ML19207B855
| ML19207B855 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 08/31/1979 |
| From: | Trimble D ARKANSAS POWER & LIGHT CO. |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| 1-089-19, 1-89-19, NUDOCS 7909050433 | |
| Download: ML19207B855 (4) | |
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.'.h ARKANSAS POWER & LIGHT COMPANY POST OFFICE BOX 551 UTTLE ROCK, ARKANSAS 72203 (5011371-4000 August 31, 1979 l-089-19 Director of Nuclear Reactor Regulation ATTN:
Mr.
R.
W.
Reid, Chief Operating Reactor Branch #4 U.
S.
Nuclear Regulatory Commission Washington, D.
C.
20555
~
Subject:
Arkansas Nuclear One-Unit 1 Docket No. 50-313 License No.
DPR-51 Long - Term Generic Issues Related to May 17, 1979 Order (File:
1510)
Gentlemen:
In accordance with the request of Dr.
D.
F.
Ross' letter of Aucust 21, 1979, we have reviewed Enclosure 1 of that letter and provide the following responses to Items 1,
4, 5,
7 and 8.
Item 1 The failure modes and effects analysis of the Integrated Con-trol System (ICS) was provided via letter from James H. Taylor to Dr.
D.
F.
- Ross, Jr.,
dated August 17, 1979.
The report, entitled " Integrated Control System Reliability Analysis",
also includes a reliability assessment of the ICS plant operating experience.
We have reviewed this report and basi-cally endorse it as applicable to our system.
Specific areas of difference are limited and will be addressed in our response to necessary system or procedural changes, if your review should come to that conclusion.
Our operating experience has lead us to believe the ICS is a reliable control system.
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.. _,g l-089-19 Mr.
R.
W. Reid August 31, 1979 Item 4 The AFW/EFW system reliability study for ANO-1, as discussed in the B&W Owners Group meeting with your staff, is currently underway by B&W.
The study for Rancho Seco (the lead plant) will be available in draft form by September 17, 1979.
The reliability study for ANO-1 will be available in draft form by October 22, 1979, with the final report being published by December 3, 1979.
Item 5 An analyses of the thermal-mechanical conditions in the reactor vessel during recovery from small breaks with extended loss of all feedwater will be available for release by B&W by December 21, 1979.
Item 7 The following item numbers refer to those of Attachment A to the previously referenced Ross letter.
Item lA:
A benchmark analysis for sequential auxiliary feedwater flow using CRAFT-2 with a 3 node steam generator representation will be sub-mitted by December 1, 1979.
Item 1B:
Justification for the relief and safety valve flow models used in the CRAFT-2 code will be provided by Septmber 30, 1979.
Item 2A:
Justification that the 3 node steam generator model used in the CRAFT-2 analysis of small breaks is adequate for the prediction of steam generator heat transfer will be provided by September 30, 1979.
Item 2B:
In response to ACRS concerns in regard to reactor system response to a stuck open PORV for the case of a small break which causes the reactor system to pressurize the PORV setpoint, we propose three (3) options with preference as listed.
Option 1 - Provide a statement that no small break with auxiliary feedwater will pressurize the system to the PORV setpoint by September 30, 1979, or 31&iY 34
~
I l-089-19 Mr.
R.
W.
Reid August 31, 1979 Option 2 - Provide a qualitative assessment of the transient by December 30, 1979, or Option 3 - Provide an analysis of the core using a 0.01 ft2 break with no AFW/EFW by February 1, 1980.
Please identify to us by September 7,
- 1979, if Option 1 will not meet your requirements so that work can proceed to meet the above schedule for Option 2.
Item 3A:
The sources of non-condensible gases in the primary system will be provided by September 30, 1979.
Item 3B:
The effect of non-condensible gas on (1) con-densation heat transfer, (2) system pressure calculations, and (3) natural circulation flow will be discussed in a report provided by October 31, 1979.
Item 3C A description of operator actions and/or emer-gency procedures necessary to preclude intro-duction of significant quantities of non-con-densible gases into the primary system will be provided by September 30, 1979.
Item 3D:
A description of operator actions to be taken in the event of a significant accumulation of non-condensible gases in the primary system will be provided by October 31, 1979.
Item 4:
The TMI analysis already performed which covers the first 100 minutes will be provided by September 30, 1979.
Item 5:
An evaluation of the recent Semiscale small break experiment with the B&W small break com-puter program will be performed and provided by September 30, 1979.
Item 6:
Pretest calculations of the LOFT small break test will be performed and provided by May 1, 1980.
Item 8:
Analysis of loss of feedwater and other antici-pated transients originally discussed in Section 8.4.1 of NUREG-0560 has been superseded and expanded by Section 2.1.9 of NUREG-0578.
On
[ji*l.E.3S
s 1-089-19 Mr.
R.
N. Reid August 31, 1979 August 9, 1979, B&W and the B&W Owners Group discussed a conceptual program to develop operator guidelines for a spectrum of abnormal transients.
The seven (7) transients selected for evaluation require detailed interface between NSSS vendor and architect engineer equipment as well as con-sideration of plant specific interactions that must be considered.
While our original program intended to use event tree method-ology for preparation of these guidelines, we believe inclusion of safety sequence analysis (SSA) into the event tree methodology would yield enhanced information.
B&W is currently developing the details of this combined approach.
Therefore, consistent with a discussion with your Mr.
R.
Capra on August 28, 1979, we will now provide a written program outline and de-tailed schedule for completion of the Abnormal Transients Operating Guidelines by September 14, 1979.
We would like to take this opportunity to address your concern with the present working relationship between NRC and the members of the B&W Owners Group.
As.you know, the B&W Owners Group has functioned to the satisfaction of all in its previous years of its existence.
We have found in the past the communications tool known as the B&W Owners Group is most effective when used as a forum for joint discussion on generic issues.
Our experience has been that the members of the Owners Group and the NRC staff are agents for a successful working relationship.
There-fore, we desire to continue to function in the same manner of greatest mutual benefit unless otherwise directed.
Very truly yours, 0 $.
David C. Trimble Manager, Licensing DCT/DGM/ew Ot <
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