ML20028G779

Application for Amend to License DPR-51,revising Tech Specs 2.1,2.3,3.1.7 & 3.5.2 Re Cycle 10 Core Design.Cycle 10 Reload Rept Encl

ML20028G779
Person / Time
Site:	Arkansas Nuclear
Issue date:	08/08/1990
From:	Carns N ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20028G780	List: ML20028G779 ML20028G781 ML20028G782
References
1CAN089006, 1CAN89006, NUDOCS 9009040124
Download: ML20028G779 (8)

Text

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ggg m Operat s August 8, 1990 ICAN089005 U. S. Nuclear Regulatory Commission Document Control Desk Mail Station P1-137 Washington, DC 20555

Subject:

Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. OPR-51 Cycle 10 Reload Report and Proposed Technical Specification Change Request Attached is the ANO-1 Cycle 10 Reload Report for your review. Included with the report are proposed Technical Specification changes required as a result of the reload.

Entergy Operations, Inc. has evaluated the proposed changes in accordance with 10CFR50.91(t)(1) using the criteria in 10CFR50.92(c) and has determined that these changes involve no significant hazards consideration. The bases for these determinations are included in the enclosed submittal.

As discussed with the staff in a July 10, 1990, telephone conversation, Enterdy Operations has initiated a re-analysis.of the Main Steam Line Break event which will also address end-of-cycle moderator temperature coefficient concerns currently under evaluation. The results of this analysis will be relayed to the NRC upon completion, currently anticipated by year's end.

This topic is discussed in the enclosed submittal.

Also discussed in the same conversation was the possible revision of the MHA and LOCA dose consequences contained in this Reload Report due to the plant modifications required to return to 100% full power. The analyses associated with these plant modifications have been completed and the results incorporated into this Reload Report so that a future revision will not be necessary.

Very truly yours, TA W' N. S. Carns NSC:JGH:cip Attachments / Enclosures 9

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90090A0124 900808 L PDR ADOCK 05000,313 / )1(

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s U. S. -. NRC '

e4 .Page 2 August 8, 1990 L cc: Mr. Robert Martin-L U. S. Nuclear Regulatory Commission

)- Region IV L 611 Ryan Plaza Drive, Suite 1000

. Arlington,'TX 76011-NRC. Senior Resident Inspector Arkansas Nuclear One - ANO-1 & 2 P . Number 1, Nuclear Plant Road Russellville, AR 72801 g

Mr. Thomas W. Alexion

NRR Project Manager, Region IV/ANO-1 U. S. Nuclear Regulatory Commission NRR Mail Stop 11-B-19 One White Flint North 11555 Rockville Pike-Rockville, Maryland 20852 Mr. Chester Poslusny NRR Pro' ject Manager, Region IV/ANO-2 V. S. Nuclear Regulatory Commission NRR Mafi Stop 11-B-19 One White Flint North 11555 Rockville Pike Rockv111e' Maryland 20852

.Ms. Greta Dicus, Director Division of Radi6tien-Control and Emergency Managen.cnt-Arkansas-Department of Health 4815 West Markham Street Little Rock, AR' 72201 I

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c. STATE OF ARKANSAS ') '

) SS i COUNTY OF PULASKI -). 'i e

I, N. S. Carns, being duly sworn, subscribe to and say that I am Vice President, ,

i Operations ANO for Entergy Operations, Inc.; that I have full authority to .

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execute this oath, that I have read the document numbered ICAN089006 and know the l p contents.thereof; and that to the best of my knowledge, information and i

.t belief the statements in it are true.- l

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DA bw: h N. S. Carns

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SUBSCRIBED AND SWORN T0 before me, a Notary Public in and for the i County and State above named, this k day of b 2 rui X ,

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ENCLOSURE- .,

t PROPOSED TECHNICAL SPECIFICATION l 4

AND  !

RESPECTIVE SAFETY ANALYSES -!

t 1N THE MATTER OF AMENDING  :

LICENSE NO. DPR-51 .>

l ENTERGY OPERATIONS, INC.  !

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ARKANSAS NUCLEAR ONE, UNIT 1 l

DOCKET NO. 50-313 i

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PROPOSED CHANGES The proposed amendment would change ANO-1 Technical Specifications 2.1, 2.3,

! 3.1.7 and 3.5.2. Revised copies of the affected pages are included in this attachment. The following changes are proposed:

a. Figure 2.1-2 is modified to reflect the Cycle 10 core design.

b. rigure 2.3-2 is modified to reflect the Cycle 10 core design.

c. Section 3.1.7 is modified to reflect an increase in the moderator temperature coeffich.% limit to +0.9 x 10 4 Ak/k/ F.from +0.5 x 10 4 Ak/k/"F. This change L implemented to make AN0-1 Technical Specifications match more recently licensed B&W units in support of the Standardized Technical Specifications project. The most limiting event for positive moderator temperature coefficient has been evaluated for 6-an MTC value of +0.9 x 10 4 Ak/k/ F.

i A The Control Rod Grcy and Power Distribution Limits of Technical Specification 3.5.2 contain the following changes:

1. Section 3.5.2.4 is modified to reflect new quadrant power tilt limits for Cycle 10. The phrase "in excess of 4.24%" was also added to action statement 3.5.2.4.2a in order to clarify the required protection system setpoint reduction.

2. Section 3.5.2.5.4 is modified to accommodate Cycle 10 changes to the control rod insertion limits and operational limits on the gray axial power shaping rods (APSRs). Corresponding changes are made to Figures 3.5.2-1 (A-C), 3.5.2-2 (A-C), and 3.5.2-3 (A-C) for the control rod insertion limits.

3, Figures 3.5.2-4-(A-C) are modified to reflect new operational power imbalance limits for Cycle 10.

4. Figure 3.5.2-5 is modified to provide new LOCA linear heat rate limits for Cycle 10.

DISCUSSION OF PROPOSED CHANGES These proposed changes are all enveloped by and presented in the AND Unit 1, Cycle 10 Reload Report. Analytical techniques and design bases were employed which have been accepted by the NRC. References noted in the Reload Report describe the techniques utilized in the analyses. The Cycle 10 Reload Report was developed based upon a Cycle 9 length of 420 EFPD. Due to outage scheduling considerations, the Cycle 9 shutdown is expected prior to reaching the 420 EFPD design length. A Cycle 10 reanalysis was performed using a 350 EFPD Cycle 9 length to determine the impact and these results have indicated that the Reload Report parameters presented here are still valid.

The effect of changes in cycle specific parameters on each accident analysis addressed in the ANO-1 FSAR has been considered in Section 7 of the Cycle 10 Reload Report. With the exception of LOCA and MHA, the radiological l

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consequences of all of the evaluated accidents with the specific nuclide inventory from Cycle 10 changed only slightly from Cycle 9. The LOCA and MHA doses. exhibited a significant reduction due to the adopt an of Regulatory Guide 1.4, Revision 2 (1974) guidelines for icaw v;acies fractions and Standard Review Plan 6.5.2, Revision 2 gus h . mes ur calculating Reactor Building Spray (RBS) performance. 'ih m e M ts also reflect the anticipated RBS throttling required to addre s N' a ccerns.

For this reload modification, with the debris resistant fuel ar.sembly design, considerations with respect to margins of safety for the fel system design, nuclear design, and thermal-hydraulic design are addressed in sections 4, 5 and 6 of the Cycle 10 Reload Report. The applicable value limits and margins have L.aen determined to be within allowable limits and requirements for acceptable Cycle 10 operation.

Bounded by the analysis of the Cycle 10 Reload Report, the quadrant power-tilt, rod position and power imbalance setpoints of Technical Specification 3.5.2 have been revised. The new setpoints of Cycle 10 assure that the maximum cladding temperature will not exceed the final acceptance criteria in 10CFR50, Appendix K, assuming worst case power distribution. These revised setpoints have been obtained using NRC approved codes and methodology, taking into account all perceived uncertainties, worst case conditions and core burnup.

The revised rod insertion limits provide assurance of achieving hot shutdown by reactor trip at any time and ensure that power peaking criteria are not exceeded. These limits preclude insertion of rod groups which could result in any single rod worth greater than the safety analysis assumption for the rod ejection transient. The physical design of the control rods has not changed nor has cycle operation, except for the change in position limits.

The simplified gray APSR position limits have been analyzed using worst case conditions and time of core life such that core peaking limits are not violated.

Based upon a worst case range of power distribution, shapes and peaking factors within the core which result in the most severe calculated consequences for the spectrum of postulated accidents, the revised linear heat rates of Technical Specification 3.5.2 continue to provide assurance that the fuel rod cladding temperature remains below the final acceptance criteria of 10CFR50, Appendix K. These parameters have been analyzed using NRC approved ECCS evaluation codes and methodology, referenced-in the Cycle 10 Reload Report, to ensu~e that the fuel pin cladding will remain intact during a LOCA and that the core remains in a safe configuration during Cycle 10 operations.

DISCUSSION OF E0C MTC CONCERNS B&W Fuel Company is in the process of evaluating their MTC calculational and measurement methodology because, with longer fuel cycles, more plants are approaching their FSAR end-of-cy.:le MTC limit. These evaluations have revealed a small non-conservative (positive) bias in the calculational 2

methodology. _ As this bias has yet to be quantified, its effect is not reflected in the MTC values presented in Table 5-1 of the Reload Report.

Because of the potential impact of these results on the Cycle 9 (current s cycle) MTC related safety analyses, Entergy Operations has administrative 1y limited Cycle 9 operations to a minimum soluble boron concentration sufficient to conservatively ensure that MTC does not exceed the analysis limit.

Due to other considerations, Entergy Operations had already decided to l re-analyze the Main Steam Line Break using the latest approved methodology. I Based upon similar work performed for other plants, it is anticipated that a significant relaxation in the end-of-cycle MTC limit will result and will more than compensate for the MTC calculation bias described above. This ,

analysis is expected to be completed early in Cycle 10 (December 1990). The )

results and the effect on the end-of-cycle MTC limits will be described to the NRC staff following completion of this analysis.

'i DETERMINATION OF SIGNIFICANT HAZARDS Entergy Operations, Inc. has performed an analysis of the proposed change in r accordance with 10CFR50.91(a)(1) regarding no significant hazards i

! consideration using standards in 10CFR50.92(c). A discussion of those l standards as they relate to this amendment request follows:  ;

Criterion 1 - Does Not Involve Significant Increase in the Probability or l Consequences of an Accident Previously Evaluated j Section 7 of the Cycle 10 Reload Report presents the results of an  !

evaluation of accidents addressed in the ANO-1 FSAR. The evaluation demonstrates that changes in the fuel cycle design and the corresponding

- proposed Technical Specification changes do not involve a significant .{

increase in the probability or consequences of an accident previously evaluated. All of the radiological. consequences are lower than the NRC acceptance criteria of NUREG-0800. The transient evaluation of Cycle 10 is bounded by previously: accepted analyses.

Criterion 2 - Does Not Create the Possibility of a New or Different Kind-of Accident from any Accident Previously Evaluated I

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The proposed changes would not create the possibility of a new or different kind of accident from any previously analyzed. The Technical Specification  ;

changes are minor limit and setpoint changes and result in no signific. int changes to the operation of the unit.

Criterion 3 - Does Not Involve a Significant Reduction in a Margin of j Safety The proposed changes do not involve a significant reduction in a margin of safety since, as shown in the Reload Report, Cycle 10 setpoints, safety limits and limiting safety system settings provide the same margins of 3

safety as previous core reloads. The NRC accepted methodology for establishing Technical Specification Limiting Conditions for Operation assure that the Final Acceptance Criteria ECCS limits will not be exceeded nor will the thermal design criteria be violated.

The Commission has provided guidance concerning the application of these standards by providing examples in 51 F.R. 7750. The proposed amendment is most closely encompassed by Example (iii): "A change resulting from a core reloading, if no fuel assenblies significantly different from those found previously acceptable to the NRC for a previous core at the facility in question are involved. This assumes that no significant changes are made to the acceptance criteria for the technical specifications, that the analytical methods used to demonstrate conformance with the technical specifications and regulations are_not significantly changed, and the NRC has previously '7und such methods acceptable."

Therefore, based on the reasoning presented above and the previous discussion of the amendment request, Entergy Operations has determined that the requested changes do not involve a significant hazards consideration.

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