaiIments.
of 5
PEOPLE SERVING PEOPLE
'fa
Director of Nuclear Reactor Regulation Attention:
Mr. Darrell G. Eisenhut, Acting Director Page 2
The proposed amendment is described below and shown on the acccmpanyi ng Technical Specification pages bearing the date of this letter in the lower right hand corner.
Pa e 3.2-3 Specification 3.2.6.a (I) is revised to reflect the results of a
recent ECCS analysis'n the power distribution limits for a
steam generator tube plugging level of 255.
Fi ure 3.2-3b The normalized hot channel factor operating envelope for a
steam generator tube plugging level of 25K is revised to reflect the results of a recent ECCS analysis.
Very truly yours, Robert E. Uhrig Vice President Advanced Systems 8 Technology REU/MAS/cph Attachment cc:
Mr. James P. O'Reilly, Region II Harold F. Reis, Esquire
~ 80050go
- g. P'~
reactivity insertion upon ejection greater than 0.3%
t /k at rated power.
Inoperable rod worth shall be determined within 4 weel s.
b.
A control rod shall be considered inoperable if (a) the rod cannot be moved by the GRDCE, or (b) the rod is misaligned from its banlc by more than 15 inches, or (c) the rod drop time is not met.
c 4
If a control rod cannot be moved by the drive mechanism, shutdown margin shall be increased by boron addition to compensate for the withdrawn worth of the inoperable rod.
5.
CONTROL ROD POSITION INDXCATXON Xf either the power range channel deviation alarm or the rod deviation monitor alarm are not operable rod positions shall be logged once per shift and after a load change greater than 10% of rated power. If both alarms are inoperable for two hour. or more, the nuclear over-power trip shall be reset to 93% of rated power.
6.
POWER DISTRIBUTION LXHXTS a.
Tarot channel factors:
(1)
With steam generator tube plugging
>22% and
<25%,
the hot channel factors (defined in the basis) must meet the following limits at all times except during low power physics tests:
Fq (Z)
< '(p 97/P) x K(Z)', for 'P >'5 Fq (Z) < (3.94) x K(Z), for P
<.5
.F~
< I 55 D-+0-2 (1 P) l li%ere P is the fraction of rated power at which the, core is operating; K(Z) is the function given in Figure 3. 2-3b; Z
is the core height location of F q'f F
, as predicted by approved physics calculations, exceeds 1.97, the power will be limited to the rated power multiplied by the ratio of 1.97 divided by the predicted F
, or augmented surveillance of hot channel factors shall be imflemented.
(2)
With steam generator tube plugging -22%, the hot channel factors (defined in the basis
) must meet the following limits at all times except during low power physics tests:
F<'(Z) '~.(1.99/P)'
Z(Z), for P >.5 F< (Z)
< (3.98) x K(Z), for P <.5 F<ri ~ 1'55
~ 1 +0 2 (1 P Where P is the fraction of rated power at which the core is operating; K(Z) is the function given in Figure 3.2-3a; Z
is the core height location of Fq.
3 ~ 2 3
I
HOT CHANNEL FACTOR NORfSLIZED OPERATING ENYELOPE (for steam generator tube plugging 25K and Fq=1.97)
. 1.0 0.9
'35 0.8 0.7 0.6 0.5 0.4 0.3 0.2 0
Bottom CORE HEIGHT (FT.)
10 12 Top Figure 3.2-3b
t'
ATTACHt7IENT 1 TABLE 1
LARGE BREAK TIt1E SE(UENCE Of EVENTS START Rx Trip Signal S.I. Signal Acc. Injection End of Bypass End of Blowdovin Bottom of Core Recovery Acc.
Empty Pump Injection DECL CD=O.4 (Sec) 0.0 0.669 0.73 15.5 27.83
- 29. 12 46.6 59.67 25.73
TABLE 2 LARGE BREAK Results Peak Clad Temp. '
Peak Clad Location Ft.
Local Zr/H20 Rxn(max)X Local Zr/H20 Location Ft.
Total Zr/H20 Rxn 5 Hot Rod Burst Time sec Hot Rod Burst Location Ft.
DECL 2136 6.0 6.945 6.0
<0.3 34.8 6.0
.. Calculation Core Power 1%t 102K of Peak Linear Power kw/ft 1025 of Peaking Factor Accumulator Water Volume (ft )
3 2200
- 11. 19 1.97 875 (per accumulator)
Fuel region + Cycle analyzed Cycle Region Unit 3 and Unit 4 Al 1 All
TABLE 3 LARGE BREAK CONTAIfli'i NT OATA {DRY CONTAIsll'1EHT)
NET FREE VOLUViE INITIALCONDITIONS Pressure Temperature RHST Temperature Service Hater Temperature Outside Temp rature SPRAY SYSTEH Number of Pumps Operating Runout Flora Pate Actuation Time
- 1. 55x10 Ft 14.7 psl a 90 F
3g F
63
'F 39 "F
2 1050
. gpm 26 secs SAFFGUAROS FAN COOLE)S Number of Fan Coolers Operati.ng Fastest Post Accident Initation of Fan Coo1ers 26 se.cs
E OilTAIfliiEi'IT DATA (DRY COlITAItliIEPPf STPllCTURAL HEAT S IlII:S Paint
'Carbon steel Carbon steel Paint Carbon steel Carbon s teel Paint Carbon steel Concrete Carbon steel Concrete Paint Carbon steel Carbon steel Paint Calboil steel Carbon steel
~ ~
THICKI'IESS
~INC(I) 0.006996 0.20 0.006996 0.006996 0.4896 0.4896 0.006996 0.2898 24.0 0.2898 24.0 0.006996 1.56
- l. 56
.0.006996 5.496 5.496 AREA'[FT 51824'-'69 996054.9 35660.11 11886.7 102000.0 34000.0 4622.69 1540.89 1277.87 425.93'aint Carbon steel Carbon steel Paint Carbon steel Paint Carbon steel Paint Carbon s'teel Alluminum Stainless steel Stainl ess. steel Stainless steel Concrete Concrete 0.006996
. 2.748 2.748 0.006996 0.03 0.006996 0.063 0.006996 0.10 0.006996 0.4404 2.1?64
- 0. 1398
. 24.0
- 24. 0 951.525 317.175
.23550.0'0368.
5 427?8.25 102400.0 768;0 3704. 0 14392.0 59132.0
TABLE 4 REFLOOD MASS AND ENERGY RELEASES - DECLG (CD = 0.4)
TrXE (SEC)
MASS FLOW'(LB/SEC)
ENERGY FLOW (10 BTU/SEC) 46.597 47.822 54.36 64 '88 78.288 94.288 111.088 128 '88
"'66'88 208 588 255.688 0.0 0.0245 34.06 77.45 82.3 100.5 250.8 276.8 285.4 292.7 300.6 0.0 0.003 0.4418 0.9665 1.025 1.131
- 1. 514 1.535 1.453
- 1. 360 1.249
TABLE 5 Broken Loop Accumulator Flow To Containment For Limiting Case Declg (CD = 0.4)
TIWE (SEC) tOSS FLOW (LB/SEC) 0.0 0.01 2.01 4.01 6.01
- 8. 01, 10.01 15
~ 01
- 20. 01 25.01 30.84 31.567 0.0 2820.8 2367.2 2082.2 1879 '
1725.0 1600.2 1369.6 1215.1 1108 '
1026.4 1017
~ 3
- FOR ENERGY FLOW, MULTIPLY MASS FLOW BY AN ENTHALPY OF 59.62 BTU/LB
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Attachment 2
The fluclear Regulatory Commission (NRC) issued a letter dated November 9,
1979 to operators of light water reactors regardinq fuel rod models used in Loss of Coolant Accident (LOCA)
ECCS evaluation models.
That letter describes a meeting called by the NPC on flovember 1, 1979 to present draft reoort HUREG 0630, "Cladding Swelling and Rupture Models for LOCA Analysis."
At the meeting, representatives of NSSS vendors and fuel suppliers were asked to show how plants licensed using their LOCA/ECCS evaluation model continued to conform to 10 CFR Part 50-46 in view of the new fuel rod models presented in draft NUREG 0630.
westing-house representatives presented information on the fuel rod models used in analyses for plants licensed with the Mestinghouse ECCS eval-uation model and discussed the potential impact of fuel rod model changes on results of those analyses.
That information vias formally documented.
in letter NS-TMA-2147, dated November 2, 1979, and formed the basis for the l(estinqhouse conclusion that the information was presented in draft NUREG 0630 did not constitute a safety problem for lJestingnouse plants and that all plants conformed with NRC regulations.
In the November 9, 1979 letter, the HRC requested that operators of light water reactors provide, within sixty (60) days, information which wi 11 enable the staff to determine, in light of the fuel rod model concerns, whether or not further action is necessary.
As a result of compilinq information for letter HS-TYiA-2147, t!estinghouse recognized a potential discrepancy in the calculation of fuel rod burst for cases having clad heatup rates (prior to rupture) siqnificantly lower than 25 deqrees F per second.
This issue was reported to the NRC staff, by telephone, on November 9, 1979, and although independent of the NRC fuel rod model
- concern, the combined effect of this issue and the effect of the HRC fuel rod models had to be studied.
Details of the work done on this issue were presented to the NRC on November 13, 1969 and documented in letter NS-TtlA-2163 dated November 16, 1979.
That work included development of a procedure to determine the clad heatup rate prior to burst and a reevaluation of operatinq Hestinghouse plants with consideration of a modified Westinqhouse fuel rod burst model.
As part of this reevaluation, the ltestinghouse position on NUREG-0630 was reviewed and it was still concluded that the information presented in draft NUREG-0630 did not constitute a safety problem for plants licensed with the l/estinqhouse ECCS evaluation model.
On December 6, 1979, HRC and Westinghouse personnel discussed the infor-mation thus far presented.
At the conclusion of that discussion, the HRC staff requested Hestinghouse to provide further detai 1
on the poten-tial impact of modifications to each of the fuel rod models used in the LOCA analysis and to outline analytical model improvements in other parts of the analysis and the potential benefit associated with those improvements.
This additional information was compiled from various LOCA analysis results and documented in letter HS-TMA-2174 dated December 7, 1979.
Another meeting was held in Bethesda on December 20, 1979 where NRC and Westinghouse personnel established:
- 1) The currently accepted procedure for assessing the potential impact on LOCA analysis results of using the
fuel rod models presented in draft fIUREG-0630 and 2) Acceptable benefits resulting from analvtical model improvements that would justify continued plant operation for the interim until differences between the fuel rod models of concern are resolved.
Part of the llestinghouse effort 'providedto assist in the resolution of these LOCA fuel rod model differences is documented in letter AS-TNA-2175, dated December 10, 1979, which contains Lestinqhouse comments on draft NUREG-0630.
As stated in that letter, t!estinghouse believes the current Westinghouse models to be conservative and to be. in compliance with Apoendix K.
~
+'
~
A.
Evaluation of the potential impact of using fuel rod models pre-sented in draft NUREG-0630 on the Loss of Coolant Accident (LOCA) analysis for Turke Point units 3
8 4 with 25%
SGTP and 5% red. TDF.
This evaluation is based on the limiting break LOCA analysis identi-fied as follows:
BREAK TYPE -
DOUBLE El'IDED COLD LEG GUILLOTINE BREAK DISCHARGE COEI.FI CIEtlT CD=0. 4 MESTINGHOUSE ECCS EVALUATION MODEL VERS IOt'I Februar, 1978 CORE PEAKING FACTOR 1.97 HOT ROD MAXItlUMTEtlPERATURE CALCULATED FOR THE BURST RFGION OF ThE CLAD -
2136 OF = PCT B
ELEVATION -
6.0 Feet.
IIOT ROD MAXItiIUYiTEtlPERATURE CALCULATED FOR A tlON-RUPTURED REGION OF THE CLAD 1976 OF
=
PCTN ELEVATION -
- 7. 75 Feet CLAD STRAIN DURIiNG BLO!IDOIIN AT THIS ELEVATION 4.00 Percent tIAXIMUM CLAD STRAIN AT THIS ELEVATIOII 8.52 Percent maximum temperature for this non-burst node occurs when the core reflood rate is GREATER than 1.0 inch per second and reflood heat transfer is based on the FLECHT calculation.
AVERAGE HOT ASSEt!BLY ROD BURST ELEVATION -
N/A HOT ASSEMBLY BLOCKAGE CALCULATED -
0.0 1.
BURST NODE Percent Feet The maximum potential imoact on the ruptured clad node is expressed in letter NS-TIIA-2174 in terms of the change in the peakino factor limit (Fg) reouired to maintain a peak clad tem-perature (PCT) of 2200 F and in terms of a change in PCT at a
constant Fg:
Since the clad-water reaction rate increases sig-nificantly at temperatures above 2200 F, individual effects (such as aPCT due to changes in several fuel rod models) indi cated here may not accuratelv apoly over large ranges, but a simultaneous change in Fg )~hich causes the PCT to remain in the neighborhood of 2200.
F justifies use of this evaluation procedure.
From NS-T!1A-2174:
For the Burst Node of the clad:
0.01 hFg -
150'F BURST NODE a,PCT Use of the NRC burst model and the revised llestinghouse burst model could require an Fq reduction of 0'.027 The maximum estimated impact of using the NRC strain model is a required Fg; reduction of 0.03.
Therefore, the maximum penalty for the Hot Rod burst node is:
hPCT1
=
.027 +.03) (150'/.01)
= 855'F
'argin to the 2200.
F limit is:
a,PCT2
2200.',.F-PCTB
64 F
The Fg reduction required to maintain the 2200' clad temperature limit is:
LE/~ = (nPCT)
APCT~) (~f~g p
)
- 01 (855
- 64) (150)
= 0.053 (but not less than zero).
2.
NON-BURST NODE The maximum temperature calculated for a non-burst section of clad typicall'y occurs at an elevation above the core mid-plane during the core reflood phase of the LOCA transient.
The poten-tial impact on that maximum clad temperature of using the NRC fuel rod models can be estimated by examining t>vo aspects of the'nalyses.
The first aspect is the change in pellet-clad gap conductance resulting from a difference in clad strain at the non-burst maximum clad temperature node elevation.
Note that clad strain all along the fuel rod stops after clad burst occurs and use of a different clad burst model can chanae the time at which brust is calculated.
Three sets of LOCA analysis results were'tudied.to established'anac'ceptable"sens'itivity to apply gene'rically in this evalution. 'he possible.PCT increase'esulting from a change in: strain (in the )lot Rod) is +20.
F.
per percent decrease in strain at the maximum clad temperature
r
locations Since the clad strain calcula during the reactor coolant s
em blowdown phase of the,acci is not changed by the use o
HRC fuel rod models, th(5 maximum decrease
'.n clad strain that must be considered here is the difference between the "maximum clad strain",and the "clad strain at the end of RCS blowdown" indicated above.
Therefore:
hPCT
= (
)
(MAX STRAIN - BLOWOOHH STRAIN)
( py ) (~0S52-0. 04 )
20 90.4 The second aspect of the analysis that can increase PCT is the flow blockage calculated.
Since the greatest value of blockage indicated by the NRC blockage model is 75 percent, the maximum PCT increase can be estimated by assuming that the current level of blockage in th analysis (indicated above) is raised to 75 percent and then applying an appropriate sensitivity formula shown in HS-TflA-2174.
Therefore, hPCT4
= 1.25 F (50 - PERCENT CURRENT BLOCKAGE)
+ 2.36oF (75-50)
= 1.25 (50 - 0.0
) + 2.36 (35-50) 121.5 oF If PCTH occurs when the core reflood rate is greater than 1.0 inch per second hPCT4
= 0.
The total potential PCT increase for the non-burst node is then hPCT5
= hPCT3 + IPCT4 = gP,4 + 0.= 'gP 4 Hargin to the 2200oF limit is hPCT6
2200oF - PCTH
224-',A F The.Fg reduction required to maintain this 2200oF clad tem-perature limit is (from HS-THA-2174) b,Fg
= (BPCT
-'hPCT.
)
(
) =-.~3~.
I.o'F aPCT but not less than zero.
The peaking factor reduction required to maintain the 2200 'F clad temperature limit is therefore the greater of LFg> andh~g><,
~ nPEtlA'Y 8.
The effect on LOCA analysis results of'sing improved analytical and modeling techniques (v:hich are currently approved for use in the Upper ffead Injection p'lant LOCA analyses) in.the reactor coolant system bio:adown calculation (SATAia computer code) has been quanti-fied via an analysis vhich has recently been submitted to the HRC for review(.
Recognizing that revien of that analysis is not yet complete and that the benefits associat d:cith those model improve-ments can change for other plant designs, the fsRC has established a
credit that is acceptable for'his interim period to help offset penalties resulting from application of the VRC fuel rod models.
That credit for tv.o, three and four loop plants is an increase in" the LOCA peaking factor limit of 0.12, 0.15 and 0.20 respectively.
The peaking factor limit adjustmeni requir d to justify plant operation for this in crim period is determined as the appropriate bFQ credit icer.",i=ied in section (5) above s:inus thebFQ
} t v calculated in.ection (h) above (but not greater than zero>.'"
'Q ADJUSTE}iT= 0.15
0.053
STATE OF FLORIDA ) ) COUNTY OF DADE ) ss. Robert E. Uhrig, being first duly sworn, deposes and says: That he is a Vice President of Florida Power & Light Company, the Licensee herein; That he has executed the foregoing document; that the state-ments made in this said 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. Robert E. Uhrig Subscribed and sworn to before me this ~g~ day of l9+a NOTARY PUBLIC, in and for the county of Dade, State of Florida hOTARY PUBVC STATS OF FLOf"3A St LARGC MY COMMlSS!Otl EXPIPES AVOOST 24, t9El Hy commi 8 s ion expires: sottoFo Ttgv RMvttARO so::otRa AINE
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