ML17340B410

Forwards Interim Annual Operating Rept for 1979.

ML17340B410
Person / Time
Site:	Hatch
Issue date:	02/28/1980
From:	Manry M GEORGIA POWER CO.
To:	James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
Shared Package
ML17340B411	List: ML17340B410 ML19344D269
References
PM-80-182, NUDOCS 8003110615
Download: ML17340B410 (8)

Text

. BOX 529100, MIAMI,FL 33152 y+Q I Idt/~

PAls '4%%

FLORIOA POWER Ik LIGHT COMPANY March 5, 1980 L-80-74 Office of Nuclear Reactor Regulation Attention: Nr. Darrell G. Eisenhut Acting Director.

Division of Operating Reactors U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Dear Nr. Eisenhut:

Re: Turkey .Point Units 3 and 4 Docket Nos. 50-250 and 50-251 This letter provides additional information in support of Florida Power

& Light Company's request for an amendment to Operating Licenses DPR-31 and DPR-41 dated February 13, 1980 (L-51-80).

After correcting input errors reported in Reportable Occurrence 250-79-33 (Letter PRN-LI-79-414 of November 15, 1979) and correcting a potential nonconservatism reported in Reportable Occurrence 250-79-33 (Letter PRN-LI-79-423 of November 23, 1979), our NSSS vendor, Westinghouse, performed an ECCS analysis for Turkey Point Units 3 and 4 with the February 1978 Appendix K evaluation model for the worst break (CD=0.4), assuming a steam generator tube plugging level. of 22Ã. This ECCS analysis yielded an allowable F of 1.89.

A second ECCS analysis for the same conditions but with the removal of a 65GF fuel temperature conservatism led to an Fq of 1.99, thus increasing the allowable F limit by 0. 10. This analysis was transmitted to you with our letter L-51-80 dated February 13, 1980. We understand that this particular model change is still under review and will 'be acted on shortly.

In the attachment and in our letter L-80-34 of January 23, 1980, it is shown that for Turkey Point the F penalty associated with a revised clad burst and flow blockage mode) is 0.05. However, there is an F<

credit of 0.15 available for technical changes derived from upper head injection investigations.

To suIIIIarize, the allowable Fq limit for Turkey Point with -224 tube plugging is as follows:

8003 yg0 gy g PEOPLE... SERVING PEOPLE

Office of Nuclear Reactor Regulation Page Two February 1978 Appendix K model 1.89 Clad burst and flow blockage penalty -0 '5 Upper head injection changes credit +0.15 Allowable Fq limit l. 99 Credit for 65op fuel temperature conservatism +0.10 All'owable Fq limit 2.09 We request, on the basis of these analyses, that you approve our submittal of February 13, 1980. for a change in Technical Specifications to an Fq limit of 1.99. When. your review of the removal of the 65oF fuel temperature conservatism has been compl'eted, the Fq limit for Turkey Point with 22K tube plugging should be increased to 2.09.

Very truly yours, Robert E. Uhrig Vice President Advanced Systems 8 Technology REU/GDW/ah Attachment cc: J. P. O'Reilly, Region II Harold F. Reis, Esquire

Attachment The Nuclear Regulatory Commission (NRC) issued a letter dated November 9, 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 HRC on November 1 1979 to present draft ressort HUREG 0630, "Cladding Swelling and Rupture Models for LOCA Analysis." At the meeting, representatives 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 HUREG 0630. Westing-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-TMA-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 Westingnouse plants and that all plants conformed with HRC regulations. In.,the November 9, 1979 letter, the HRC requested that operators of light water reactors provide, within sixty (60) days, information which will enable the staff to determine,'n light of the fuel rod model concerns, whether or not further action is necessary.

As a result of compilina information for letter HS-TYiA-2147, Westinghouse recognized a potential discrepancy in the calculation of fuel rod burst for cases having clad heatup rates (prior to rupture) significantly lower than 25 deqrees F per second. This issue was reported to the NRC staff, by telephone, on Hovember 9, 1979, and although independent of the HRC'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 HS-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 nlants licensed with the Westinqhouse 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 Westinghouse, to provide further detail on the poten-tial impact of modifications to each of the fuel'od models used in the LOCA analysis and to outline analytical model improvements in other Darts 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

I fuel rod models present n draft HUREG-0630 and 2) Accable benefits ti res ul ng from anal yti ca model i mprovements that would j ti fy us continued ol ant ooerati on for the i nteri m unti l differences between the fuel rod models of concern are resolved .

Part of the ltestinghouse effort provided to assist in the resolution of these LOCA fuel,rod model differences is documented in letter HS-THA-2175, dated December 10, 1979, which contains llestinqhouse comments on draft NUREG-0630. As stated in that letter, Mestinghouse believes the current llestinghouse models to be conservative and to be in compliance with Appendix K.'

A. Evaluation of the po.-ntial impact of using fuel rod ,odels pre-8"k~ii sented in draft NUREG-0630 on the Loss of Coolant Accident (LOCA) conservati sm removed.

54 "ih2H ., F d This evaluation is based on the limiting break LOCA ana>lysis identi-fied .is follows".

BREAK TYPE DOUBLE ENDED COLD LEG GUILLOTINE BREAK DISCHARGE COEFFICIENT CD=0.4 WESTINGHOUSE ECCS EVALUATION tiODEI VERSION Februar, 1978 CORE'EAKING FACTOR 1.99 HOT ROD tlAXINUH-TEt1PERATURE CALCULATED FOR THE BURST REGION OF THE L ~gQ F = PCT B

ELEVATION 6.0 Feet.

!IOT ROD MAXIMUM TEMPERATURE CALCULATED FOR A NON-RUPTURED REGION OF

, THE CLAD- 'F = PCT N

ELEVATION 7.75 Feet CLAD STRAIN DURING BLO!<DO!JN AT THIS ELEVATION 4.29 Percent HAXIiviUbl CLAD STRAIN AT THIS ELEVATION - 8.92 Percent Maximum temperature for this non-burst node occurs when the core ref lood rate is. greater 'han 1.0 inch per second and ref lood heat transfer is based'n'the ,FLECHT calculation.

AVERAGE,'HOT ASSEMBLY ROD BURST .ELEVATION - N/A Feet HOT ASSEMBLY BLOCKAGE CALCULATED - '.0 Percent

1. 'BURST NODE The maximum potential imoact on the ruptured clad node is expressed in letter NS-TNA-2174 in terms of the change in the

',peakina .factor limit (Fg) reauired to maintain a peak caid 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 sPCT due to changes in several fuel rod models) indicated .here may not accuratelv apply over large ranges, but a simultaneous change in Fg which causes the PCT to remain in the neighborhood of 2200. F justifies use of this evaluation'rocedure.

From NS-TMA-2174:

For the Burst Node of the clad:

0.01 hFg ~ ~ 150'F BURST NODE aPCT Use of the NRC burst model and the revised Westinghouse burst model 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:

BPCTl 027 + 03 150 F/ 01 855 P Margin to .the 2200.', F limit is:

hPCT2 = 2200.',.F- PCTB = 40 F The Fg reduction required to maintain the 2200' clad temperature limit is:

= (aPCT - .01 aF ')

aFg APCT2) ('150 F

01

= (B55 -40.) (;>>)

= 0.054 "(but not less than zero).

2. NON-BURST NODE The maximum temperature calculated for a non-burst section of clad typically occurs at an elevation above the core mid-plane

'during the core reflood phase of the I OCA transient. The poten-tial impact on that maximum clad temperature of using the NRC fuel rod models can be estimated by examining two aspects of The first aspect is the change in pellet-clad gap the'nalyses.

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 which burst is calculated. Three sets of LOCA analysis results were"studied to established'n ac'ceptable"s'ensi t'ivity. to apply gene'rically in this evalution. The possib'le 'PCT from a change in strain (in the'IIot Rod) is +20. F increase'esulting.

per percent decrease in strain at the maximum clad temperature

locations; Since the clad strain calculated during the reactor coolant syst~blo:;down phase of the accidents not changed by the use of /i~fuel rod models, the maximum Mcrease in clad strain that must be considered here is the difference bet;;een the "maximum clad strain" and .he "clad strain at the end of RCS bIowdown" indicated. above.

Therefore:

= 20 F -

!PCT -

) {HAn STRAIit BLOHDOHh STRAIH)

Ol

= { 20 ( 0.0892-0.0429)

Ol) 92.6 The second aspect of the ana1ysis that can increase PCT is the flo,v blockage calcu"laied. Since the greatest value of blockage indicated by the NRC blockage model is 75 percent, the maximu.",.

PCT increase can be estimated by assuming that the cu. rent level of blockage in the analysis (indicated above) is rais d to 75 percent and then aoplying an appropriate sensitivity formula shown in HS-TttA-Zlr 4.

Theref ore, hPCT4 = 1.25oF (50 PERCDT CURRENT BLOCKAGE)

+ 2.36 F (75-50)

= 1.25 (50 - 0.0 ) + 2.36 (75-50) 121.5 oF If PCT~l occurs>>hen the core ref lood rate is greater t!ian 1.0 inch per s cond BPCT4 = 0. The total potential PCT increas for th non-burst node is then APCT5 = BPCT3 + DPCT4 = 92.6 + 0 = 92.6'F Hargin to the 2200oF limit is APCT6 = 2ZOOoF - PCT~<] = 2]7 oF Th .Fg reduction required to maintain this 2200oF clad tem-perature. limit is ( rom HS-THA-2174)

AFAR, = (~PCT - L.CT ) (' }

= -.1244 10'F ~PCT hei( but not less than z ro.

The pea'~ing factor reduction r quired to maintain the 2200 'F clad temperature limit is ther fore ihe greater of LF(B and+/<<,II ol i Zk FgPE~,;y TY 0 i054 4

B. The effect on LOCA analysis results of using improved analytical and modeling techniques (whicn are currently appro;ed for use in the Upper Head Injectic~ pl nt LOCA analys s) in the reactor coolant system blowdown calculation (SAT'H cc.-;,-uter 'cod ) has" be n quanti-fied via an analysis wnicn has recentl been sub..ittc d to th lK for review. Recognizing that review of that analysis is not yet complete and that the benefits ass"ciated with those model i-.prcve-ments can change for. oihcr plant designs, th HRC has established a credit thai is acceptable for this in-'crim period to help offset penalties resulting from application of ihe tlRC fuel rod models.

That credii for tvo, thro e and four loop plants is an increase in the LOCA peaking =actor limit of 0.12, 0.15 and 0.20 respectively.

C: Tho peaking factor limit adjustment r quired to justify plant operation for this interim p riod is c termined as the appropriate

..t ir'..

~Fgll, cl orts+

A 1 =.:

-i<1- 8

~",5 ""

ln c:c.r ~isc.Ir (P i. <<p.,

)

'r.<< 't

.i....i c ion (A) above (but n"t greater than 7e,0]'."-

o P, iA

calculated'n Fg ADJUSTY~=">,'7 = . 0.15 - 0.054