ML17326B304
| ML17326B304 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 01/11/1988 |
| From: | Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | Murley T NRC, NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| Shared Package | |
| ML17326B305 | List: |
| References | |
| AEP:NRC:0916AF, AEP:NRC:916AF, NUDOCS 8801130230 | |
| Download: ML17326B304 (29) | |
Text
REQULAT INFORMATION DISTRIBUTION STEM <RIDS)
ACCESSION NBR: 8801130230 DOC. DATE: 88/01/11 NOTARIZED:
NO DOCKET 0 FACIL: 50-316 Donald C.
Cook Nuclear'oeer Plant'nit 2i Indiana 5
05090316 AUTH. NAME AUTHOR AFFILIATION ALEXICH> M. P.
(formerly Indiana 8c Michigan Ele REC IP. NAME RECIP lENT AFFlLIATION MURLEYiT. E.
NRC No Detailed Affi1iation Given
SUBJECT:
Application for amend to License DPR-7@i extending surveillances that Tech Specs require to be performed beginning 880302 until Unit 2 refueling outage in June 1'F88.
NRC response by 880226 requested.
DISTRISUTION CODE:
AOOID COPIES RECEIVED: LTR J ENCLf SIIE; Ot TITLE:
OR Submittal:
General Distribution NOTES:
RECIPIENT ID CODE/NAME PD3-3 LA NIQGINQTONi D COPIES LTTR ENCL 1
0 1
REC IP IENT ID CODE/NAME PD3-3 PD COP IES LTTR ENCL 5
5 INTERNAL: ARM/DAF/LFMB NRR/DEST/CEB NRR/DEST/RSB ILRB REG FILE 01 EXTERNAL:
LPDR NSIC 1
0 1
1 1
1 1
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NRR/DEST/ADS NRR/DEST/MTB NRR/DOE*/TSB OQC/HDS1 RES/DE/EIB NRC PDR 1
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TOTAL NUMBER OF COPIES REQUIRED:
LTTR 20 ENCL 17
0 h'
Indiana Michigan Power Company P.O. Box 16631 Columbus, OH 43216
à AEP:NRC:0916AF Donald C.
Cook Nuclear Plant Unit 2 Docket No. 50-316 License No.
ADDITIONALSURVEILLANCE INTERVAL EXTENSIONS FOR UNIT 2 CYCLE 6 U.S. Nuclear Regulatory Commission Attn:
Document Control Desk Washington, D.C.
20555 Attn:
T.
E. Murley January ll, 1988
Dear Dr. Murley:
This letter and its attachments constitute an application for amendment to the Technical Specifications (T/Ss) for the Donald C.
Cook Nuclear Plant Unit No. 2.
Specifically, we request an extension for certain surveillances that the T/Ss require to be performed beginning March 2, 1988.
We are requesting relief from these T/S requirements until the Unit 2 refueling outage, which is expected to begin by early June 1988.
The nature of these surveillances is such that they cannot be performed at power without violating T/S requirements or placing the unit or personnel in jeopardy.
Therefore, to avoid unnecessary
- shutdown, we ask that your review of this request be performed on an expedited basis and that you respond to us by February 26, 1988.
This is the second of two submittals that request surveillance interval extensions for Unit 2 Cycle 6.
The changes requested in this letter supplement the extension requests subroitted in AEP:NRC:0916AE, dated October 28, 1987.
Those changes were granted as Amendment 97 to the Unit 2 T/Ss.
The changes were divided into two separate submittals because the changes requested in our earlier submittal required the most immediate regulatory action in order to avoid unit shutdown.
In this submittal we are also requesting two changes to correct minor editorial errors.
A description of the proposed changes and our analyses concerning significant hazards considerations are contained in Attachment 1
to this letter.
The proposed revised T/S pages are contained in.
=. 8801130230 880iii U},
Dr. T.
E. Murley AEP:NRC:0916AF All of the requested surveillance extensions are associated with surveillances normally performed during refueling outages.
The extensions are necessary because the current cycle has been lengthened due to a self-imposed operating limit of 80% of rated thermal power and the occurrence of various unanticipated outages.
The significant outages which have occurred during the current Unit 2 cycle are described in the table below.
Date of outa e
Len th of outa e
Cause of outa e
March 3
- April 21, 1987 49 days Steam generator and ice condenser surveillance work July 14-22, 1987 8 days Problems experienced with main generator exciter August 27 October 11, 1987 45 days Problems experienced with RCP hatch evaluations/steam generator eddy current testing During the August 27, 1987 Unit 2 steam generator inspection
- outage, an effort was made to perform as many surveillances as possible.
Significant progress was made in reducing the number of T/Ss for which we need to request extensions.
Our progress was constrained because Unit 1 was in a refueling outage at the same time.
Therefore, other surveillances could not have been completed without significantly increasing the length of the Unit 2 outage.
If our request for extension is not granted, we would be required to shut down Unit 2 for a period of approximately four weeks in order to perform the surveillances addressed in this letter.
This would constitute a burden whi.ch would likely be placed on our customers.
Similar surveillance interval extensions were recently granted by the NRC for Cook Nuclear Plant Unit 1 via Amendment 100 to the Unit 1 T/Ss.
These extensions permitted us to continue operation into t'e Unit 1 Cycle 9-10 refueling outage, which ended in October 1987.
At your staff's request, we have reviewed the test results for those Unit 1 surveillances that were granted
Dr. T.
E. Murley AEP:NRC:0916AF extensions.
Nothing was found during this review that would indicate that the extensions requested in this submittal, or our previous submittal (AEP:NRC:0916AE), would result in the unit being in an unsafe condition.
Some of the T/S 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 are as follows:
Letter No.
Date T
S Pa e No.
AEP:NRC:0692AJ May 30, 1986 3/4 3-44 AEP:NRC:0931B July 22, 1987 3/4 7-20 AEP:NRC:0916W March 26, 1987 3/4 3-11 We believe that the proposed changes will not result in (1) a significant change in the types of effluents or a significant increase in the amount of any effluents 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.
C. Callen of the Michigan Public Service Commission and Mr. G.
Bruchmann of the Michigan Department of Public Health.
Pursuant to 10 CFR 170.12(c),
an application fee of $150.00 is necessary.
Please use check number 295-0220 dated October 22, 1987, which has previously been mailed to your office.
Dr. T.
E. Murley AEP NRC 0916AF This document has been prepared following Corporate procedures which incorporate a reasonable set of controls to ensure its accuracy and completeness prior to signature by the undersigned.
Sincerely, M. P. Al ich Vice President Attachments cc:
John E. Dolan V. G. Smith, Jr.
- Bridgman G. Bruchmann R.
C. Callen G. Charnoff NRC Resident Inspector
- Bridgman A. B.. Davis - Region III
Attachment 1 to AEP:NRC:0916AF Reasons and 10 CFR 50.92 Significant Hazards Evaluation for Changes to the Technical Specifications for Donald C.
Cook Unit 2
Attachment 1 to AFP:NRC:0916AF Page 1
As discussed in our cover letter, the purpose of this proposed amendment is to prevent an unscheduled outage before our next refueling outage, which is expected to begin by early June 1988.
This submittal requests extensions for surveillances that cannot be performed at power without violating T/S requirements or placing the unit or personnel in jeopardy.
The surveillances covered in this letter are due between March 2,
- 1988, and the expected entry into the refueling outage.
Our previous submittal for Unit, 2 Cycle 6 surveillance extensions, AEP:NRC:0916AE, requested relief for surveillances due by December 31, 1987.
The extensions were granted as Amendment 97 to the Unit 2 T/Ss.
Amendment 97 added the following T/S 4.0.7.
We are proposing to add the same reference to the surveillances addressed by this submittal.
4.0.7 By specific reference to this section, those surveillances which must be performed on or before July 1, 1988 and are designated as 18-month surveillances (or required as outage-related surveillances under the provisions of Specification 4.0.5) may be delayed until the end of the cycle 6-7 refueling outage (currently scheduled to begin during the latter part of the second quarter of 1988).
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 2 1988 refueling outage.
The specific T/Ss for which we are requesting relief in this submittal and the surveillance due dates are provided below:
(1) 4.5.2.d.j 4.5.3.1 (2) 4.7.7.1 Descri tion of Chan e
Delay RHR auto-closure interlock testing Delay steam generator snubber functional testing Due Date 03/02/88 03/09/88 (3) 4.1.3.3 Delay analog rod position 03/21/88 indication functional testing
Attachment 1 to AEP:NRC:0916AF Page 2
T S Affected (4) Table 4.3-1, Items 7 6 8
4.3.2.1.2 Table 4.3-2, Item 4.d Table 4.3-10, Items 2,3,
& 11 Descri tion of Chan e
Delay RTD calibrations Due Date 05/08/88 (5) Table 4.3-2, Items l.a, 2.a, 3.a.l, 3.b.l, 3.c.l, 4.a Delay testing of ESF manual
. 05/15/88 actuation switches (6) Table 4.3-1, Items 7,9,610 Table 4.3-2, Item ld 4.3.2.1.2 Table 4.3-6, Item 2
4.4.11.1.b Delay pressurizer pressure calibrations, interlock function testing, and PORV calibrations 05/18/88 The reasons and 10 CFR 50.92 analyses for each of the six surveillance groups is discussed separately below.
In addition to the surveillance interval extensions, we are also proposing two minor. editorial changes to correct errors in the present T/S pages.
The first of these changes adds the word "by" between the words "OPERABLE" and "the" in T/S 4.3.2.1.1.
The second change deletes a redundant "the" from T/S 4.5.2.g.
These changes are purely editorial in nature and will not increase the probability or consequences of a previously analyzed accident, create the possibility of a new or different kind of accident, or involve a significant reduction in a margin of safety.
Attachment 1 to AEP:NRC:0916AF Page 3
1 RHR Auto-Closure Interlock We are requesting an extension for the residual heat removal (RHR) auto-closure interlock test required by T/S 4.5.2.d.l.
An extension is also requested for T/S 4.5.3.1 since it references T/S 4.5.2.
The RHR auto-closure interlock automatically isolates the RHR system from the RCS if RCS pressure is above 600 psig.
In order to demonstrate operability of the auto closure interlock, it is necessary to open the RHR isolation valves in the cooldown line from the hot leg in order to verify that the valves would automatically close with the RCS pressure above 600 psig.
This cannot be accomplished with the unit operating (i.e., with the RCS fully pressurized) because it would result in exposing the RHR system to pressures higher than the RHR safety valves'etpoints.
Previous surveillance testing has demonstrated that the auto-closure interlock is very reliable.
The previous test results give us no reason to believe the auto-closure interlock would be inoperable during the extension period.
The calibration for the RCS wide-range pressure transmitters, which provide input into the interlock, can be done at power and will be performed by the Harch 2,
- 1988, due date.
- Thus, the only portion of the interlock for which surveillances will not be current is the portion from the bistable of the RHR suction valves through valve operation.
Additionally, we note that when the unit is operating (i.e., not on RHR), the RHR suction valves are closed and procedures require power to be removed from the valve operators.
This precludes inadvertent valve opening and thus alleviates the need for the auto-closure interlock to function.
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.
Attachment 1 to AEP:NRC:0916AF Page 4
Criterion 1 The surveillance test history of the auto-closure interlock has shown that the system is highly reliable, and gives us no reason to believe the equipment would be inoperable during an extension period.
The wide-range pressure transmitters, which provide input into the auto-closure interlock, will have a current calibration.
Additionally, we note that when the RHR system is not in service, power is removed from the suction valve operators, thus preventing inadvertent valve opening and eliminating the need for the auto-closure interlock.
For these
- reasons, we believe the extension we are requesting will not result in a significant increase in the probability or consequences of a previously evaluated
- accident, nor will it result in a significant reduction in a margin of safety.
Criterion 2
This extension will not result in a change in plant configuration or operation.
Therefore, the extension should not create the possibility of a new or different kind of accident from any previously evaluated or analyzed.
Criterion 3
See Criterion 1, above.
- 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.
N
Attachment 1 to AEP:NRC:0916AF Page 5
2 Steam Generator Snubbers This change would delay functional testing of steam generator snubbers required by T/S 4.7.7.1.c.
The extension is needed from March 9, 1988, until the refueling outage.
The steam generator snubbers for which we need extension are those numbered 91 and 92 in T/S Table 3.7-9.
The extension is requested because these snubbers are inaccessible during power operation, and because T/S 4.7.7.1.c specifically requires the testing to be performed during shutdown.
Both snubbers required to be tested were selected randomly, i.e., neither of them are being tested as a result of a previous failure.
Based on past snubber tests, we believe that granting this change will not'ause a significant risk to the public health and safety.
The requirement to test the steam generator snubbers was established in 1983.
Thirteen of the thirty-two snubbers in Units 1 and 2 have been functionally tested, and of the thirteen tested only one failed, that being a failure to lock-up in compression.
The problem was not generic, and the snubber passed the subsequent retest in 1985.
Visual inspection of the steam generator snubbers per T/S 4.7.7.1.a is not required until after the scheduled outage start
- date, and for this reason no extension for T/S 4.7.7.1.a is requested.
We believe that the surveillance history of our visual inspections gives further support for the position that the change will not result in a significant risk to public health and safety.
Visual inspections have been performed on steam generator snubbers at the Cook Nuclear Plant since 1975.
These inspections are performed at least once per refueling cycle.
No problem or potential problem has been revealed by these inspections.
All snubbers have been found to be acceptable and no generic problems have developed.
A similar request for an extension for Unit 1 snubber surveillances was approved by the NRC on December 20, 1986, via Amendment 100 to the Unit 1 T/Ss.
The Safety Evaluation Report for that amendment required that the snubber functional testing surveillance requirements be revised to increase the snubber testing sample size at least in proportion to the increase in the length of the refueling cycle beyond 18 months.
We intend to impose this requirement on ourselves for the Unit 2 steam generator snubbers as well.
This will require us to perform functional testing of at least one more steam generator snubber during the upcoming refueling outage.
It should be noted that a
request to permanently increase the surveillance interval for
Attachment 1 to AEP:NRC:0916AF Page 6
snubber testing to 24 months has been submitted in our letter AEP:NRC:0931B, dated July 22, 1987.
10 CFR 50.92 Criteria Per'0 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
Thirteen steam generator snubbers have been functionally tested at the Cook Nuclear Plant since 1983 with only one failure, the cause of which was not generic.
Visual inspections have been performed on snubbers since 1975, revealing no problems or potential problems.
Based on this surveillance history, we have no reason to believe the steam generator snubbers will be inoperable during the extension period.
Thus, it is believed that this change will not result in a significant increase in the probability or consequences of a previously evaluated
- accident, nor will it significantly reduce a margin of safety.
Criterion 2
Delaying the snubber functional tests will not result in a change in plant design or operation.
Therefore, we believe that the change will not create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated.
Criterion 3 See Criterion 1, above.
- Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing
Attachment 1 to AEP:NRC:0916AF Page 7
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 of occurrence or consequences of a previously analyzed accident, but where the results are within the limits established as acceptable.
We believe this change falls within the.scope of this example, for the reasons cited above.
Therefore, we believe this change does not involve a signiricant hazards consideration as defined in 10 CFR 50.92.
Attachment 1 to AEP:NRC:0916AF Page 8
3 Rod Position Indication S stem This change would delay functional testing of the rod position indicator (RPI) cnannels required every 1S months by T/S 4.1.3.3.
The extension is needed from March 21, 1988, until the refueling outage.
Although T/S 4.1.3.3 is only applicable in Modes 3, 4, and 5, we believe relieE is needed from this T/S to continue operation in Modes 1 and 2 since T/S 3/4.1.3.2 requires the RPI channels to be operable in these modes.
The surveillance we perform to satisfy T/S 4.1.3.3 is far more stringent than the channel functional testing that the T/S requires.
Our test is actually a calibration of the RPI channels over the rod insertion range.
Since rods must be inserted to perform the calibration, it cannot be performed at power because to do so should violate the rod insertion limits of T/S 3.1.3.6.
We believe that the extension we are requesting will not result in significant risk to public health and safety.
The operability of the RPI channels is functionally verified once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per T/S 4.1.3.2 by comparison to the demand position indication system.
These comparisons would be expected to indicate significant degradation in the RPI channels.
Indication that the core is performing as designed is provided by the incore Elux
- maps, which are taken at least once every 31 effective full power dayg to satisfy the requirements of T/Ss 4.2.2.2 (F (Z)) and 4.2.3 (F
H).
Core performance is also indicated by the 3xcore detectors, which are used to measure the quadrant power tilt ratio per T/S 4.2.4.
These surveillances would also be expected to indicate discrepancies between indicated and actual rod position.
- Lastly, we note that our review oE the surveillance history of the RPI channels gives us no reason to suspect the channels would be inoperable during the extension period.
10 CFR 50.92 Criteria Per 10 CFR 50.92, a proposed amendment will not involve a signiEicant 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.
Attachment 1 to AEP:NRC:0916AF Page 9
Our evaluation of the proposed change with respect to these criteria is provided below.
Criterion 1
/
T/S-required comparison oE the RPI channels to the demand position indication system would be expected to indicate signiEicant degradation in the RPI channels.
In addition, other surveillances, such as the determination of the quadrant powe" tilt ratio and incore flux mapping, provide a comparison of core performance to design and would be expected to indicate significant deviations of the control rods from their indicated position.
Also, the RPI channel surveillance history is good and provides no reason to believe the channels would be inoperable during the extension period.
For all these
- reasons, we believe the change will not involve a significant increase in the probability or consequences of a previously analyzed accident and that it will not involve a signiEicant reduction in a margin of safety.
Criterion 2 The proposed change will not result in a change in plant conEiguration or operation.
- Thus, the change should not create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated.
Criterion 3
See Criterion 1, above.
- Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing 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.
Attachment 1 to AEP:NRC:0916AF Page 10 4
RTD Calibration" Extensions are requested for the calibration of resistance temperature detectors (RTDs).
The T/S surveillances involving the RTD calibration are listed below.
Des~cri tion 1.
Table.4.3-1, Item 7
Overtemperature delta T channel calibration 2.
Table 4.3-1, Item 8
Overpower delta T channel calibration 3.
4.3.2.1.2 Total interlock function test for P-12 4.
Table 4.3-2, Item 4.d Steam flow in two steam lines
-- high coincident with T.avg low-low channel calibrationg 5.
Table 4.3-10, Item 2
Reactor coolant outlet temperature THOT channel calibration 6.
Table 4.3-10, Item 3
Reactor coolant inlet temperature TCOLD channel calibration 7.
Table 4.3-10, Item 11 Reactor coolant system subcooling margin monitor channel calibration The extension is needed from May 8, 1988, until the refueling outage.
The extensions requested in this category are for the calibration of the sensors only.
The calibration procedure requires data to be taken at RCS temperatures ranging from 0
approximately 250 F through operating temperatures.
This procedure cannot be performed at power because of the low temperatures necessary for the calibration.
The channels involved with the RTDs are subject to T/S required channel checks and/or channel functional tests.
This testing, which will continue during the extension period, would be expected to provide indication of RTD drift.
We have found RTDs at the Cook Nuclear Plant to be very stable, and have not experienced significant drifting problems.
- Lastly, we note that the Unit 2
Attachment 1 to AEP:NRC:0916AF Page ll RTDs are new, having been replaced during the last refueling
- outage, which ended in July 1986.
(The replacements were, made in order to satisfy equipment qualification requirements, and not because of problems with the previous RTDs.)
This increases our confidence in the dependability of the devices.
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.
Criterion 1 The RTDs at the Cook Nuclear Plant have traditionally been very stable.
In addition, the devices are new, having been replaced during the last refueling outage.
Channels involving the RTDs are subject to T/S-required channel checks and/or channel functional
- tests, which will continue to be performed during the extension period.
For these
- reasons, we believe the extension we are requesting will not result in a significant increase in the probability or consequences of a previously evaluated
- accident, nor will it result in a significant reduction in a margin of safety.
Criterion 2 This extension will not result in a change in plant configuration or operation.
Therefore, the extension should not create the possibility of a new or different kind of accident from any previously evaluated or analyzed.
Criterion 3
See Criterion 1, above.
- 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
Attachment 1 to AEP:NRC:0916AF Page 12 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.
Ve 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.
Attachment 1 to AEP:NRC:0916AF Page 13 5
ESF Manual Actuations Extensions are requested for the testing of ESF manual actuation switches specified in T/S Table 4.3-2, Table Notation (1).
The specific table entries for which extensions are requested are provided in the table below.
Table 4.3-2 Item No.
l.a Safety injection, turbine trip, feedwater.isolation, and motor-driven auxiliary feedwater pumps 3.a.l 3.b.l Containment spray Containment isolation Phase "A"
Containment isolation Phase "B"
3.c.l Containment purge and exhaust isolation 4.a Steam line isolation The testing of the manual actuation switches is required per T/S Table 4.3-2 to be performed every 18 months during shutdown.
The extension is needed from May 15, 1988, until the refueling outage.
As indicated in Note 1 for T/S Table 4.3-2, the circuitry associated with manual actuation of ESF functions is channel functional tested monthly.
The only portion of the channel not tested is the manual actuation switches.
Previous surveillance testing, of the switches have shown them to be highly reliable; in
- fact, there has never been a failure of any of the ESF manual switches detected during surveillance testing of the switches in either unit.
Additionally, we note that the manual circuitry serves as a backup to automatic logic, which initiates the same ESF functions.
The automatic logic is subjected to T/S-required channel
- checks, channel functional tests, and channel calibrations to verify its operability.
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:
Attachment 1 to AEP:NRC:0916AF Page 14 (1) involve a significant increase in the probability or consequences of an acc'dent 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 surveillance test history of the ESF manual switches is excellent, indicating no failures of the switches in either unit.
The majority of'the manual circuitry is subject to a channel functional test on a monthly basis.
This channel functional testing will continue to be performed during the surveillance extension period.
Additionally, we note that the manual circuitry serves as a backup to automatic circuitry, which initiates the same ESF functions.
For these
- reasons, we believe the extension we are requesting will not result in a significant increase in the probability or consequences of a previously evaluated
- accident, nor will it result in a significant reduction in a margin of safety.
Criterion 2 This extension will not result in a change in plant configuration or operation.
Therefore, the extension should not create the possibility of a new or different kind of accident from any
,previously evaluated or analyzed.
Criterion 3
See Criterion 1, above.
- 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.
Ve 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.
Attachment 1 to AEP:NRC:0916AF Page 15 6
Pressurizer Pressure Calibrations We are requesting an 'extension for the performance of channel calibrations and interlock testing involving the pressurizer pressure instrumentation.
The affected T/Ss are as follows:
Re uirement 4.3.1.1.1, Table 4.3-1, Item 7
Calibration for over-temperature delta T
4.3.1.1.1, Table 4.3-1, Item 9
Calibration for pressurizer pressure-low reactor trip 4.3.1.1.1, Table 4.3-1, Item 10 Calibration for pressurizer pressure-high reactor trip 4.3.2.1.1, Table 4.3-2, Item ld Calibration for pressurizer pressure-low ESF actuation 4.3.2.1.2 Interlock total function testing 4.3.3.'5, Table 4.3-6, Item 2
Calibration of remote shutdown monitoring instrumentation for pressurizer pressure 4.4.11.1.b Calibration of power operated relief valves The extensions are needed from May 18, 1988, until the'Unit 2
refueling outage.
Performance of this testing is not considered to be prudent at power.
The testing involves operational difficulties; in particular, testing involves an increase in the probability of a reactor trip (or Engineered Safety Features actuation that could cause a reactor trip or other severe operating difficulty).
During the testing, the trip logic is reduced.
The testing generally would require several shifts to complete, during which time an erroneous signal to any of the remaining channels could cause a trip.
Attachment 1 to AEP:NRC:0916AF Page 16 In addition, performance of the testing at power may involve a decrease in pers'onnel safety.
Testing of equipment would be performed at much higher temperatures and pressures than normally encountered during testing with the unit shut down.
This presents the potential.for increased risk of personnel accidents or injury.
One final difficultywith testing at power involves the configuration of the pressurizer pressure instrumentation.
Two of the instruments (NPS-153 and NPP-153) share a common sensing line.
Testing one of the transmitters poses the risk of perturbing the input to the other transmitter, which could result in a trip.
The instrumentation channels for which we are requesting surveillance interval extensions are subject to T/S-required channel functional testing and/or channel checks.
The channel functional tests we perform 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, which is qualitatively verified during channel checks.
- Thus, the testing we will continue to perform would be expected to provide indication of the operability of the systems, and would provide indication of significant degradation.
- Lastly, we note that based on our review of the surveillance history for the equipment, we believe the equipment will remain operable during the extension period.
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.
Attachment 1 tc AEP: NRC: 0916AF Page 17 Criterion 1 The applicable channel functional tests and channel checks should ensure that these systems will perform as designed.
Additionally, based on the surveillance history" of the equipment, we believe that the equipment will remain operable during the extension period.
We therefore believe the extension we are requesting will not result in deterioration to the extent that the instrumentation cannot perform its intended function.
- Thus, the proposed change should not result in a significant increase in the probability or consequences of a previously evaluated accident, nor should it result in a s=gnificant reduction in
- a. margin of safety.
Criterion 2 The proposed change will not result in a change in plant configuration or operation.
- Thus, the change should not create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated.
Criterion 3 See Criterion 1, above.
- 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 that may result in some increase to the probabili'ty 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.