ML17324B117
| ML17324B117 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 10/31/1986 |
| From: | Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17324B118 | List: |
| References | |
| AEP:NRC:0967E, AEP:NRC:967E, NUDOCS 8611070301 | |
| Download: ML17324B117 (11) | |
Text
REQUL*TORY INFORNATION DISTRIBUTION SYSTEH (RIDS) cg
~
ACCESSION NBR: 8611070301 C. DATE: 86/10/31 NOTARI2:
NO DOCKET FACIL: 50-315 Donald C.
Cook Nuclear Power Plant>
Unit i. Ind iana 5
05000315 AUTH. NANE AUTHOR AFFILIATION ALEXICH,N. P.
Indiana Sr Nichigan Electric Co.
RECIP. NANE RECIPIENT AFFILIATION DENTON> H. R.
Office of Nuclear Reactor Regulationz Direc tor (p ost 851125
SUBJECT:
Application for amend to License DPR-58> changing Tech Specs to extend deadline for surveillances normally performed during shutd own. Fee pai d.
DISTRIBUTION CODE:
AOOID COPIES RECEIVED: LTR J ENCL I SIZE: ~bi/
TITLE:
OR Submittal:
Qeneral Distribution NOTES:
RECIPIENT ID CODE/NANE PWR-A EB PWR-A FOB PWR-A PD4 PD 01 PWR-A PSB INTERNAL: ADli/LFNB NRR/DHFT/TSCB 04 COPIES LTTR ENCL 1
5 5
1 1
0 1
1 1
1 REC IP IENT ID CODE/MANE PWR-A EICSB PWR-A PD4 LA WIQQINQTONi D PWR-A RSB ELD/HDS3 NRR/ORAS RQN3 COPIES LTTR ENCL 2
2 1
0 1
1 0
1 0
1 EXTERNAL:
EQ8cQ BRUSNE> S NRC PDR 02 1
1 1
LPDR NSI C 03 05 1
1 1
TOTAL NUNBER OF COPIES REQUIRED:
LTTR 23 ENCL 19
(
n 3'n 'jtj >
( (
}'I PS J4 R
E C}h
-, E '}plhj(
/
4 INDIANA8 MICHIGAN ELECTRIC COMPANY P.O. BOX 16631 COLUMBUS, OHIO 43216 October 31, 1986 AEP:NRC:0967E Donald C.
Cook Nuclear Plant Unit No.
1 Docket No. 50-315 License No.
DPR-58 ADDITIONAL SURVEILLANCE INTERVAL EXTENSIONS FOR UNIT 1 CYCLE 9 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Denton:
This letter and its attachments constitute an application for amendment to the Technical Specifications (T/Ss) for the Donald C.
Cook Nuclear Plant Unit 1.
Specifically, we request an extension for certain surveillances which the Technical Specifications require to be performed beginning December 20, 1986.
A description of the proposed
- changes, the reasons for the changes, and our analyses concerning significant hazards considerations are contained in Attachment 1 to this letter.
The proposed revised Technical Specification pages are contained in Attachment 2.
We are requesting an extension for a number of surveillances that are normally performed during shutdown.
While it is possible to perform these surveillances at power, we believe that it is not prudent to do so for the reasons stated in Attachment 1.
These surveillances are due beginning December 20, 1986 through July 31, 1987.
Therefore we request that this T/S change be reviewed on an expedited basis.
Some of the Technical Specification pages affected by this submittal are pages for which changes are pending due to prior submittals.
The proposed changes contained in this submittal are in addition to our previous requests, and do not supersede them.
The pages included in this category and the applicable prior submittals which have not yet been processed are provided in the table below.
Letter No.
Date Technical S ecification Pa e Nos.
AEP:NRC'0967D AEP:NRC:0895D AEP:NRC:0856J AEP:NRC:0856I October 1,
1986 January 27, 1986 October ll, 1986 May 19, 1986 3/4 3-12, 3/4 3-33, 3/4 3-13 3/4 3-56 3/4 3-56 3/4 3-13, 3/4 3-15 3/4 3-33a, 3/4 3-56 sbii070301 8bi031 PDR ADOCK 050003i5 P
PDR gly-b14$
~,
'E C
~ ~
1 1
4 p
~
J 1
y4
\\
A T
~
I
'I ~
1'
Mq. Harold R. Denton AEP:NRC:0967E In this letter we are requesting an exemption for some of the same T/Ss which were addressed in our letter AEP:NRC:0967D.
The reason for this additional request is that different justifications for the changes are being supplied in this submittal.
We believe these new justifications are appropriate since the sameT/S changes would now be applied to different instruments than those of concern in AEP:NRC:0967D.
For several items in our previous letter we'requested an exemption for only portions of the equipment required by the T/Ss (Table 4.3-1 Items 11, 14, 15; Table 4.3-2 Items 5.a, 6.a, 7.a; Table 4.3-7 Items 5, 7).
For it'ems which we are requesting additional exemptions, as mentioned in Attachment 1 to this
- letter, we have included a reference to Section 4.0.6 on the T/S pages in Attachment 2,
even if an exemption was requested in our letter AEP:NRC:0967D.
We believe that these proposed changes will not result in (1) a significant change in the types of effluents or a significant increase in the amount of any effluent that may be released offsite, or (2) a significant increase in individual or cumulative occupational radiation exposure.
These proposed changes have been reviewed by the Plant Nuclear Safety Review Committee (PNSRC) and will be reviewed by the Nuclear Safety and Design Review Committee (NSDRC) at their next regularly scheduled meeting.
In compliance with the requirements of 10 CFR 50.91(b)(1),
copies of this letter and its attachments have been transmitted to Mr. R.
CD Callen of the Michigan Public Service Commission and Mr. G. Bruchmann of the Michigan Department of Public Health.
I Pursuant to 10 CFR 170.12(c),
we have enclosed an application fee of
$150.00 for the proposed amendments.
This document has been prepared following Corporate procedures which incorporate a reasonable set of controls to insure its accuracy and completeness prior to signature by the undersigned.
Very truly youris, cm Attachments M.
. Al ich
~p,P Vice president S
i
~ (g(p lOP cc:
John E. Dolan W.
G. Smith, Jr.
- Bridgman R.
C. Callen G.
Bruchmann G. Charnoff NRC Resident Inspector
- Bridgman
Attachment 1 to AEP:NRC:0967E Reasons and Significant Hazards Considerations for Changes to the Technical Specifications for D.
C.
Cook Unit 1
Attachment 1 to AEP:NRC:0967E Reasons and Significant Hazards Considerations for Changes to the Technical Specifications for D.
C.
Cook Unit 1
AEP:NRC:0967E Attachment 1
Page 1
This proposed amendment requests an extension for surveillances that must be performed between December 20, 1986 and July 31, 1987.
This request is a supplement to our letter AEP:NRC:0967D, dated October 1, 1986.
As was done in AEP:NRC:0967D, we propose to reference the following Section 4.0.6:
4.0.6 By specific reference to this section, those surveillances which must be performed on or before July 31,
- 1987, and are designated as 18-month surveillances (or required as outage-related surveillances) may be delayed until the end of the Cycle 9-10 refueling outage (currently scheduled to begin during the second quarter of 1987).
For these specific surveillances under this section, the specified time intervals required by Specification 4.0.2 will be determined with the new initiation date established by the surveillance date during the Unit 1 1987 refueling outage.
We reference this Specification by footnote in all surveillances that require this extension.
This footnote will be applicable to the following Technical Specifications with the indicated surveillance due date.
Dates given include the grace period allowed by T/S 4.0.2.
Table 4.3-1, Items 7, 8,
9, 10, 11, 14, 15 Descri tion of Chan e
Delays portions of the following channel calibrations:
steam generator level steam generator mismatch pressurizer level overtemperature/overpressure delta T
pressurizer pressure Due Date 12/20/86 12/20/86 12/23/86 01/31/87 04/16/87 Table 4.3-2, Items l.c, l.d, l.e, l.f, 2.c, 3.b.3, 4.c, 4.d, 5.a, 6.a, 6.b, 7.a, 8.a, 8.b 4.3.2.1.2 Delays portions of the following channel calibrations:
steam generator level steam generator mismatch lower containment pressure steam line pressure overtemperature/overpressure delta T
P-ll, P-12 interlocks 4kV loss of voltage pressurizer pressure 12/20/86 12/20/86 12/27/86 12/31/86 01/31/87 01/31/87 02/08/87 04/16/87 Table 4.3-6 Items 2, 3,
5 Delays portions of the following channel calibrations:
steam generator pressure pressurizer level pressurizer pressure 12/20/86 12/23/86 04/16/87 Table 4.3-7 Items 1, 5,
6, 7
4.4.1l.l.b Delays portions of the following channel calibrations:
steam generator level pressurizer water level containment pressure steam line pressure Delays PORV channel calibration 12/20/86 12/23/86 12/27/86 12/31/86 04/16/87
AEP:NRC:0967E Attachment 1
Page 2
We are requesting an extension for the performance of channel calibrations on the following instrumentation:
steam generator level, steam generator
A channel calibration consists of calibrating the primary sensor, and alarm and trip-functions.
Also during a calibration, a signal is injected into the first instrument after the primary sensor, and the appropriate bistables and/or alarms that feed the Solid State Protection System (SSPS) are checked.
During the extension period the normal channel functional tests and channel checks will be performed.
During channel functional tests, a simulated signal is injected into the channel as c'lose to the primary sensor as practicable to verify operability.
The channel functional tests currently performed on 'this instrumentation are far more stringent than required.
These tests not only demonstrate channel functionality, but also verify calibration of trip setpoints, actuations, and alarms.
The only portion of the channel that is not tested is the sensor.
The operability of the sensor is qualitatively verified each shift by channel checks which compare the output of independent sensors that measure the same parameter.
We believe that performance of the applicable channel functional tests and channel checks is adequate to ensure operability of this instrumentation during the extension period.
In addition, the primary sensors for the steam generator level and mismatch, pressurizer pressure and level, steam line pressure and delta T/Tavg instrumentation, (which have a qualified life of 10 years) were replaced during the 1985 Unit 1 refueling outage.
These sensors were extensively tested by means of a static alignment and a bench check prior to installation, and were calibrated after installation.
Performance of these channel calibrations involves operational difficulties, in particular, an increase in probability of a reactor trip (or Engineered Safety Features Actuation which could cause a reactor trip or other severe operating difficulty).
During the performance of these channel calibrations, the trip logic is reduced.
These channel calibrations generally require several shifts to complete, during which time one erroneous trip signal to any of the remaining channels will cause a reactor trip, thus causing the plant to operate with a higher than normal risk of a reactor trip.
Another problem with performing these surveillances is that many of them have never been performed at power and do not currently have procedures to do so.
While most of the current procedures can be revised to include the steps necessary to complete the surveillance at power, some may present operational difficulties that are not known at this time.
In addition, performance of some of these surveillances at power may involve a decrease in personnel safety.
Several of these surveillances involve testing equipment at much higher temperatures and pressures than when done during shutdown.
This presents the potential for increased risk of personnel accidents or injury.
AEP:NRC:0967E Attachment 1
Page 3
The above justification applies to all Reactor Trip System Instrumentation and all Engineered Safety Feature Actuation System Instrumentation (Tables 4.3-1 and 4.3-2) discussed in this letter.
The justification also applies to the Remote Shutdown Monitoring Instrumentation,and the Post-Accident Monitoring Instrumentation (Tables 4.3-6 and 4.3-7) except that channel functional tests are not performed and channel checks are done monthly.
Also included in this application is T/S 4.3.2.1.2, Engineered Safety Features logic interlocks (P-11 and P-12).
The interlocks are tested each month by an automatic actuation logic test.
This extension is requested because the 18 month total interlock function is tested during our delta T/Tavg and pressurizer pressure channel calibrations.
10 CFR 50.92 Criteria Per 10 CFR 50.92, a proposed amendment will not involve a significant hazards consideration if the proposed amendment does not:
(1) involve a significant increase in the probability or consequences of an accident previously analyzed, (2) create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or (3) involve a significant reduction in a margin of safety.
Our evaluation of the proposed change with respect to these criteria is provided below.
Criterion 1
The applicable channel functional tests and channel checks will ensure that these systems will perform as designed.
As stated
- above, many of the sensors were replaced in the last outage.
We believe the extension we are requesting will not result in deterioration to the extent the sensors cannot perform their intended function.
Therefore, we believe the extension will not result in a significant increase in the probability or consequences of-a previously evaluated accident.
Criterion 2 We believe that this extension will not result in a change in plant configuration or operation.
The applicable channel functional tests and channel checks will demonstrate that no significant degradation to the systems has occurred.
Therefore, the extension we have requested will not create the possibility of a new or different kind of accident from any previously evaluated or analyzed.
Criterion 3
The actions described above ensure the operability of the equipment involved.
We believe this surveillance interval extension will not significantly affect the ability of the equipment to perform its safety function.
Therefore, this change will not result in a significant reduction in a margin of safety.
AEP:NRC:0967E Attachment 1
Page 4
Lastly',
we note that the Commission has provided guidance concerning the determination of significant hazards by providing certain examples (48 FR 14870) of amendments considered not likely to involve significant hazards consideration.
The sixth of these examples refers to changes which may result in some increase to the probability or consequences of a previously evaluated accident, but the results of which are within limits established as acceptable.
We believe this change falls within the scope of this example, for the reasons cited above.
- Thus, we believe this change does not involve a significant hazards consideration as defined in 10 CFR 50.92.