ML17324B090
| ML17324B090 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 10/01/1986 |
| From: | Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17324B091 | List: |
| References | |
| AEP:NRC:0967D, AEP:NRC:967D, NUDOCS 8610140316 | |
| Download: ML17324B090 (53) | |
Text
e REQULATORY INFORMATION DISTRIBUTION SYSTEN (RIDS)
ACCESSION NBR: 8610140316 DOC. DATE: 86/10/01 NOTARIZED: NO DOCKET FACIL: 50-315 Donald C. Cook Nuclear Power Planti Unit ii Indiana 0 05000315 AUTH. MANE AUTHOR AFFILIATION ALEX ICHOR N. P. Indiana 5 Ni ch 1 gan El ec tr i c Co.
RECIP. NANE RECIPIENT AFFILIATION DENTON> H. R. Office of Nuclear Reactor Regulation> Director (post 851125 SUB JECT: Application For amend to License DPR-58i requesting extension for certain surveillances required to be performed b eg inn ing 861220. Extension needed because Cycle 9 lengthened due to limit of operation at 90/ power. Fee paid.
DISTRIBUTION CODE: A001D TITLE: OR COPIES RECEIVED: LTR Submittal: General Distribution ENCL SI ZE: 3 NOTES:
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t IN DIANA 8 M I CH I G AN ELECTRIC COM PAN Y P.O. BOX 16631 COLUMBUS, OHIO 43216 AEP:NRC:0967D Donald C. Cook Nuclear Plant Unit No. 1 Docket No. 50-315 License No. DPR-58 SURVEILLANCE INTERVAL EXTENSION 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 for the Donald C. Cook Nuclear Plant Unit l.
Specifically, we request an extension for certain surveillances which the Technical Specifications require to be performed beginning December 20, 1986.
Many of these surveillances can only be performed during shutdown; therefore, to avoid unnecessary shutdown of the plant, we ask that your review of this request be performed on an expedited basis and that you respond to us by December 15, 1986 '
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.
Most of the surveillance extensions are associated with items required during the refueling outage. The reason for the change is that the length of the refueling cycle has been extended for reasons which were not anticipated and are described below. This surveillance extension is needed because Unit 1 Cycle 9 has been lengthened due to a self-imposed limit of operation at 90$ of rated thermal power. This limit was imposed as a precautionary measure following discovery of abnormal degradation of steam generators in D. C. Cook Unit 2. Although the abnormal degradation has thus far been limited to the Unit 2 steam generators, we have adopted the additional chemistry controls and reduced power operation in both Units 1 and 2 as a precautionary measure until the cause of the degradation can be determined as not affecting Unit 1. In
'addition to the cycle lengthening due to reduced power operation, Unit 1 Cycle 9 was also lengthened by an unanticipated turbine trip which occurred in May of 1986. The corrective action and recovery from this trip required a 41-day outage. Lastly, Unit 1 experienced an outage of 10 days because of an unanticipated turbine trip in July 1986.
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r Mr. Harold R. Denton AEP:NRC:0967D If extension for those surveillances which can only be performed during shutdown is not granted, we will be required to shut down the unit for approximately 37 days beginning on December 15, 1986. This 37-day outage period will occur at the beginning of our winter peak power period. The loss of D. C. Cook Unit 1 would result in a loss of revenue to the American Electric Power system of approximately twelve million two hundred thousand dollars ($ 12,200,000) during this 37-day period, a burden that would likely be placed on our customers.
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 submi.ttal 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 Ne Date Technical S ecification Pa e Nos AEP:NRC'0972 January 17, 1986 3/4 7-17 AEP:NRC:0931 August 23, 1986 3/4 7-28 AEP:NRC:0895D January 27, 1986 3/4 3-13 AEP:NRC:0856J October 11, 1985'ay 3/4 3-56 AEP:NRC:0856I 19, 1986 3/4 3-56 3/4 4-14 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. 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), we have enclosed an application fee of
$ 150.00 for the proposed amendments.
We will be requesting similar extensions, for surveillances which can be performed at power'ut may place the unit in jeopardy of unexpected trips, by a similar letter under separate cover. We have only included in this letter surveillance extension requests required between December 15, 1986 and February 28, 1987. An additional letter requesting surveillance extensions from March 1, 1987 to the expected entry into the 1987 refueling outage will be sent under separate cover. We have broken our request in this fashion because of the large number of T/S change requests included and because the T/S changes requested by this letter require the most immediate regulatory action to avoid unit shutdown.
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y Mr. Harold R. Denton AEP:NRC:0967D 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 undersigned.
Very truly yours, M. . Alexich Vite President ~ l~)4'~l pm Attachments 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:0967D Reasons and Significant Hazards Considerations for Changes to the Technical Specifications for D. C. Cook Unit 1
- I AEP:NRC:0967D Attachment 1 Page 1 As stated in our letter, the purpose of this proposed amendment is to prevent an unscheduled surveillance outage before our next refueling outage currently scheduled to begin on May 23, 1987. This submittal requests extensions, for surveillances that must be performed during shutdown or that present such operational difficulty that performing the surveillance is not practical at power. We propose to add the following requirement to Section; 4.0:
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.
Descri tion of Chan e Due Date (1) Table 4. 3-1 Delay steam generator 12/20/86 Items 14, 15 level channel calibration, including steam/feedwater Table 4.3-2 flow mismatch, turbine trip Items 5a, 6a, 7a and feedwater isolation, motor-driven and turbine-Table 4.3-7, driven feedwater pumps, and Item 7 post-accident monitoring instrumentation (2) Table 4.3-1 Delay pressurizer water 12/23/86 Item 11 level channel calibration for reactor trip and post-Table 4.3-7 accident monitoring instru-Item 5 mentation (3) 4.5.2.d.l .Delay testing of the 01/10/87 4.5.3.1 RHR/RCS interlocks
i'A' AEP:NRC:0967D Attachment 1 Page 2 Descri tion of Chan e Due Date (4) 4.7.8 Delay steam generator 12/29/86 snubber functional test Delay Grinnell small 02/17/87 bore snubber functional test (5) 4.4.6.1.b Delay channel calibration 01/01/87 of containment sump level 02/06/87 and flow monitoring instrumentation (6) 4.8.2.5.2.d Delay N-train battery 01/13/87 service test.
(7) 4.6.5.9 Delay divider barrier seal 02/05/87 test and inspection (8) Table 4.3-7 Delay PORV limit switch 02/08/87 Xtem 12 calibration (9) 4.5.3.1 Delay visual inspection of 02/12/87 4.5.2.d.2 containment sump (10) 4 '.9.2.b Delay reactor coolant pump 02/19/87 spray/sprinkler system testing and inspection (11) 4.8.1.1.2.b Delay diesel testing 02/20/87 4.8.1.2 including relief valve 02/27/87 4.4.11.3 testing and essential 02/28/87 4.7.4.l.b service water valve testing (12) Table 4.3-2 Delay auxiliary feedwater 02/23/86 Item 6.d pump testing including ESF actuation on a loss of main 4.7.1.2.b feedwater pumps
'I (13) 4.3.1.1.3 Delay time-response testing 02/26/87 4.3.2.1.3 for reactor trip instrument-ation and engineered safety features instrumentation A description of the proposed changes, the reasons for the changes, and our analyses concerning significant hazards considerations for each group of extension requests are given in the remainder of this attachment.
We also request a change to correct a typographical error on page 3/4 0-3. The phrase "Semiannually or every 5 months" should read "Semiannually or every 6 months." This change is purely administrative in nature and therefore does not increase the probability or consequences of a previously evaluated accident, does not create the possibility of a new or different accident from any previously evaluated, and does not reduce a margin of safety. Therefore this change does not involve a significant hazards consideration as defined by 10'FR 50.92.
AEP:NRC:0967D Attachment 1 Page 3 (1) STEAM GENERATOR LEVEL (2) PRESSURIZER LEVEL (3) RHR/RCS INTERLOCKS We are requesting an extension for the performance of channel calibrations on instrumentation for steam generator level, pressurizer level and RHR/RCS interlocks required by 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 and 7; and T/Ss 4.5.2.d.l and 4.5.3.1. A channel calibration consists of calibrating the primary sensor, alarm and trip functions. 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.
Channel functional tests simulate a signal into the channel as close to the primary sensor as practical 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 channel functional tests and channel checks is adequate to ensure operability of this instrumentation during the extension period. In addition, the primary sensors for this instrumentation, which have a qualified life of 10 years, were replaced during the 1985 Unit 1 refueling outage. These sensors were extensively tested by performing a static alignment and a bench check prior to installation, and were calibrated after installation.
Performance of these channel calibrations involves operational difficulties, including the increase in probability of a reactor trip. 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 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.
The above-mentioned calibrations pose additional operational difficulties. The steam generator level, Sets I and II, channel calibration is particularly difficult to perform at power because the feedwater level controller for the steam generator must be placed in manual for the duration of the test. This test takes an average of several shifts for both protection sets. During this time feedwater flow must be controlled by a dedicated operator.
The calibration of pressurizer level instrumentation Set III is also difficult to do at power. The primary sensor (NLP-153) has a common sensing line with two pressurizer pressure sensors (NPS-153 and NPP-153). The fluid in this line must be partially drained to perform this test. If all the fluid in the sensing line is lost, the two sensors not being tested (NPS-153 and NPP-153) will cause a reactor trip.
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AEP:NRC:0967D Attachment 1 Page 4 Testing interlocks between the reactor coolant system (RCS) and the residual heat removal (RHR) system is required every 18 months. T/S 4.5.2.d.l requires that this interlock perform its function when the RCS is above 600 psig. Therefore it is not possible to perform this test at power without it violating the T/S. This interlock is required because otherwise would be possible to open the valves that isolate the RCS from the RHR system. Opening these valves would result in the overpressurization and possible failure of the RHR system.
10 C R 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 channel functional tests and channel checks will ensure that these systems will perform as designed. As stated above, 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 channel functional tests and channel checks will demonstrate that no significant 'degradation to the system 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 have been taken to 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.
'I AEP:NRC:0967D Attachment 1 Page 5 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.
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AEP:NRC:0967D 4l Attachment 1 Page 6 (4) STEAM GENERATOR SNUBBERS AND GRINNELL SNUBBERS This change would delay steam generator snubber and Grinnell small-bore .
snubber functional tests required by T/S 4.7.8. The surveillance due date is December 29, 1986 for the steam generator snubbers and February 17, 1987 for the small-bore snubbers. We are requesting an extension of the surveillance deadline from the required date to the refueling outage currently scheduled to begin on May 23, 1987. The extension is being requested because both the steam generator snubbers and all but one of the small-bore snubbers are inaccessible during power operation, and T/S 4.7.8.c requires the testing to be performed with the reactor shut down. Extension is requested only for the functional tests, since the deadline for the visual tests is after the anticipated beginning of the scheduled refueling outage. The snubbers selected for testing during the next outage are as follows:
- 1. Reactor Coolant System: T/S Nos. 1, 2, 3, 7, and 15.
- 2. Component Cooling Water System: T/S No. 55.
- 3. Main Steam System: T/S No. 46.
- 4. Reactor Coolant Pump Seal Water Supply System: T/S No. 38.
- 5. Steam Generator Snubbers: T/S Nos. 83 and 86 on Steam Generator No. 12.
All samples are random, that is, no retests are the result of a previous failure.
there Based on past snubber tests, we believe that, if this change is will be no significant risk to the public health and safety. Ingranted, 1978, numerous small-bore snubbers manufactured by the Grinnell Co. were found to lock up at a rate higher than design specifications recommended due to the factory settings of lock-up and bleed rates. All the Grinnell small-bore snubbers were tested in 1978 and settings were adjusted as necessary. The requirement to test the large-bore steam generator snubbers was established in 1983 in conjunction with the new Technical Specifications. Six of the sixteen snubbers have been tested, and of the six tested, one failed to lock up in compression. The problem was not generic, and the snubber passed the subsequent retest in 1985.
As mentioned above, the visual inspections of snubbers would not be required until after the beginning of the next refueling outage, but belief that the surveillance history of our visual inspections gives furtherour it is support for the position that the change will not result in a significant risk to public health and safety. Visual inspections are performed on small-bore snubbers at least once per refueling period. Over 21 T/S inspections have been performed on the "accessible" and "inaccessible" snubbers, resulting in 7 unsatisfactory findings; that is less than 1$ of the total population (cumulative). Visual inspections have been performed on the steam generator snubbers since 1975. These inspections are performed at least once per
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I, a AEP:NRC:0967D Attachment 1 Page 7 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.
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 As noted above, since the 1978 test results, all Grinnell snubbers teste'd have been found operable. With regard to the steam generator snubbers, the one snubber failure was not caused by a generic problem. No generic problems have developed with regard to the snubbers. The visual inspections have revealed only 1% of the total number tested to be unsatisfactory for the Grinnell snubbers, and all visual inspections of the steam generator snubbers have been acceptable. On the basis of the past history of the snubbers, believed that this change will not result in a significant increase in the it is probability or consequences of a previously evaluated accident.
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 We believe that an extension of the time for surveillance requirements will not result in a significant reduction in the margin of safety for the reasons given in 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 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. Therefore, we believe this change does not involve a significant hazards consideration as defined in 10 CFR 50.92.
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AEP:NRC:0967D Attachment 1 Page 8 (5) CONTAINMENT SUMP INSTRUMENTATION This change will delay channel calibration of the containment sump level and flow monitoring system required by T/S 4.4.6.1.b. An extension of the surveillance deadline from January 1, 1986 to the refueling outage, currently scheduled to begin May 23, 1987, is requested. It is noted that T/Ss allow one of the three leak detection systems to be inoperable for 30 days. The extension is being requested because the instruments are located in areas of containment which are inaccessible during power operation due to ALARA concerns.
This system provides an early indication of RCS pressure boundary degradation.
We believe this change will not result in a significant risk to the public health and safety because of the number of backup systems available (humidity monitors, containment atmosphere gaseous radioactivity monitoring channels, the containment water level instrumentation, and RCS water inventory balances required by T/S 4.4.6.2.1 every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />).
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 evaluated, (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 instrumentation is required for both leak detection and for post-accident monitoring purposes. The change we are requesting is to defer a required calibration until the 1987 refueling outage. During this period, should a leak or accident occur, we believe that the intended safety function would be served by the alternative equipment mentioned above. Therefore, we believe this change will not result in a significant increase in the probability or consequences of a previously evaluated accident.
I Criterion 2 The change to extend the surveillance interval for channel calibration does not affect normal or accident plant operation nor decrease the level of protection available to prevent accidents. In an accident these instruments do not provide an actuation signal to safety equipment. Therefore we believe the change will not create the possibility of a new or different kind of accident from any previously evaluated.
AEP:NRC:0967D Attachment 1 Page 9 Criterion 3 We believe that an extension of time for surveillance requirements for channel calibration of the containment sump level and flow monitoring system will not result in a significant reduction in the margin of safety, for the reasons given in 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 of occurrence or consequences of a previously analyzed accident, but where the results are within limits established as acceptable. We believe this change falls within the scope of this example. Therefore, we believe this change does not involve a significant hazards consideration as defined in 10 CFR 50.92.
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AEP:NRC:0967D Attachment 1 Page 10 (6) N-TRAIN BATTERIES This change will allow a delay of the N-train battery capacity testing required by T/S 4.8.2.5.2.d. The time extension we are requesting to avoid a shutdown is from January 13, 1987 until the refueling outage scheduled to begin on May 23, 1987.
At the Donald C. Cook Nuclear Plant, the electrical loads associated with each unit's turbine-driven auxiliary feedwater pump (TDAFP) are supplied through the unit's N-train battery charger. The N-train battery itself is kept continuously charged by being connected in series with the charger. In the event power to the battery charger is lost, the TDAFP loads are supplied through the N-train battery until power to the charger is restored.
T/S 4.8.2.5.2.d requires verification every 18 months (during shutdown) that the battery capacity (with the charger disconnected) is adequate to supply and maintain its associated emergency loads. Typically, this test results in discharging the battery below the minimum operability levels called for by T/S 4.8.2.5.2. T/S 3.8.2.5 requires that the TDAFP be declared inoperable in Modes 1, 2, and 3 if the N-train battery is inoperable. Per T/S 3.7.1.2, therefore, we are allowed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to restore the battery (and thus the pump) to operable status before shutdown is required, provided neither of the other two motor-driven auxiliary feedwater pumps is inoperable. The battery service test required by T/S 4.8.2.5.2.d cannot be performed at power because the test and subsequent restoration of the battery to the T/S required charge level cannot be completed in the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allotted. Additionally, T/S 4.8.2.5.2.d specifically requires the testing to be performed during shutdown.
Justification for Chan e It is our belief that the temporary relief requested from the requirements of T/S 4.8.2.5.2.d will not pose a significant risk to public health and safety. In the approximately 8 years that the N-train batteries have been in service, they have regularly passed the testing required by T/S 4.8.2.5.2.d. Weekly checks of electrolyte level, specific gravity, and voltage are conducted as required by T/Ss. Visual inspections of the battery per T/S 4.8.2.5.2.c.l are conducted quarterly, even though T/Ss only require this every 18 months. The surveillances noted, as well as others required by T/S 4.8.2.5.2.b to be performed quarterly, would be expected to provide early indication of battery deterioration. Also, this battery passed its 5-year 80%
discharge test in 1984 which demonstrated adequate performance capability.
In addition it should be noted that the N-train battery has an amp-hour rating significantly greater than is needed to operate the N-train system.
\ 4r AEP:NRC:0967D Attachment 1 Page 11 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 As described above, the N-train battery has an excellent record for passing the surveillance test required by T/S 4.8.2.5.2.d. Thus, we have no reason to suspect that the batteries would fail their next required test, even if the surveillance interval was extended. Additionally, other tests are routinely conducted which would be expected to provide early indication of battery deterioration. Thus, we do not believe the change would significantly increase the probability or consequences of a previously evaluated accident, nor should it significantly reduce a safety margin.
Criterion 2 Because the change will not introduce any new operating configuration or, equipment from that presently allowed by T/Ss, the change should not create the possibility of a new or different kind of accident from any which has previously been evaluated.
Criterion 3 For the reasons given in Criterion 1 we believe this change does not involve a significant reduction in a margin of safety.
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.
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AEP: NRC: 0967D Attachment 1 Page 12 (7) DIVIDER BARRIER SEAL AND INSPECTION We are requesting an extension for testing and inspection of the divider barrier seal, as specified by T/S 4.6.5.9, from the required February 5, 1987 date until the refueling outage scheduled to begin on May 23,'987. The divider barrier seal serves to limit the ice condenser bypass leakage from the lower to upper volume in the event of an accident. The divider barrier seal is a purely passive component and is not subject to an active failure.
The reason for the extension request is that the seal is not accessible during power operation. Our review of recent surveillance test results has indicated that the seal is in excellent condition. The passive nature of the seal allows it to perform its safety function without dependence on any other safety system. Therefore, based on the passive nature of this system and the test history of the seal we do not believe the surveillance extension will significantly impact the ability of the seal to perform its safety 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 evaluated, (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 belo~.
Criterion 1 As stated above, the purpose of the divider barrier seal is to limit leakage bypassing the ice condenser between the lower and upper containment volumes. Based on previous operational experience of the two units, we believe that the extension we are requesting will not result in seal deterioration to the extent that the barrier cannot perform'ts intended
- function. Hence we have concluded that the extension we have requested will not involve a significant increase in the probability or consequences of previously evaluated accidents.
Criterion 2 Recent data has shown that the extension will not result in seal degradation, and for this reason we believe the surveillance extension will not result in a change in plant con'figuration or operation. We believe the extension we have requested will not create the possibility of a new or different kind of accident from any previously evaluated or analyzed.
AEP:NRC:0967D Attachment 1 Page 13 Criterion 3 Because of the passive nature of this system and its excellent test .
,history, we believe this surveillance 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.
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.
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AEP:NRC:0967D Attachment 1 Page 14 (8) PORV LIMIT SWITCHES The purpose of this change is to request an extension for the channel calibrations of the PORV limit switches required by'T/S Table 4.3-7, Item 12.
The extension is needed from February 8, 1987 until the Unit 1 refueling outage currently scheduled to begin on May 23, 1986. Calibration of the PORV limit switches cannot be performed at power because the limit switches are located inside the pressurizer doghouse, which is inaccessible during power operation. It is noted that the T/Ss allow the PORV acoustic monitor to be substituted for the PORV limit switches. This substitution will not relieve us of the need to calibrate the PORV limit switches because the acoustic monitor calibration will also be due prior to the refueling outage, and like the limit switch calibration cannot be performed at power because of inaccessibility.
Justification For Chan e An extension of the surveillance interval for calibrating the PORV limit switches should not pose significant risk to public health and safety. The PORV'imit switches use mechanical stops which indicate whether the PORVs are open or closed. The nature of these devices is such that they generally are not subject to drift in mechanical adjustment. Previous calibration tests performed on the switches have not indicated any problems with failure of the switches. Monthly channel checks are performed which indicate whether the limit switches are providing indication in the control room. If the PORVs were to open even slightly, indication would be provided to the operators by changing temperatures in the PORV discharge lines. Additional indication could also be provided by pressurizer relief tank (PRT) pressure and temperature instrumentation.
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.
AEP:NRC:0967D Attachment 1 Page 15 Criterion 1 Backup instrumentation (i.e., PORV discharge line temperature, PRT pressure, and PRT temperature) exists which provides indication of PORV status. Additionally, the PORV limit switches have not been subject to failure. Thus, we do not anticipate that the extension we have requested will involve a significant increase in the probability or consequences of 'an accident previously evaluated, nor will they involve a significant reduction in a margin of safety.
Criterion 2 This change introduces no new modes of plant operation or new equipment.
Thus, it is not anticipated that the change could introduce the possibility of any new or different kinds of accidents from any that have been previously analyzed or evaluated.
Criterion 3 See Criterion 1 above.
L'astly, 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
'n some increase to the probability of occurrence or consequences of a previously analyzed accident, but the results of which are clearl'y within the
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, limits established as acceptable. We believe these changes fall within the scope of this example. Therefore we believe this change does not involve a significant hazards consideration as defined in 10 CFR 50.92.
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AEP:NRC:0967D Attachment 1 Page 16 (9) CONTAINMENT SUMP INSPECTION The purpose of this change is to delay the visual inspection of the containment sump and its associated subsystem inlets, required by T/S 4.5.2.d.2. Because T/S 4.5.3.1 references T/S 4.5.2, relief from the requirements of T/S 4.5.3.1 with regard to the specific requirements of T/S 4.5.2.d.2 is also necessary.
The purpose of the surveillance requirements of T/S 4.5.2.d.2 is to provide assurance that the sump will be available as a source of RCS water during the recirculation phase of a LOCA. The visual inspection must verify that the sump is free of debris which may clog it, and that its components are free of corrosion and distress.
Justification for Chan e Records researched indicate that no evidence of structural distress or corrosion of the sump components has been detected. Thus, we have no reason to suspect that any will be detected by the next required inspection. As for.
debris, containment tours which include sump areas are generally conducted following reactor trips. T/S 4.5.2.c requires a visual inspection of containment prior to establishing containment integrity. Visual inspections of the sump areas were conducted following the May and July 1986 trips described in the main body of this submittal. Loose articles which have the potential to become sump debris are kept tied down inside containment.
Lastly, we note that plant administrative procedures require housekeeping inspections whenever the containment is entered during power operation.
For these reasons, we do not believe that a delay in the required visual inspection will significantly increase the risk of the sump being disabled by debris.
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 evaluated, (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 sump components have not indicated evidence of st): <<"tural distress or corrosion. Additionally, administrative controls exist whiiii are designed to ensure that the containment is kept free oF debris which coiilil clog the sump.
AEP:NRC:0967D Attachment 1 ,"
Page 17 Thus, we do not believe that the change would involve a significant increase in the probability or consequences of a previously evaluated accident, nor should it'esult in a significant reduction in a margin of safety.
Criterion 2 We believe that our past evidence on sump corrosion and distress, as well as our administrative controls on housekeeping, is sufficient to ensure that the time extension 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 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.
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AEP:NRC:0967D Attachment 1 Page 18 (10) REACTOR COOLANT PUMP SPRAY/SPRINKLER SYSTEM This change would allow a delay of the reactor coolant pump spray/sprinkler system testing and inspection required by T/S 4.7.9.2.b. The time extension required to avoid a shutdown would be from February 19, 1987 to the refueling outage scheduled to begin on May 23, 1987. The extension is being requested because the system is inaccessible during power operation.
Since the system was installed, it has not failed a surveillance test from the standpoint of being incapable of performing its intended safety function of extinguishing a fire.
We believe that this change would not result in a significant risk to the public health and safety, based on the inspection and test history of the system.
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 On the basis of the test history of the reactor coolant pump spray/sprinkler system, we believe that the extension will not result in a significant increase in the probability or consequences of a previously evaluated accident.
Criterion 2 The surveillance extension will not result in a change in plant configuration or operation. Therefore, this change will not create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated.
AEP:NRC:0967D Attachment 1 Page 19 Criterion 3 We believe that an extension of llance interval will not result surveillan o thee surve in a significant re d uct ion in the mar~in g of safet sa e y, for the reasons given in Criterion 1 above.
Lastly, we note that the Commission has p rovid g " g azards by providing certain examples (48 FR s gnz scant e s xth in some increase too thee pro
'f ility o amendments considered not likel to invo these examples robabil a
p refers to changes c which may result of occurrence or conse quences ue of a previously analyzed accident en , b ut th e results of which a re within the limits established as acceptable.e. Wee b elieve1 these chan g es fall within,the scope of this example. Thereforee we b e 1'ieve this chan g e does not involve a significant fi i hazards consideration a s d e ne d n 10 CFR 50.92.
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AEP:NRC:0967D Attachment 1 Page 20 (11) DIESEL-GENERATOR TESTING An extension of the surveillance interval is requested for the surveillance requirements of T/S 4.8.1.1.2.b. These surveillances are required by T/Ss to be performed during shutdown. The requirements include subjecting the diesel to an inspection in accordance with manufacturer's recommendations, as well as testing to verify that the diesel generator and its associated circuitry are capable of energizing, sequencing and shedding the emergency loads upon receipt of the appropriate signal. An extension of the surveillance interval is also necessary for part of the requirements of T/S 4.8.1.2, since T/S 4.8.1.1.2.b is referenced there. The extension needed for both T/Ss is from February 20, 1987 until the D. C. Cook Unit 1 refueling outage, which is currently scheduled to begin on Hay 23, 1987.
During this approximately three-month period, each diesel generator would accumulate six additional starts and six to eight additional running hours.
This is considered insignificant in light of the wear history of each machine.
A review of diesel generator repair records from 1980 to the present indicated no problem areas which in our judgement would be significantly affected by the proposed, surveillance interval extension. Also, a review of previous test results did not indicate any reason to suspect th'at the diesel generator associated circuitry (i.e., energizing, sequencing, and shedding the various emergency loads) would not pass required surveillance tests with the surveillance interval extended.
Two other extensions related to the diesel generators are also necessary to avoid a shutdown. These are for the requirements of T/Ss 4.4.11.3 and 4.7.4.1.b. T/S 4.4.11.3 requires testing of the emergency power supply for the power operated relief valves (PORVs) and their associated block valves.
Since this testing involves cycling the PORVs and block valves, it is generally performed during shutdown and in conjunction with the diesel generator testing requirements of T/S 4.8.1.1.2.b, as suggested by T/S 4.4.11.3. T/S 4.7.4.1.b involves testing automatic valves in the essential service water (ESW) system. Per T/Ss, this surveillance testing must be performed during shutdown. Since some of the ESW valves which are required to be tested involve cooling water for the diesel generator and its associated equipment, this testing is generally conducted in conjunction with the diesel generator testing of T/S 4.8.1.1.2.b.
The extension for both the ESW valves and the PORV emergency power supply are needed for the period February 20, 1987 to the Unit 1 refueling outage.
Previous test results do not indicate any reason to suspect that the valves and their associated circuitry would not pass the required surveillance with the extended interval.
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
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III AEP:NRC:0967D Attachment 1 Page 21 (3) involve a significant reduction in a margin of safety.
h Our evaluation of the proposed change with respect to these criteria is provided below.
Criterion 1 For the diesel-generator machinery, the extension will result only in approximately six additional starts and six to eight additional run hours.
This is considered insignificant with regard to the wear history of each machine. For the diesel-associated circuitry, the ESP automatic valves, and the PORV emergency power supply, our review of previous test data has not indicated any reason to believe the equipment would not pass the required surveillance tests with the extended interval. Thus, it is our belief that these changes will not result in a significant increase in the probability or consequences of a previously evaluated accident, nor will they result in a significant reduction in a margin of safety.
Criterion 2 No new modes of plant operation, nor any changes to plant equipment will occur as a result of this change. Thus, we believe that these changes will not create the possibility of a new or different kind of accident from those previously analyzed or evaluated.
Criterion 3 The discussion under Criterion 1, above, is also applicable to Criterion 3.
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 of occurrence or consequences of a previously analyzed accident, but the results of which are clearly within the limits established as acceptable. Ve believe these changes fall within the scope of this example. Therefore we believe this change does not involve a significant hazards consideration as defined in 10 CFR 50.92.
AEP:NRC:0967D Attachment 1 Page 22 (12) AUXILIARYFEEDWATER PUMP TESTING Extensions are necessary for the channel functional test of T/S Table 4.3-2 Item 6.d and for T/S 4.7.1.2.b from February 23, 1987 to the Unit 1 refueling outage'/S Table 4.3-2 Item 6.d requires a channel functional test to be performed on an 18 month basis. To perform this testing during power operations would involve tripping at least one main feed pump, which would result in a reduction of power and cause a thermal transient to be imposed on the plant.
T/S 4.7.1.2.b requires testing to demonstrate that the motor- and turbine-driven auxiliary feedwater pumps start and that the associated automatic valves actuate to their correct position upon receipt of the appropriate signal as listed in T/S Table 4.3-2. Per T/S 4.7.1.2.b, this testing must be performed during shutdown.
Although we are requesting an extension of the surveillance requirement for these tests, in practice the essential portions of these T/Ss (that is, startup of the auxiliary feedwater pumps when required and movement of the valves to their correct position) have occurred several times since the last refueling outage. Portions of this system are actuated when the unit trips.
The last such trip occurred on July 22, 1986. Prior testing experience with regard to these surveillances has indicated no significant problems when the surveillance was performed. Although we recognize that not all the actuation signals have been challenged as a result of spurious trips, we feel that our recent experience, in conjunction with the excellent test history in this area, justifies our request to extend the surveillance interval.
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 evaluated, (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 As stated above, portions of the system have undergone a challenge due to a recent actuation (that is, a unit trip). It is recognized that not all the actuation signals have been challenged as a result of the trip; however, we believe that the recent actuation, in conjunction with the previously stated
W AEP:NRC:0967D Attachment 1 I Page 23 excellent test history, shows that a surveillance interval extension does not involve a significant increase in the probability or consequences of an accident previously evaluated.
Criterion 2 The surveillance extension will not result in a change in plant configuration or operation. We believe that the auxiliary feedwater system if will serve its intended safety function required, and that the extension we are requesting is not significant with regard to impact on public health and safety. Therefore, this change will not create the possibility of a new or different kind of accident from any previously analyzed or evaluated.
Criterion 3 Based on the performance history noted above we believe this surveillance 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.
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 S0.92.
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AEP:NRC:0967D Attachment 1 Page 24 (13) TIME-RESPONSE TESTING We request an extension for the time-response testing required by T/S 4.3.1.1.3 and 4.3.2.1.3 for the instrumentation in Tables 3.3-1 and 3.3-3.
These tests cannot be performed at power without violating T/Ss. T/Ss 3.3.1.1, 3.3.2.1 and 3.0.3 require the plant to be shutdown reactor trip or engineered safety feature (ESF) instrumentation is not if sufficient operable. Both reactor trip and ESF trains (all channels) must be taken out of service during this test because the same test signal goes into all trains.
Upon receipt of a test signal, the trains which are not in test mode would initiate protective functions such as safety injection and containment spray.
Therefore, the time-response tests must be done during shutdown.
The time-response test for which we are requesting exemption has not had a failure since initial startup calibrations in 1974. The equipment for which time-response testing is required is also subject to a surveillance program of channel checks, channel functional tests, and channel calibrations. Except for those few (9 out of 81 sensors) calibrations for which we are requesting exemption, these surveillances will be performed, ensuring the operability of the instrumentation involved. Therefore we believe that this extension for time-response testing will not adversely impact the ability of this equipment to perform its safety 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 signi.ficant 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 As stated above, the time-response testing has not had a failure in over 11 years. The normal surveillance program will ensure the operability of the equipment during the extension period. Therefore we believe that this proposed extension will not involve a significant increase in the probability or consequences of an accident previously analyzed.
Criterion 2 The surveillance extension requested for the time-response testing will not change the design or operation of the plant. Therefore we believe that this change will not create the possibility of a new or different kind of accident from any previously analyzed or evaluated.
AEP:NRC:0967D Attachment 1 Page 25 Criterion 3 As stated above, the surveillance program will ensure that the equipment is operable and that it is capable of performing its safety function. The delay in performing the time-response testing will not affect the ability of this equipment to perform its safety function. Therefore we believe this change does not involve a significant reduction in a margin of safety.
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 of occurrence or consequences of a previously analyzed accident, but the results of which are 'clearly within the limits established as acceptable. We believe these changes fall within the scope of this example. Therefore we believe this change does not involve a significant hazards consideration as defined in 10 CFR 50.92.
Attachment 2 to AEP:NRC:0967D Proposed Revised Technical Specifications Pages for D. C. Cook Unit 1