ML20134M973

Safety Evaluation Re Core Reload Methodology Changes.Changes Acceptable

ML20134M973
Person / Time
Site:	Fort Calhoun
Issue date:	08/26/1985
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20134M967	List: ML20134M964 ML20134M973
References
NUDOCS 8509040361
Download: ML20134M973 (4)

Text

-

s,

%, UNITED STATES

[ ,, g NUCLEAR REGULATORY COMMISSION ,

7, j WASHINGTON, D. C. 20655

]

\, . . . . . / l l

l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT NO. 1 l DOCKET NO. 50-285 CORE RELOAD METHODOLOGY CHANGES

1.0 INTRODUCTION

By letter dated June 13,1985 (Ref.1), the Omaha Public Power District (OPPD) submitted reload methodology changes which affect the previously approved report entitled " Reload Core Analysis Methodology Overview",

OPPD-NA-8301-P. These methodology changes are due to the following modifications and fuel system design changes which will be made to the Cycle 10 core of Unit 1 of the Fort Calhoun Station:

a) Part length poison rods will be inserted in peripheral assemblies to further reduce the flux to the reactor vessel beltline welds.

b) The Themal Margin / Low Pressure (TM/LP) calculators will be converted from the "early system" type which

~~

'did not monitor the axial shape index to the " standard system" type units which include axial shape dependency.

c) Mini-CECOR/Better Axial Shape Selection System (BASSS) will be installed on the plant computer for incore monitoringoftheDNBandLinearHeatRate(LHR)

Limiting Conditions for Operation (LCO).

These changes are incorporated in the revised methodology overview topical report,OPPD-NA-8301-P, Revision 1(Ref.2).

8509040361 e5002625 DR ADOCK 050

I .

2.0 FUEL SYSTEM DESIGN In order to further reduce the flux to the reactor vessel beltline welds to minimize the RTNDT shift, Cycle 10 may contain part length poison rods. 1 These rods are composed of 4B C over the middle 50% of length with inert material comprising the balance of the rod. When used, these rods will be inserted in peripheral assemblies located near the critical weld locations.  !

I Any resulting increases in radial power peaking due to these part length rods will have to be evaluated during the Cycle 10 reload review. Also, improvements in the monitoring systems to accommodate radial peaking increases are discussed in Section 3.0 of this SE. Based on this, the incorporation of part length poison rods into the fuel loading pattern

% is acceptable.

I

. 3.0 SETPOINT CENERATION

The Fort Calhoun Reactor Protection System (RPS) previously utilized the "early system" local power density trip and TM/LP trip. This TM/LP trip didnotronitortheaxialshapeindex(ASI). The TM/LP calculators are now being converted to the " standard system" type units which include axial shape dependency. Both of these systems are described in CENPD-199, '

~

l Revision 1 (Ref. 3) and are acceptable for use in CE reactors.

Monitoring of the LCO for peak linear heat rate and DNS will be performed with the Better Axial Shape Selection System (BASSS). This system has been reviewed and approved by the NRC and is presently being used in the l Calvert Cliffs units. The incore monitoring of DNB with BASSS affords a

! gain in operating margin as compared to the excore ASI monitoring system.

i This is due to a reduction in ASI uncertainty, use of the actual Control ElementAssembly(CEA)grouppositioninsteadoftheexcoresystem's

! assumption that the CEAs are inserted to the maximum allowed limit, and use of the actual radial peaking factors instead of the excore system's  ;

l assumptions that radial peaks are at the Technical Specification limits.

, 2 1 <

~ , - - . - - , - --

. - - - n.---

. . _ i l

. l

- BASSS uses input from the CECOR code as well as from the plant computer for LCO monitoring. The power distribution uncertainties which are used by OPPD were approved during our review of CECOR (Reference 4), which has been  :

operating on the CE CDC computer. For Fort Calhoun, OPPD plans to eventually utilize a minicomputer version of.CECOR (mini-CECOR) which has been developed by CE. In order to justify the continued use of the previously approved 7

power distribution uncertainties with mini-CECOR, it must be shown that identical results are obtained from both versions of the CECOR code.

Therefore, OPPD has agreed to submit results of QA testing performed to validate that the conversion of CECOR from a CDC mainframe computer to an ERF minicomputer gives the same calculational results. Upon review ind approval of this validation, the use of mini-CECOR will be acceptable.

4.0 CONCLUSION

S As a result of our review, we find the proposed core reload methodology changes.for Fort Calhoun Cycle 10 to be acceptable. The continued use of the previously approved power distribution uncertainties is, however, contingent upon the results of QA testing performed to validate the conversion of the CECOR code to a minicomputer.

Principal Contributor:

L. Kopp, DSI .

Date: August 26, 1985 s

3

REFERENCES

1. Letter from R. L. Andrews (OPPD) to E. J. Butcher (NRC), " Core

' Reload Methodology Changes for Cycle 10", LIC-85-237, dated June 13, 1985.

2. " Reload Core Analysis Methodology Overview", OPPD-NA-8301-P, Rev. 01, Omaha Public Power District, June 1985.

3. "CE Setpoint Methodology", CENPD-199-P, Rev. 1-P, Combustion Engineering, March 1982.

4. " INCA /CECOR Power Plant Uncertainty." CENPD-153-P, Rev. 1-P-A, Combustion Engineering, May 1980.

i O-4

. - _ ___. _ _ _ -_.