ML19210B371

Interim Rept on Fuel Densification for Met Ed TMI-1.

ML19210B371
Person / Time
Site:	Oconee, Three Mile Island
Issue date:	04/30/1973
From:	METROPOLITAN EDISON CO.
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ML19210B370	List: ML19210B369 ML19210B371
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NUDOCS 7911070694
Download: ML19210B371 (7)

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INTERIM REPORT ON FUEL DENSIFICATIO'!

FOR THE METROPOLITAN EDISON THREE MILE ISLAND NUCLEAR STATION UNIT 1 r

APRIL 1973 1584 303 2495 4911070SPf

1. INTRODUCTIO" At the presen t time , the Regulatory Staf f is evaluating topical reports BAW-10055, " Fuel Densification Report", and LAN-1388, " Fuel Densification Report for Oconee Uni t 1". It is enticipated that by May 1,1973, the evaluation will be comoleted of the applicaticn of E6W analytical models in accordance with the guidelines set forth in the " Technical Report on Densificaticn of Light Water Reactor Fuels", USAEC, of Novetber 14, 1972.

The purpose of this interim report is to: (1) provide the Staff with suf ficient information to compare Three Mile Island - Unit 1 (TMI-1) with Oconee 1 and to describe any changes in methods or application thereof in the final report for TMI-1, and (2) present c preliminary eva'.uation of the rated power capability of TMI-l even considcring fuel densification effects.

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2. S U'.'PAPli Based on the analysis performed for TMI-1 using the methods outlined in BAW-10055, including a change in the power spike model and a change in input to this model (see Section 3), the follcwing conclusions are nade assuming that the fuel pellets densify up to 96.5!! theoretical density.

1. C1cdding collapse will not occur prior to 21,000 effective full pcuer hours of operation (greater than one fuel cycle) .

2. The mechanical performance of B&W fuel rods will not be impaired.

3. The interim acceptance criteria for ECCS are =ct.

4. The reactor can be safely operated at a power level of 2535 Ifit with minor codifications to the reactor protection system se tpoints.

a. A reduction in the overpower trip se tpoint from 114 to 112T; of 2335 MWt.

b. A reduction in allowable imbalance limits.

5. The increase in power peaking from gaps in fuel rods due to the assured densification can be compensated for by base load cperation as cpposed to load.follouing.

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3. RESULTS 3.1 Power Spike Model The power spike model described in DA'J-10055 will be used as described, except for one minor modification to the model and one change to input as described in the following subcecticns which are revised versict.s of the same subsections in B!.W-10055. .

2.3.4. Calculations of the Marimum Veid Size in A::ial Interval The equation for the determination oS,the ma:;:inun gap size asha function of ar.ial positica has been nodified fron AL = 0.965 - Di + 2r( ) + 0.004 L(2) 2 _

to the following at = 0.965 - ci + 0.004 L(z) 2 The term c(z) has been elininated from the determination of =aximu=

axial gap as a result of discussions with the s taf f and their evaluation of the Point Beach 2 Nuclear Plant Analysis.

2.4.1 Pcwer Soikes Due to a Sincie Gan and 2.4.2 Power Scikes Dee to Co-?laner Caos The calculaticnal cathods described in 3A.i-10055 will be changed to reflect the use of the DOT, Sn transport theory code in both X-?

and R-Z geometries. The power spike in the adjacent rods eill be obtained using EDT in X-Y gecnetry and the variation with gap lencth will be obtained T584 306 9

3. Result;, (centine:d) us in g R-Z ge or..e try . The ch an ge redefines the po er Ipike as the power increase in the adjacent pin rcther than that used in the third sentence in s ubs e c tion 2. 4.1. , "(i.e. , the percentage increase at the edge of the surrounding fuel region)". B&W has known that this definition of the power spihe due to a single gap was conservative and unrealistic and the true power increase in the adjacent pin has been determined. The results have been compared to those reported by Brookhaven National Laboratory in their Deccaber 1972 report, "Feaking Factors in Pressurized Water Reactors with Fuel Densification". The ins t figure in this report (un-numb e re d) ,

" Correction Factor F on Finite Gap Lengths", shows BNL results that are less than those values calculated by B&W for the same fuel enrichnent.

Therefore, the BOW method is conservative cenpared to the BNL method.

The DOT transports theory code in X-Y geotetry will also be used instead of the PDQ07 dif fusion theory code to determine the power spike in a center rod duc to more than one gap at the sane antal height in a surrounding array of fuel rods. The use of the trensport calculation provides a more accurate description of the physical system than does the diffusion theory calculation.

For the TMI-1 reactor these correction f actors will be calculated for the second zone fuel enriche.ent (2.75 w/o as-built), since this zone experiences the highest power peaking of the three fuel enrichnents in Core *l.

The preliminary data indicate that the power spike factor shoun in Figure 2-12 for theTMI-l reactor will be about 1.11 to 1.20 at 140 inches.

The f ollowing discussion en several additional sections of B.i%-10055 is presented to indicate the cancer in which the final analysis for TMI-l.

will be performed.

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3.2 Thc rna l Annivs's .

The application of the power spike factor in the analysis of fuel tcrperatures and D::BP analysis will be that approved by the staff for B A'.'-10 05 5 . The as-built data for TAI-l will, of course, be used in the evaluation.

The preliminary evaluation has shoun that the DNBR less due to deacification frc= about 92.5 to 90.5% T.D. results in a power peaking rargin reduction of about 4 ta 5%.

The fuel reiting criterion will be abcut 21 kU/f t based en a taximum first cycle burnup of 20,800 Eld /ttU, uhich includes a 10,% uncertainty margin.

These effects will be compensated for by reducing of the overpcrar trip setpoint from 114 to 112% of 2535 E!t and changing the operational ,

philosophy f rom load following to base loaded cperation.

3.3 Nuclear Analvsis The channe in operation from load following to a base loaded philosophy will reduce pcwer peaks more than enouch to .cc=pcasate for the effects derivec in section 3.2. This change in operation will allev greater imbalance trip limi ts , for T:iI-l than for Ocence I in spite of the fact that the average density is 92.5% cocpared to 93.5% T.D. and the fuel enrichment censidered in the deternination of the power spike facter is 2.75 w/o compared to 2.15 w/o for Oconee I.

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ow w . .3 3.4 Sa fe ty Analvsis e The ground rules and acceptance criteria adopted for the Oconee 1 report vill be achered to in the T!'I-l c cl: sis.

Mechenical An alvets 3.5 The applica ica of the 2. cl. m colle, se ,2ds.1 vill be the sa ue fci to o, a>.!;catica v. cm >_ar A tor ccone 1. . ae ca l- en 'm' till be !- tha an-u n ? ;

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.S lt e ch.m ical , ris, (coatinutd) data. T. e re f ore ' clad e 'iapse is not prodicted prior to 21,000 ef fective full pc.;c: hourc of operatien, 4

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