ML17339A998

Tech Spec Page 3.2-3 & Figure 3.2-3b Reflecting Results of Eecs Analysis on Power Distribution Limits for Steam Generator Tube Plugging Level of 25%.Related Matl Encl

ML17339A998
Person / Time
Site:	Turkey Point
Issue date:	04/29/1980
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17339A996	List: 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 ML17339A998
References
NUDOCS 8005050282
Download: ML17339A998 (41)

Text

800SOSO P F'W reactivity tl:tion upon ejection greater th0.3%

k/k at rate'd power.

Inoperable rod worth shall be determined within 4 weeks.

b.

A control rod shall be considerecl inoperable if (a) t:he rod cannot be moved by the'RDCE, or (b) the rod is misaligned from i.ts banlc by more than 15 inches, or (c) the rod droI> time is not met.

c.

Xf a contxol rod cannot be moved by the drive mechanism, shutdown margin shall be increasecl by boron addition to compensate for the withdrawn wort:h of the inoperable rod.

5.

CONTROl. ROD POSITION INDXCATXON If eit:her t:he powex'ange channel deviation alarm or the rod deviation monitor alarm are not operable rod positions shall be logged once per shift and aft:er a load change greater than 10% of rat:ed power.

IE both alarms are inoperable for tvo hours or more, the nuclear over-power trip.shall be'reset to 93% of rated power.

6.

PONER DISTRIBUTION LIHXTS a.

Hot 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 lov power physics tests:-

I:< (Z) <;(1.97/P) x K(z),. for v >

5 Fq (Z) < (3.94) x K(Z), fox P 5

Ail

< i'55 [l.+0.2 (l-P) ]

lihere P i~s~ 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 Fq.

If P

, as predicted by approved physics calculations, exceeds 1.97, the pover will be limited to the ratecl power multiplied by the ratio of 1.97 divided by the predicted P

, or augmented surveillance of hot channel factors shall be implemented.

(2)

With steam genexator tube plugging <22%, the hot channel fact:ors (clefinecl in the basis

) must meet. the following limits at all times except during lov power physics test:s:

Pci'(Z) '6.(l.'99/P)'

K(Z), for P >.5 I'q (Z)

< (3 98) x K(Z), for I

<.5 F'

l.55

[ l.+0.2 (1-P)]

M>ere 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

C r

c

~

~

HOT CHANNEL FACTOR NORf'1AL IZED OPERATING ENVELOPE (for steam generator tube plugging 25% and Fq=1.97) 1.0 0.9 M~.35 -::...:

0.8 0.7 0.6

~

I

~

~

~

0.5 0.4 0.3 0.2 "0

Bottom CORE HEIGHT (FT.)

10 12 Top

~

~

Figure 3.2-3b

l t

p, r

ATTACHMENT 1 TABLE 1 LARGE BREAK TINE SE(UENCE OF EVENTS START Rx Trip Signal-S.I. Signal Acc. Injection End of Bypass End of BloIidown Bottom of Core Recovery Acc.

Elllpty Pump Injection DECL Cp=0.4 (Sec) 0.0 0.669'.73 I5.5 27.83

29. 12 46.6

59. 67 25.73

TABLE 2 LARGE BREAK Res ul ts Peak Clad Temp. '

Peak Clad Location Ft.

Local Zr/H20 Rxn(max)X Local Zr/H20 Location Ft.

Total Zr/H20 Rxn /

(lot Rod Burst Time sec Hot Rod Burst Location Ft.

DECL 2136 6.0 6.gn5 6.0

<0.3 34.8 6.0

.,Calculation Core Power i&t 102% of Peak Linear Power kw/ft 102% of Peaking Factor Accumulator Water Volume (ft )

3 2200 11.19 1.97 875 (per accum'ulator)

Fuel region + Cycle analyzed Cycle Region Unit 3 and Unit 4 All

'All

TABLE 3 LARGE BREA',

CONTAItll'iENT DATA (DR'f CON fAI'll)EHT)

PET FREE VOLUVi~

1N1 rIAL CONDITIONS Pressure Temperature RllST Temperature Service Hater Temperature Outside Temperature SPRAY SYSTEM Number of'umps Ope) ating Runout F lo bon steel Paint Carbon sLeel Carbon s teel Paint Carbon'teel Concrete Carbon steel Concrete Paint Carbon steel Carbon steel Paint Carbon ste at Carbon steel THICKI'IESS

~INCH 0.006996 0.20 0.006996, 0.006996 0.4896

0. 4896 0.006996 0 2898 24.0 0.2898 24.0 0.006996 1'6

1. 56

.0.006996 5.496 5.496 AREA'(FT

-51824 69 996054.9 35660.11 11886.7 102000.0 34000.0 4622.69 I

1540 89 1277.87 425 93 Paint Carbon steel Carbon steel Paint Carbon s teel Paint Carbon steel Paint Carbon Steel Alluminum Stainless steel Stainless. 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 2.1264 0.1398

.24.0 24.0 951.525 317.175

.23550.0'0368.

5 42278. 25 3.02400. 0 768.0 3704. 0 1439?.0 59132.0'

0 TABLE 4 REFLOOD ASS AND ENERGY RELEASES DECLG (CD = 0.4)

TIME (SEC)

MASS FL011'LB/SEC)

ENERGY FLOlrl

'10 BTU/S EC) 46.597

47. 822

54. 36 64.488 78.288 94.288 111.088 128.688

" 166 '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)

TINE (SEC)

WSS FLO~i (LB)SEC) 0.0 0.01 2.01 4.01 6.01

8. 0 l.

10. 01 15.01 20.01 25'..01

30. 84 31.567 0.0 2820.8 2367.2 2082 '

1879.4 1725.0 1600.2 1369. 6 1215".1 1108.2 1026.4 1017.3

• FOR ENERGY FLOW, MULTIPLY (ASS FLO'A BY AN ENTNALPY OF 59.62 BTU/LB

I.F000 I.2500 TURKEY PDIHT IIUCLEAR POMER STATIOII - 25 SGTP 5

TDr DOUBLE ENDED COLD LEG GUILLOTI4E CO>0.4 oviLITY or rLUID BURSTS' 00 rTI

)

PEAK>

6 00 fT(>l I.0000 o

0.7500

~C c7 0.5000 0.2500 0.0 o

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Figure 1 Fluid quality - DECLG (CD=0.4)

LJ let IIII I

O 50.000 e0.000 20.000

'TURKEY POlkT ¹USLEAR POSER STATlD¹ - 25 SGTP 5

TDF.

DOUBLE E¹DED COLO LEG GUlLLOTl¹E CD 0.4 H ASS V ELOC l T Y DURST.

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. TURKEY POlNT NUCLEAR POVER STAT?Dll - 25 SGTP 5

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To,aoo TURKKY PalRT ROCLKAA POMER STATlalC - 25 SGTP 5

TOF OOUBLE ENOEO COLO LEG GUILLOTlNK CO=0.4 CORK PR.OROP 50.000 (K

25,000 0.0

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Core Pressure Drop - DECLG {CD=0.4)

2S00.0 TURKEY POINT NUCLEAA POUEA STATjojj - 25 SCTP 5

TDF DDUDLE ENDED COLD LEG I"UjLLMlRE.

CD=0 ~ 4 CLAD AYC.TEHP,HOT AOD BURST ~

8 ~ 60 FT(

)

PEAK ~

S.OO FTl<l

~

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a iSOO.O I

a 106G.O C

LJ 500.00 0.0 aa TjjjE <SECj Figure 7

Peak Clad Temperature

- OECLG (CD=0.4)

2000.0 TURKEY POI)(T l(UCLEIcR POMER.STAT[OR - 25 SGTP 5

TOF DOUBLE E)(OEQ COLO LEG CUILLOTIRE Co=0.a FLUID <EHPEAATURE BURST<

6.00 FT(

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PEA)l<

6 ~ 00 FT(I) 1250,0

),00 500.00 250,00 0.0 CI o

C7 TI4E (SEC)

Figure 8

Fluid Temperature - DECLG (CD=0.4)

1000.0 5000.0 TURKEY POiRT NUCLEAR POSER STATION - 23 SGTP 3

TOF OOUBLE EHOCO COLO Lf0 CUlLLOrfNK CO>0e C

l-FLOYRATE CORE BOrrOR ',

t i TOP 2500.0 0,0

-2500.0

-5000.0

-1000.0 Cl CI

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~f CI CI CI IA Figure 9

Core Flo>v - Top and Bottom - DECLG {CDc0.4}

20.000

}7.500 TURKEY POTRT t!UCLEAR POMER STATtOH " ?5 SCTP 5

TOF DOUSLE EROEO COLO LEG GUILLOTlHE CO=0.I UATER LEVEL(FT) 15.001 DOWNCOMER 12.500 10.001

7. 5000 5.00; 2.5000 0.0 C)

C)

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D T)NE LSECl Figure 10a Reffood Transient - DECLG (CD=0.4)

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CD CD CD Fiaure ll, Accumulator Flout (Blowdown} -

DECLG (CD=0.4}

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Figure 12 Pump ECCS Flow (Reflood) - DECLG (CD=0.4)

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Figure 13 Containment Pressure

- DECLG (CD=Oe4) s I

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}.0000 TURKEY PO}RT HVCLEhA POKIER STAT}O" 85 SGTP 5 TOF QOUBLE EAOEO COlD LEG GU}LLOTINE CO=0 ~ I PO'UEA o 0,8000 0,6000 O.l000 Ct O. 0,2000 0.0 o O Dow R T}HE )SEC) Cl CtO Figure 14 Core Power - DECLG {CD=0.4)

5.GO&)l TURKEY POINT NUCLEAR POVER STATION - 25 SGTP 5 TOF ~ 00UDLK EHOEO COl0 IEG CUILIOTIME CO=0. l BREAX ENERCY Ll s.OON) T 3.OOKl07 2.0OEl07 I.OOKED)7 0.0 CI CI CI8 g CI CI Vl TINE (SKC) Figuv'e 15 Break Energy Released To Containment - DECLG (Coc0.4)

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J

Attachment 2 The Nuclear Regulatory Commission (NRC) issued a letter dated November', 1979 to operators of light water reactors reqarding fuel rod models used in Loss of Coolant Accident (LOCA) ECCS evaluation models. That letter describes a meeting called by the NPC on f!ovember 1, 1979 to present draft renort NUREG 0630, "Cladding Swelling and Rupture Models for LOCA Analysis. " At the meeting, representati ves of HSSS 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. llesting-house representatives presented information on the fuel rod models used in analyses for plants licensed with the Westinghouse ECCS eval-uation model and discussed the potential impact of fuel rod model changes on results of those analyses. That information was formally documented. in letter NS-THA-2147, dated November 2, 1979, and formed the basis for the Westinqhouse conclusion that the information was presented in draft NUREG 0630 did not'constitute a safety problem for Westinghouse plants and that all plants conformed with NRC regulations. In the November 9, 1979 letter, the NRC requested that operators of light water reactors provide, within sixty (60) days, information which will enable the staff to determine, in light of the fuel rod model concerns, whether or not further action is necessary. As a result of compili nq information for letter NS-TYiA-2147, Westinghouse 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 HRC staff, bv 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 NRC 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-Tf1A-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 Westinghouse plants with consideration of a modified Westinqhouse fuel rod burst model. As part of this reevaluation, the Westinghouse 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 olants licensed with the Westinqhouse ECCS evaluation model. On December 6, 1979, flRC and Westinghouse personnel discussed the infor-mation thus far presented. At the conclusion of that discussion, the NRC staff requested Westinghouse to provide further detail 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 NS-TNA-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

~g ~ I ~ fuel rod models presented in draft iIUREG-0630 and 2) Acceptable benefits ~ 'resulting from analytical model improvements that would justify continued olant operation for the interim until differences between the fuel rod models of concern are resol ved. Part of the Westinghouse effort 'providedto assist in the resolution of these LOCA fuel rod model differences is documented in letter i)S-TNA-2175, dated December 10, 1979, which contains I'estinqhouse comments on draft NUREG-0630. As stated in that letter, l~estinghouse believes the current Westinghouse models to be conservative and to be in compliance with Apoendi x K.

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 5 4 with 25/ SGTP and 55 red. TOF. This evaluation is based on the l.imiting break LOCA analysis identi-fied as follov(s: BREAK TYPE DOUBLE ENDED COLD LEG GUILLOTINE BREAK DISCHARGE COEFFICIENT CD=0.4 MESTINGHOUSEECCS EVALUATION MODEL VERSION Februar, 1978 CORE PEAKING FACTOR 1 ~ 97 HOT ROD HAXIl1UYi TEMPERATURE CALCULATED FOR T!IE BURST REGION OF ThE CLAD - 2136 OF = PCTB ELEVATION - 6.0 Feet. HOT ROD MAXINUYi TEMPERATURE CALCULATED FOR A NON-RUPTURED REGION OF THE CLAD 1976 OF = PCT N ELECTION -

7. 75 Feet CLAD STRAIN DURING BLO!JDONN AT THIS ELEVATION 4.00 Percent IlAXINUiMCLAD STRAIN AT THIS ELEVATIONl 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 ASSEtlBLY 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-Tl1A-2174 in terms of the change in the peakina factor limit (Fg) reauired 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) indicated here may not accuratelv apoly over large rangesz

but a simultaneous change in FQ which causes the PCT to remain in the neighborhood of 2200.'F justifies use of this evaluation procedure. From NS-TNA-2174: For the Burst Node of the clad: 0.01 (6FQ -> ~ 150 F BURST NODE aPCT Use of the NRC burst model and the revised westinghouse burst model could require an Fq reduction of 0.027 The maximum estimated impact of using the NRC strain model is a required FQ: reduction of 0.03. Therefore, the maximum penalty for the,Hot Rod burst node is: APCT1 = (.027 +.03) (150'/.01) = 855'F 'Margin to the 2200. F limit is: 6PCT2 = 2200.;.F-PCT0 = 66' The EQ reduction required to maintain the 2200'F clad temperature 'limit is: a600 = (6PCT) 6PCT2) (~f~g F~) = (B55 - 64) (;50) = 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 taro aspects of the'nalyses. The first aspect is the change in pellet-c)ad 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 change the time at Nhich brust is calculated. Three'sets of LOCA analysis results were'tudieQ.to establi'shed'an.ac'ceptable"sens'it'ivity to apply gene'rically in this evalution. The possible.PCT increase" resulting from a change in: strain (in the llot Rod) is +20. F. per percent decrease in strain at the maximum clad temperature

~ ) 4~

locations. Since the clad strain calculated during the reactor coolant syst ~ blowdown phase of the,accide is not changed by the use of f fuel rod models, th6 maximu crease ',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: hPCT3 = ( 01 t ) (MAX STRAIN - BLOl<00<lH STRAIN} 20 F ( py ) (g. 0852-0. 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 th iNRC 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 llS-Tf<<A-2174. 'I Ther efore, hPCT4 = 1 25oF (50 PERCENT CURREHT BLOCKAGE) + 2.36oF (75-50) = 1.25 (50 - 0.0 ) + 2.36 (75-50) 121:5'F 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 + hPCT4 = gQ 4 + iQ = 90 4 Hargin to the 2203oF limit is hPCT6

2200oF - PCT(l

224'-~F The.FQ reduction required to maintain this 2200oF clad tem-perature.limit is (from NS-THA-2174) hFQ = (hPCT - hPCT. ) (. ) =-.<~4-. ~ 10F hPCT hFQ< andh'Q><, '~PEII TY B. The effect on LOCA analysis results of using improved analytical and modeling techniq es (which are currently approved for use in the Upper ffead Injection plant LOCA analyses) in.the reactor coolant system bio:idown calcula ion (SATAsl computer code) has been quanti-fied via an analysis which has recently'been submitted to the i<RC for review. Recognizing that review of that analysis is not yet complete and that the benefits associated with those model improve-ments can change for 'other 'plant designs, the llRC has established a credit that is acceptable for'his interim period to help offset penalties resulting fron application of the VRC fuel rod mod ls. That credit for two, three and four loop plants is an increase in'he LOCA peaking factor limit of 0.12, 0.15 and 0.20 respectively. The peaking factor limit adjustment required ta justify plant. operation for this in.crim period is determined as te appro".riate 5FQ cr'edit identified in section (B) above, minus trek FQ calculated in section (A) above (but not greater then seri>P."'"'" Fil ADJUsTviEnT = 0.15

0.053 P h h 4 Mr~

COUNTY OF DADE STATE OF FLORIDA . ) ) ) 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 day of 19@a NOTARY PUBLIC, in and for the county of Dade, State of Florida NOTARY PVBUC STATE OF FLORIOA st LARGIE MY COMMISSION EXPIRES AUGUST 24, 155 Ny commission expires. BONOEO THRU 51AYNARO RONDINO

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