ML18030B299
| ML18030B299 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 04/18/1986 |
| From: | Whitaker M TENNESSEE VALLEY AUTHORITY |
| To: | Grace J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| References | |
| NUDOCS 8605120215 | |
| Download: ML18030B299 (8) | |
Text
TENNESSEE VALLEY AUTHORITY 5N 157B Lookout. Place "IJ April 18, 1986 U.S. Nuclear Regulatory Commission Region II Attention; Dr. J.
Nelson Grace, Regional Administrator 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323
Dear Dr. Grace:
BROWNS FERRY NUCLEAR PLANT UNITS 1, 2; AND 3 N
-OIE REGION II INSPECTION REPORT 50-259/86-05, 50-260/86-05, AND 50-296/86-0 SUPPLEMENTAL RESPONSE Please refer to R.
H. Shell's April 4, 1986 letter to you which provided our response to Inspection Report Nos. 50-259/86-05, 50-260/86-05, and 50-296/86-05 for Browns Ferry Nuclear Plant.
Enclosed is a supplemental response addressing Unresolved Items 86-05-01 (FSAR update) and -09 (radiation monitor setpoints) which we indicated would be submitted by April 17, 1986.
t On April 17, Mike Hellums of my staff and Art Johnson of your staff discussed
'a one day extension.
If you have any questions, please get in touch with R.
E. Rogers at FTS 858-2723.
To the best of my knowledge, I declare the statements contained herein are complete and true.
Very truly yours, TENNESSEE VALLEY AUTHORITY M. B.'hitaker, Deputy Director Nuclear Safety and Licensing Enclosure cc:
Mr. James Taylor, Director (Enclosure)
Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Washington, D.C.
20555 8605120215 8604%8 PDR ADOCK 05000259 8
PDR An EQual Opportunity Employer WFoi
'i
RESPONSE
NRC INSPECTION REPORT NOS.
50-259/86-05, 50-260/86-05, AND 50-296/86-05 JOHN A. OLSHINSKI'S LETTER TO S. A. WHITE DATED MARCH 5, 1986 SUPPLEMENTAL RE)
- )NSE Unresolved Item 86-05-01 IE Inspection Report 86-05 identifies an unresolved item related to past annual Final Safety Analysis Repoxt (FSAR) updates.
Specifically, the inspection report identified FSAR changes which appeared to change original FSAR commitments without documented justification ox safety evaluations.
TVA has reviewed the four changes specified in IE Inspection Report 86-05 paxagxaph 5.a.
As discussed
- below, TVA has continued to maintain the original commitment.
1.
Section 5.3.5.1 of the FSAR was changed to state that secondary containment relief panels m~a be visually inspected xather than are visually inspected as originally stated.
Mechanical Maintenance Instruction (MMI)-14 pxesently requires periodic visual checks of the secondaxy containment relief panels.
2.
Section 3.8.5 of the FSAR was changed to state that the gas pressure in the standby liquid control system accumulators can be measured rather than is measuxed as previously stated.
MMI-40 presently requires the gas pressure to be checked periodically.
3.
Section 5.3.5.3 of the FSAR was changed to state that the cooling water supply to the equipment area cooling units was initially tested with.
emergency equipment cooling water and can now be tested rathex than is now tested as previously stated.
Technical Instruction-33 checks the cooling watex supply to the equipment area cooling units periodically.
4.
Section 7.5.4.2.5 of the FSAR was changed to state
- that, a reactor trip signal to be generated by source xange monitors (SRMs) m~a be used for the performance of core alterations xather than is used as previously stated.
General Operation Instruct,ion 100-3 requires the SRMs to be connected in reactor protection system scram mode during a refueling outage.
The procedure (Standard Practice 1.13) that'will be used for the next FSAR
'pdate specifies that each change be formally evaluated and justified.
Oux evaluation can find no NRC requirement for a specific safety evaluation for each FSAR change.
Refexring to 10 CFR 50.71(e) in the Fedexal Register dated May 9, 1980, the supplementaxy information published on page'30615
- states, "No analyses other than those alxeady prepared or submitted pursuant to NRC requirements (eithex originally with the application, or as part of the operating license review process, ox as required by 50.59 or other NRC
requirements, or to support license commitments) are required to be performed by the licensee because of this rule."
TVA interprets this statement to indicate NRC recognition that there are other mechanisms which adequately control the change process for nuclear power plants.
Xn the case of the four identified changes in the inspection report, changes to the implementing procedures would have received the Plant Operations Review Committee review and evaluation for the need for a unresolved safety question determination as a minimum.
As a result, we find it unlikely that the past Final Safety Analysis Report update process would cause the facility to be in an unanalyzed configuration.
We believe our practices on this matter are consistent with NRC intentions 'for 10 CFR 50.71 requirements, and appear to be consistent with
~ other utilities approach to this issue.
Unresolved Item 86-05-09 Browns Ferry is not committed to a specific methodology for instrument setpoint determination such as that contained in Regulatory Guide 1.105 "Instrument Setpoints for Safety-Related Systems."
The normal pxactice is to compensate for the manufacturer's stated accuracy and this may be augmented by observed calibration experience.
V An exception to this practice involves the calibration technique for the radiation monitors mentioned in the inspection repoxt.
For this case, full compensation for manufacturex's accuracy is not practical since the accuxacy of the instrument to measure mr/hr is comparatively poor compaxed to its ability to respond to increases in count, rate.
If these inaccuracies were treated in a classical manner (such as suggested in Regulatory Guide 1.105) the calculated setpoint would be much lower.
These ponitors typically xead about 20-30 mx/hx due to an internal bug source.
A combination of this background source reading and a fully compensated setpoint would encroach upon the normal operational margin to the instrument trip point.
In turn, this would increase the likelihood of spurious safeguard system actuations.
The purpose of this xadiation monitor is to detect gross radiation increases.
In view of this and based on the preceding discussion, we believe it is appropriate not to apply the standard approach to setpoint determination for these instruments.
The current procedure sets these instruments at 92 mr/hr to provide margin to the technical specification value between calibxation intervals.
This semiempixical approach accomplishes the following:
1.
maintains a setpoint below the technical specification value.
2.
provides a reasonable operating margin to minimize spurious safeguard system actuations.
3.
is set low enough to assure a high degree of confidence that abnormal radiation releases are detected and mitigated.
We are in the process of converting to a more xigoxous approach to establishing field setpoints.
The program will utilize the Instrument Society of Amexica standaxd 67.04.