ML20064G551

Fuel Bundle Loading Error Reanalysis for the Ja Fitzpatrick Nuc Pwr Plant for Reload 2,Suppl 2

ML20064G551
Person / Time
Site:	FitzPatrick
Issue date:	11/30/1978
From:	Brugge R, Zull L GENERAL ELECTRIC CO.
To:
Shared Package
ML20064G547	List: ML20064G544 ML20064G551
References
NEDO-24129-2-S02, NEDO-24129-2-S2, NUDOCS 7812120082
Download: ML20064G551 (11)

Text

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NEDO-24129-2 Supplement 2 1

Class 1 November 1978 FUEL BUNDLE LOADING ERROR REANALYSIS FOR THE JAMES A. FITZPATRICK NUCLEAR POWER PLANT FOR RELOAD NO. 2 Prepared by:

L. M. Zull

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Sr. Licensing Engineer Approved by

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R. . Brugge, M ger Operating Licenses II NUCLE AR ENERGY PROJECTS Divls ON

• GENE R AL ELECTalc COMPANY SAN JOSE, CALIFORNI A 95125 7 81212 o o g;t t

, .- NED0-24129-2 IMPORTANT NOTICE REGARDING

., CONTENTS OF THIS' REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for the Power Authority of the State of New York (The Authority) for The Authority's use with the U. S. Nuclear Regulatory Commission (USNRC) for amending the Authority's operating license of the James A. FitzP,atrick Nuclear Power Plant. The information contained in this report is believed by General Electric to be an accurate and true representation of the facts known, obtained or provided to General Electric at the time this repert was prepared.

The only undertakings of the General Electric Company respecting information in this document are contained in the contract between The Authority and General Electric Company (GE Requisition No. 300-8856 dated July 31, 1978 and The Authority's Change Order No. 9 to Contract AP-3) and nothing contained in this document shall be construed as changing said contract.

The use of this information except as defined by said contract, or for any purpose other than that for which it is intended, is not authorized; and with respect than that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither General Electric Company nor any of the contributors to this document makes any representa-tion or warranty (express or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information.

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. .- - NEDO-24129-2 l.0 Introduction and Summary ,

This supplement to the James A. F.tzPatrick Reload 2 licensing submittal (Reference 1) presents the results of a reanalysis of the rotated bundle loading error event using the variable water

. gap method recently approved by the Nuclear Regulatory Commission (Reference 2).

The results of the analysis show that the minimum CPR for the most limiting rotated bundle in the co.re (8DRB283) does not exceed the Scfety Limit (1.07) for all exposure during Cycle 3.

The rotat?d bundle event becomes the limiting event which determines the cgerating limit for 8X8R fuel for exposure between 60C3 and EOC3-2 Gwd/t. The operating limit is increased from its previous value of 1.20 (Reference 1) to 1.22. The operating limits for the other fuel types in the Reload 2 core remain unchanged fron those presented in Reference 6.

2. 0 Background A loading error for a fuel reload reference core is defined (Reference 3) as:

(1) A reload bundle is rotated, but inserted in its proper core location; or a bundle is inserted correctly, but in an improper core location; and ,

(2) the error is not discovered in the subsequent core verification, and the reactor is operated.

The rotated bundle and mislocated bundle loading erro 6 vents for the FitzPatrick Reload 2 core were previously analyzed us '.ig the standard procedure (analytical models and assumptions) described in References 4

( and 5. The results are. presented in Table 1. The results for the most limiting loading error event (a rotated 80RB283 bundle) were previously presented in Section 15 in the Reload 2 licensing submittal

(Reference 1).

l Because the analysis of the rotated bundle event using the standard i

method resulted in an MCPR of 0.96 (which is below the Reload 2 core Safety Limit of 1.07) the event was reanalyzed using the new variable water gap method. This method was approved by the NRC in Reference 2.

3.0 New Rotated Bundle Analysis Procedure The principle difference between the standard analysis procedure and t the new analysis procedure is the modeling of the water gap along the l

axial length of the bundle. The pr'vious e analysis used a uniform water gap, whereas the new analysis utilizes a variable water gap which is representative of the actual condition. 'The effect of the variable l

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. NED0-241.9-2 water gap is to reduce the power peaking and R-factor in the upper regions'of the.. limiting fuel rod' This results in the calculation of a reduced aCPR for the rotated bundle. The calculation was performed using the same analytical models as were previously used (Reference 4 and 5). The only change is in the simulation of the water gap which more accurately represents the actual geometry.

In the new analyses, the axial alignment of a 180 rotated bundle conservatively ignores the presence of the channel fastener. The more limiting condition of assumi'ng that the spacer buttons are in contact with the top guide is assumed. There is no known loading that could bend or break the channel spacer button during the insertion of a ISO rotated bundle, since both the top guide and spacer button are chamfered to provide lead-in. For a properly assembled bundle, no gechanism exists which could invalidate the assumption that a 180 rotated bundle leans to one side.

It should be noted that proper orientation of bundles in the reactor core is readily verified by visual observation and assured by verification procedures during core loading. Five separate visual indications of proper bundle orientation exist:

(1) The channel fastener assemblies, includir.g the spring and guard used to maintain clearances between. channels, are located at one corner of each fuel assembly aojacent to the center of the control rod.

(2) The identification boss on the fuel assembly handle points toward the adjacent control rod.

(3) The channel spacing buttons are adjacent to the control rod passage area.

( (4) The assembly identification numbers which are located on the fuel assembly handes are all readable from the direction of the center of the cell.

(5) There is cell-to-cell replication.

Experience has demonstrated that these design features are clearly visible so that any misloaded bundle would be readily identifiable during core loading verificatjon. Figures 1, 2 and 3 dgnote a normally loaded bundle, a 180 rotated bundle, and a 90 rotated bundle, respectively. Actual experience (References 4 and 5) has demonstrated that the probability of a rotated bundle is low.

4.0 Analysis Results The results of the rotated oundle analysis for the FitzPatrick Reload 2 core using the new rotated bundle analysis procedure are presented in Table 2. The results show that the minimum CPR for the most limiting rotated bundle in the core (8DRB283, an 8X8R type fresh fuel bundle) does not exceed the Safety Limit (1.07) for all exposures during Cycle 3.

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.' NEDO-24129-2 rotated bundle event becomes the With limiting anevent jnitialwhich CPRdetermines of 1.22, the,he t operating limit for 8X8R fuel for exposures Letween BOC3 and EOC3-2 GWd/t. Previously (Reference 1),

the loss of 807 feedwater heating transient determined the operating li::;it for 8X8R fuel for this exposure range. The operating limits for 7X7 and 8X8 fuel in this exposure range remain unchanged.

For exposure between EOC3-2 GWd/t and EOC3, the operating limits for all fuel types (7X7, 8X8 and 8X8R) remain unchanged. The new MCPR operating limits for Cycle 3.are summarized in Table 3.

5.0 References

1. " Supplemental Reload Licensing Submittal for the James A. FitzPatrick-Nuclear Power Plant for Reload No. 2," June 1978 (NED0-24129) .

2. Letter, D. G. Eisenhut (NRC) to R. E. Engel (GE), May 8, 1978.

3. " Generic Reload Fuel Application Licensing Topical Report,"

March 1978 (NEDE-240ll-P-3 and NED0-240ll-2)

4. Letter, R. E. Engel (GE) to D. Eisenhut (NRC), " Fuel Assemble Loading Error," MFN-219-77, June 1, 1977.

5. Letter, R. E. Engel (GE) to D. Eisenhut (NRC), " Fuel Assembly Loading Error," MFN-457-77, November 20, 1977.

6. " Raised Safety / Relief Valve Setpoint Reanalysis for the James A. FitzPatriqk Nuclear Power Pi *it for Reload No. 2,"

September 1978 (NED0-24129-1).

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. . , NE00-24129-2 Table 1 '

James A.' FitzPatrick Nuclear Power Plant Reload 2, Cycle 3 Fuel Bundle Loading Error Results Using

.. Standard Analysis Procedure -

Loading' Error Description Initial CPR Minimum CPR Rotated Bundle 80RB283 1.20 0.96 I 80RB265L 1.20 1.02 MjslocatedBundle 80RB283 in 70250 Location 1.21 1.09 80RB265L in 70250 Location 1.21 1.11 l .

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, NED0-24129-2 5 -

• - Table 2 James A. FitzPatrick Nuclear Power Plant Reload 2, Cycle 3

.. Fuel Bundle Loading Error Results Using New Analysis Procedure

.,, Loading Error Description Initial CPR** Minimum CPR Rotated Bundle

• 80RB283 1.22 > 1.07 80RB265L 1.22 > 1.07

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• Variable water gap method of analysis

• • Initial CPR for nonrotated bundle. Includes 0.02 CPR penalty for uncertainties in GEXL predictions using axially varying R-factors

NEDO-24129-2

. Table 3

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FitzPatrick Cycle 3 Operating Limits With Revised

• S/RV Setpoints

. And New Rotated Bundle Loading Error Analysis Exposure Range MCPR Operating Limit **

7X7 gXg 8X8R From 80C3 to EOC3-2 GWd/t 1.21 1.22 1.22 From EOC3-2 GWd/t to EOC3-1 GWd/t 1.25 1.33 1.33 From EOC3-1 GWd/t to EOC 3 1.30 1.37 1.37 5

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• Described in Reference 6 j **The Safety Limit MCPR is 1 07 .

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1 Figure 3. Rotated Bundle, 90 Degree Rotation

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