ML17174A353

Forwards Info Re Generic Cladding Swelling & Rupture Model Info Submitted by NSSS Vendors Re NUREG-0630.GE Approved Models Conservative.Westinghouse Plants Comply W/ Regulations

ML17174A353
Person / Time
Site:	Dresden, Quad Cities, Zion
Issue date:	01/08/1980
From:	Peoples D COMMONWEALTH EDISON CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
NUDOCS 8001150519
Download: ML17174A353 (8)

Text

--

e9 Commonwealth Aon

.One First National Plaza, Chicago, Illinois

. Address Reply to: Post Office Box 767 Chicagb, llli.nois 60690.

. January* 8, 1980 Mr. Darrell G. Eisenhut, Acting Director Divisi.on of Operating Reactors*

Office of Nuclear Reactor Regulation U.S. Nuclear* Regulatory Commission Washington, DC 20555

Subject:

Confirmation of Generic Cladding Swelling and Rupture Models for LOCA Analysis, Dresden Station Units 1, 2 and 3 Quad Cities Station Units 1 and 2 Zion Station Units 1 and 2 NRC Docket Nos. 50-10/237/249, 50-254/265, and 50-295/304 Reference (a):

D. G. Eisenhut letter to All Power Reactor Licensees dated November 9~ 1979 (b):

R. H. Buchholz (GE) letter to D. G. Eisenhut dated November 2, 1979 (c) :

"General Electric Comp~ny Analytical Model for Loss-of-Coolant Anaiysis in Accordance

. with 10 CFR 50 ;\\ppendix K", NED) -20566, January 1976.

(d):

R. H. Buchholz (GE) letter to D. G. Eisenhut dated November 16, 1979 (e):. T. M. Anderson (W) letter to D. G. Eisenhut, NS-TMA-2147, dated November 2, 1979

( f) :

( g) :

Dear Mr.. E*isenhut:

T.. ~. Anderson (W) letter to D. G. Eiienhut, NS-TMA-2163, da't:ed Dece.mber 16, *1979, *

• T. M. Anderson (W) letter to D. G. Eisenhut, NS-TMA-2174, dated December *7, 1979 As requested in Reference (a), Commonwealth Edison has reviewed the generic cladding swelling and rupture model information submitted to the NRC by the nuclear steam supply (NSSS) vendors.

Enclosed is a discussion of that information At>39 which addresse*d the applicability to our operating units A flD/J:

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hEd" Commonwealt 1son NRC Docke Nos. 50-10/237 /249, 50-254/265~ and 50-295/304 Mr. Darrell G. Eisenhut 2 -

January 8, 1980 In the caae of General Electric. units, as discussed in Reference (b), the ORNL Multi-Rod burst Test data is consistent wlth and.in fact c6mplements, over the BWR rang~ of inter~st, the data.used by GE as a basis !or the formulation of the NRC approved Appendix K model (Reference c).

For this reason*GE tbok the position at the October 30, 1979 meeting with the NRC on this subject that the recent data from ORNL did not pertain to ~he.BWRi At the October 30th meeting, the NRC concurred with this position.

The NRC reiterated their conclusion of the lack of an impact upon the BWR at the public NRC Commission Briefing of November 2, 1979.

Subsequent to'the November 2nd briefing, GE was asked to reverify the sensiti-(,ity of its model to changes in ramp rates to perforation.

The GE response (Reference d) concluded that.the GE model is conservative with respect to the NRC fuel rupture curves.

General Electric concluded that the ORNL data did not provide any new data which could be used as a basis for consideration* of potential model changes.

Furthe~more, GE. has concluded~ upon review of the dr~ft of NUREG-0630 which addresses models for LOCA analyses, that the ORNL data does not affe*ct the approved GE Appendix K model.

Therefore, there is no basis to change either the individual plant ECCS analysis or the individual plant Technical.Specifications.

In the case of the Westinghouse units, Westinghouse representatives presented information at the November 1, 1979

(

meeting calle'd by the NRC for plants licensed with the West-inghouse ECCS evaluation model, and discussed the potential impact of fuel rod model changes on result-$' of those analyses.

That information was formally documented in Reference (e), and form~d the basis for the. Westinghouse conclusion that t~e in-formation presented in <lraft NUREG-0630 did not constitute a

~afety problem for Westinghouse.plants and that all plants conformed ~ith NRC regulations.

As a result of compiling the information in.Reference (e), 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 degrees 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 ~ffect of this issue and the effect of the NRC fuel rod models. had to be studied.

Details of

.the work done on thi~ issue were presented to the NRC on November 13, 1979 and doc~mented in ~eference (f).

That work included d~velop ment of a procedure to determine the clad heatup rate prior to burst and. a reevaluation of operating Westinghous~ pl~nts with consideration of a: modified Westinghouse fuel rod burst model..

As part of this reevaluatiori, the Westinghouse position on NUREG.

063Q did not constitute a safety problem for plants licensed with the Westinghous~ ECCS evaluation model.

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-Commonwealth Edison NRC Dock, Nos. 50-10/237 /249,

~0-254/265, and 50-295/304 Mr. Darrell G. Eisenhut 3 -

January 8, 1980 On December 6, 1979, NRC and Westinghouse personnel discussed the information thus far presented.

At the conclusion of that discussion, the NRC staff requested Westinghouse to provide further detail on the potential impact of modifications to ea~h 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 improve-ments.

This additional information was compiled from various LOCA analysis results and documented in Reference (g).

Enclosed is the plant sp~cific evaluation of the potential impact* of using the fuel rod models presented in draft.NUREG-0630 *an the Loss of Coolant Accident analysis for Zi6n Units 1 and 2.

One (1) signed original ahd seventy-nine (79) copies of this letter ate submitted for your informatioh.

It is on the basis of.the foregoing discussi6n and the enclosed evaluations that it is judged that no further actions are required.

Very truly yours,

}.

• • ~..

D. L. Peoples Director of Nuclear Licensing Enclosure

f t r.

ENCLOSURE 'l Confirmation of Generi~ tladding Swelling and Rupture Models for LOCA Analysis Dresden Units 1, 2 and 3; Quad Cities Units 1 and 2 I.

The following is in response to the diiective contained in in the D. G. Eisenhut letter of November

~, 1979 which requested confirmation that the fuel rod strain-fl6w blockage models used for Dresden Units 2 and 3, and Quad Cities Units 1 and 1 are conservative with respect to the NRC model which was derived from ORNL ~ulti-Rod Test Facility Data.

Since we do not technically have or use our own models in this area, the following discussion pertains to the use of General Electric Company approved models in association with the operation of Dresden Units 2 and 3 and Quad Cities Units 1 and 2.

Through the D. G. Eisenhut letters of November 9 and 27, 1979 and other correspondence from GE, we are aware of th~ review*

effort GE has applied to this iss~e.

We are also aware of the communications between GE and the NRC on this subject.

We herein affirm that based on our correspondence from GE that it is our understanding that over the operating range of interest pertinerit to the above named units that the-.models applied by GE and approved by the NRC are of a conservative level equivalent to the NRC model.

II.

Based on the fact that the Dresden l fuel is no~ prepressurized and the operating conditions for this BWR unit are substantially the same as for the GE-BWR units discussed above, the same conclusion has been reached for Dresden 1, even though this fuel was not analyzed by General Electric.

Therefore, we affirm

• ~.

that over the operating range of interest pertinent to Dresden

~

1, that to the best of our knowledge, the models applied in ECCS evaluation of that unit are conservative, and no further assessment is required.

ENCLOSURE 2 Confirmation of Generic Cladding Swelling and Rupture Models for LOCA Analysis Zion Units 1 and 2 A.

Eval~ation of the potential impact of using fuel rod models presented.in draft NUREG-0630 on the Loss of Coolant Accident (LOCA) analysis for Zion Units land 2..

This evaluation is based on the limiting break LOCA analysis identified as follows:

BREAK TYPE -

DOUBLE ENDED COLD LEG GUILLOTINE BREAK DISCHARGE COEFFICIENT -

0.8 WESTINGHOUSE ECCS EVALUATION MODEL VERSION -

FEBRUARY, 1978 CORE PEAKING FACTOR -

1.86 HOT ROD MAXIMUM TEMPERATURE CALCULATED FOR THE BURST REGION OF THE CLAb -

2174

°F = PCT B ELEVATION -

5.75 FEET HOT ROD MAXIMUM TEMPERATURE CALCULATED FOR A NON-RUPTURED REGION OF THE CLAD -2008°F = PCTN Elevation -

7.75 FEET CLAD STRAIN DURING SLOWDOWN AT THIS ELEVATION -

4.1 PERCENT;MAXIMUM CLAD STRAIN AT THIS ELEVATION -

8.3

. PERCENT MAXIMUM TEMPERATURE FOR THIS NODE OCCURS WHEN THE CORE RE-FLOOD RATE IS GREATER THAN 1.0 INCH PER SECOND AND REFLOOD HEAT TRANSFER IS BASED ON. THE FLECHT CALCULATION.

AVERAGE HOT ASSEMBLY ROD BURST ELEVATION -

N/A HOT ASSEMBLY BLOCKAGE CALCULATED *- 0. 0 PERCENT

1.

• BURST NODE The maximum potential impact on the ruptured clad node is expressed in letter NS-TMA-2174 in terms of the change in the peaking factor limit (FQ) required to maintain

• peak clad temperature (PCT) of 2200 degrees F and in

. terms of a change in PCT at a constant FQ.

Since the

~lad-~ater reaction rate increase~ significantly at t~mperatuie~ above 2200 degrees F, individual effect~

(such as A PCT due to changes in several* fuel.rod models) indicated here. may not accurately apply over large ranges, but a si~ultaneous change in FQ which tauses the PCT to r~~ain in the neighborhood of 2200 degrees F justifies us~ of this evaluation procedure.

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From NS-'IMA-2174:

For the Burst Node of the clad:

0.01AFQ~...r1500F BURST NOOE.1PCT Use of the NRC burst model could require an FQ reduction of 0.015

- The minimum 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:

4PCT1 * (.015 +.03) (150°F/.Ol)

• 675°F Margin to the 2200°F limit is:

.The FQ reduction required to maintain the 2200°F *clad temperature limit is:

2.

~ FQa * (4PCT1 _ APCT2) (

• 675 - 26)

.014F*~

150 ~F

• .04 (but not less than zero).

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 LOCA transient.

The potential impact on that maximum clad temperature of using the. NRC fuel rod models*

can be.estimated by examining two aspects of the analyses.

The

first aspect is the change in pellet-cl~d gap conductance resulting from a difference in clad strain at the non-burst maximum clad tem-perature 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 establish an acceptable sen-

. sitivity. to apply generically in this evaluation.

The possible PCT increase resulting from a change in strain (in the Hot Rod) is +20.°F per* percent decrease in strain at the maximum clad temperature.

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Since the clad strain calculated d~ring the reactor coolant system blowdown phase of the accident is not changed by the use of NRC fuel rod models,* the maximum decrease in clad strain that must be con-sidered here is the difference between the "maximum clad strain" and the "clad strain at the end of RCS blowdown" indicated.above.

Therefore, A.PCT3 (200f

) (MAX STRAIN -

BLOODOWN STRAIN)

.01 strain

= ( 20) (.083 -

.041)

-:-or

=

84 The second aspect of the analysis that can increase PCT is the fl~w. block-age calculated.

Since the g~eatest value of blockage indicated by the NRC blockage model is 75 percent, the maximum PCT increace can be est~~ated by assuming that the current level of blockage in the analysis (indicated above) is raised to 75 percent and then applying an appropriate sensitivi'ty formula shown in NS-TMA-2174.

Therefore,

1. 25°F (SO - PERCENT CURRENT BLOCKAGE)

+ 2.36°F (75-50)

= 1.25 (50-0) + 2.36 (75~50)

= 12l°F If PCTN occurs when the core reflood rate is greater than 1.0 inch per second APCT4 = 0.

The total potential PCT increase fer the. non-burst node is then Margin.to the 22QOOF limit is The FQ reduction required to maintain this 2200°F clad temperature limit is (from NS-TMA-2174).

AF~

AF~ -

-.11 but not less than zero.

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F B.

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The peakirig factor reduction required.to maintain the 2200°F clad tempera tu re limit is there fore the greater of.6 iQ8 and*.

~ FQN'.

or i

• AFQPENALTY =

• 04

.. i\\

The effect on LOCA analysis results of using impro~~i)analytical and modeling techniques (which are currently* approved for use in the Upper Head Injection plant LOCA analyses) in the reactor coolant system blowdown claculation (SATAN computer code) has been quantified via an analysis which has recently been sub-mitted to the NRC for review.

Recognizing that review of that analysis is not yet complete and that the benefits associated with those model improvements can change for other plant designs, the NRC has established a credit that is acc:eptable for this interim perio~ to help offset penalties resulting from application of the NRC fuel rod models.

That credit for two, 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 adjustment required to justify plant operation for this interim period is determined as the ap~

propriate4FQ credit identified in section (B) above, minus theAFQPENALTY calculated in section (A) above (but not greater than zero).

FQ ADJUSTMENT= 0.20 -

0.04

= 0 D.

Furthermore, the proposed amendment to the Zion Technical Specifications submitted by the O. L. Peoples letter to H. R; Denton dated October 22, 1979 justifies an increase in the allowable LOCA peakirig factor limit from the current value of 1.86 t6 a value of 2.26~

This value is based on an ECCS reanalysis performed using the Westinghouse ECCS Evaluation

• Model approved by the NRC Staff in August, 1978.

These results further demonstrate the existence of4F0 credits previously discussed in Sections B and C.