1CAN089002, Revised Application for Amend to License DPR-51,increasing Reactor Power to Level of 100% & Borated Water Storage Tank Tech Spec Change
| ML20059D534 | |
| 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 | |
| ML20059D538 | List: |
| References | |
| 1CAN089002, 1CAN89002, NUDOCS 9009070087 | |
| Download: ML20059D534 (5) | |
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- W t u u m nx3 August 8, 1990 1CAN089002 U. S. Nuclear Regulatory Commission Document Control Desk Mail Station P1-137 Washington, DC 20555
Subject:
Arkansas Nuclear One - Unit 1 i
Docket No 50-313 License No. DPR-51 I
Revised Request for License Amendment to Increase Reactor Power to a Level of 100%
i and Borated. Water Storage Tank Technical Specification Change Gentlemen:
i On September 26, 1989, Entergy Operations transmitted to you a' Request for License Amendment to Increase Reactor Power to a Level of 100% (ICAN098903).
Due to events that have occurred since that time, this letter is being.
l written to supersede the September 26, 1989 request.
Facility Operating License Amendment No.120, transmitted by NRC letter dated May 16, 1989 (1CNA0589P4), authorized operation of AN0-1 up to a maximum steady-state reactor core power level of 2054 megawatts thermal _ (80%.
1 l
of full power).
That amen /nent was in response to our request dated-l April 24,1989 (1CANG48dM) as supplemented on May 5, 1989 (1CAN058903),
addressing a newly identified postulated small break in the High Pressure L
l Injection (HPI) hystem which was found to be not bounded by existing small break Loss ci Coolant Accident (LOCA) analyses.
The basis for the amendment was a fo; mal Appendix K LOCA Evaluation Model, of which-the analysis results were submitted, demonstrating that the ANO-1 HPI configuration would provide-adequate core cooling in the event of a complete HPI line break at an operating power of-80% of full power.
I Additionally, while performing a review of the Decay Heat Removal' System as 1
part of the ANO Design Configuration Documentation Project, several calculation errors and inconsistencies were identified, which together incorrectly characterized the flow capabilities of the Low Pressure Injection (LPI) and Reactor Building Spray (RBS) pumps when the pumps would l
be aligned to take suction from the Reactor Building sump.
Specifically, L
the adequacy of the Net Positive Suction Head (NPSH) was in question under l
these circumstances.
While restricted to 80% power, adequate margin existed l
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to assure the pumps would function properly under Design Basis Accident t
conditions.
This event was reported in Licensee Event Report 50-313/89-044-00 dated January 15, 1990 (1CAN019008);
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August 8, '.990 To allow resumption of full power operation (2568 megawatts thermal),
Entergy Operatiens is implementing changes and permanent system 4
modifications to 6ddress both the postulated HPI line break and the LPI and RBS NPSH concerns during the upcoming ninth refueling outage (IR9).
A considerable effort has been made to address all aspects of the HPI small break LOCA an lysis.
An extensive modification is being. implemented for the HPI system to Jd four new injection lines each with individual safety grade flow instrumentation.
The new design'has been modeled in detail to assess the system response to a spectrum of RCS breaks.
This system modification, along with providing acceptable H ystem performance, will provide the control room operator with specif lowrate information for proper assessment and respsnse to any small break LOCA. 'Our reanalysis conforms to the original Appendix K LOCA Evaluation Model, i
The initial findings of the Reactor Building (RB) sump NPSH issue have been thoroughly reviewed and the interrelationships between associated analyses have been identified.
A reanalysis of a number of complex analyses has been performed including:
Revisions of the BWST inventory, post accident RB water level, and RBS and LPI NPSH calculations.
Full computer re-analysis of the post accident RB pressure and temperature profiles (C0PATTA).
Complete re-analysis of the post LOCA offsite dose utilizing an updated l
methodology consistent with a recent Standard Review Plan.
l Re-analysis of the RBS and RB sump pH calculations consistent with the l
Standard Review Plan, t
ihese re-analyses now provide updated, consistent design bases for ANO-1 that resolve the concern identified, and verify acceptable performance of the ECCS and RBS systems to allow 100% full power operation.
l Therefore, continued limitation of ANO-1 operation below the 100% power level will no longer be necessary af ter completion of the modifications and tests scheduled for implementation during the 1R9 Refueling Outage.
Engineering Reports are enclosed detailing the analysis results for the HPI line break and the RBS and LPI pump NPSH resolution.
These results The information contained in the HPI line break engineering considered proprietary to its author, Babcock and Wilcox (B&W) port is
, since it contains methods and data specific to B&W ECCS analysis which would provide a competitor an unfair advantage if used.
Due to the proprietary nature of B&W has requested that it be withheld from public disclosure with 10CFR2.790 a non proprietary version of this material and an affidavit In accordance from B&W containing a full statement of the reasons separate cover by B&W.
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i U. S. NRC Page 3 August 8, 1990 In advance of modification implementation and predicated upon the successful post-implementation testing of the design changes, Entergy Operations requests a license amendment to resume full power operation at steady-state reactor core power levels not in excess of 2568 megawatts. We request that this be granted prior to the end of 1R9, currently scheduled for completion on December 1, 1990.
We are available to meet with the NRC Staff in your of fices should further clarification be necessary.
The License amendment request, associated Technical Specification changes, and the basis for these changes are attached.
There is currently a pending Technical Specification change dated June 14, 1990 (ICAN069002) relating to the attached change to Specification 3.1.2.10.
In accordance with 10CFR50.91(a)(1), and ~using the criteria in 10CFR50.92(c), Entergy Operations has determined that the change involves no-significant hazards consideration.
Very truly yours, N. S. Carns NSC:lw Attachments cc:
Mr. Robert Martin U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76011 NRC Senior Resident Inspector Arkansas Nuclear One - ANO-1 & 2 Number 1, Nuclear Plant Road Russellville, AR 72801 Mr. Thomas W. Alexion-NRR Project Manager, Region IV/ANO 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 Project Manager, Region IV/ANO-2 U. S. Nuclear Regulatory Coinmission NRR Mail Stop 11-B-19 j
One White Flint North i
11555 Rockville Pike Rockville, Maryland 20852 1
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Page:4" August 8 ;19901 n
-Ms.GretaDieus,Directok=
1 Division ofLRadiation Control?
'and Emergency; Management
. Arkansas Department of Health-4815 West Markham; Street.-
'Little Rock,' AR 72201'
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I, N. S. Carns, being duly sworn, subscribe to and say that-I am Vice President, Operations ANO for Entergy Operations, Inc. ; that I have full authority to execute this oath; that I have read the document numbered ICAN089002 and know the contents thereof; and that to the best of my knowledge, information and belief the statements in it are true.
W N. 5. Carns SUBSCRIBED AND SWORN T0 beforo me, a Notary Public in and for the County and State above named, this b day of d1 /L an E 1990.
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