ML17334B233
| ML17334B233 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 10/05/1988 |
| From: | Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | Murphy T NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-1.097 AEP:NRC:0773AB, AEP:NRC:773AB, NUDOCS 8810130179 | |
| Download: ML17334B233 (117) | |
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ACCEMENTED DIST BUYION DEMONSTRA.i.ON SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:8810130179 DOC.DATE: 88/10/05 NOTARIZED: NO
'DOCKET FACXL:50-315 Donald C.
Cook Nuclear Power Plant, Unit 1, Indiana 05000315
50-316 Donald C.
Cook Nuclear Power Plant, Unit 2, Indiana 6
05000316 AUTH.NAME AUTHOR AFFILIATION ALEXICH,M.P.
Indiana Michigan Power Co. (formerly Indiana 6 Michigan Ele.
RECIP.NAME RECIPIENT AFFILIATION MURPHY,T.E.
Document Control Branch (Document, Control Desk) l
SUBJECT:
Provides updated status rept re implementation of Reg Guide B6 R
1.97,Rev 3 for plant.
DISTRIBUTION CODE:
A003D COPIES RECEIVED:LTR ENCL SIZE:
TITLE: OR/Licensing Submittal:
Suppl 1 to NUREG-0737(Generic Ltr 82-33)
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Indiana Michigan Power Company P.O. Box 16631 Columbus, OH 43216
à AEP:NRC:0773AB Donald C.
Cook Nuclear Plant Units 1 and 2
Docket Nos.
50-315 and 50-316 License Nos.
DPR-58 and DPR-74 REGULATORY GUIDE 1.97, REVISION 3 U.S. Nuclear Regulatory Commission Attn:
Document Control Desk Washington, D.C.
20555 Attn:
T.
E. Murley October 5, 1988
Dear Dr. Murley:
The purpose of this letter is to provide you with an updated status report (attached) that details our compliance with Revision 3 of Regulatory Guide 1.97.
In our previous submittal (AEP:NRC:0773AC, dated June 20, 1988),
we had committed to providing you this updated status report by October 1,
1988.
This status report consolidates, clarifies and updates all of our previous submittals (AEP:NRC:07730, dated October 15, 1985 and AEP:NRC:0773S, dated June 29, 1987).
We again request that your final licensing decision concerning Regulatory Guide 1.97, Revision 3 be based on the attached updated status report.
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. Alexich Vice President MPA/eh Attachment cc:
D. H. Williams, Jr.
W.
G. Smith, Jr.
- Bridgman R.
C. Callen G. Charnoff G.
Bruchmann A. B. Davis
- Region III NRC Resident Inspector
- Bridgman (oo>
'h
Dr. T.
E. Murley AEP:NRC:0773AB bc:
S. J.
Brewer/R.
A. Kraszewski S.
H. Horowitz/T. 0. Argenta/R.
C. Carruth J. J. Markowsky/S.
H. Steinhart/P.
G. Schoepf J.
G. Feinstein P.
A. Barrett M. L. Horvath
- Bridgman J.
F. Kurgan J.
B. Shinnock J.
F. Stang, NRC - Washington, D.C.
AEP:NRC:0773AB DC-N-6015.1
ATTACHMENT TO AEP'NRC'0773AB STATUS REPORT IMPLEMENTATION PLAN OF REGULATORY GUIDE 1.97, REVISION 3, FOR THE DONALD C.
COOK NUCLEAR PLANT UNITS 1 AND 2 8810130179 I
ATTACHMENT TO AEP:NRC'0773AB Page 1
1.0 BACKGROUND
In August 1984, American Electric Power Service Corporation (AEPSC) contracted the engineering consulting firm of DiBenedetto, Farwell &
Hendricks to perform the detailed design study required to determine the status of the Cook Nuclear Plant Units 1 and 2 compliance with Regulatory Guide 1.97, Revision 3.
A preliminary Status Report (AEP:NRC:0773J dated February 28, 1985),
based on preliminary findings by our consultant, was submitted to the NRC staff.
A Final Status Report (AEP:NRC:07730 dated October 15, 1985) was then issued based on further work done by our consultant, with the understanding that further updating may be necessary.
We subsequently issued an update to our October 15, 1985, submittal (AEP:NRC:0773S dated June 29, 1987) that responded to numerous questions asked by the NRC in a Preliminary Technical Evaluation Report and also submitted various specific deviations from Regulatory Guide 1.97, Rev.
3 recommendations.
In order to consolidate, clarify, and update our previous submittals, AEP:NRC:0773J, AEP:NRC:07730, AEP:NRC:0773S, we are submitting this document for NRC review and final licensing action.
It should be noted that because design change activities have been ongoing since 1986 with respect to Regulatory Guide 1.97, some planned actions identified in AEP:NRC:07730 have since been completed.
These items are identified within Section 3.0 of this letter as "No Further Action Required" in the Remarks/Action Req'd column and listed as "CMPLT" in the Unit 1 or Unit 2 Schedule column.
2.0 STATUS REPORT Section 3.0 of this attachment contains information regarding instrument range, environmental qualification, seismic qualification, quality assurance, redundancy, power supply, location of display,
- remarks, and, as appropriate, an upgrade schedule for each type A,B,C,D and E variable listed in Regulatory Guide 1.97, Revision 3.
The format and content of Section 3.0 is
ATTACHMENT TO AEP'NRC:0773AB Page 2
consistent with the requirements of Section 6.2 of Supplement No.
1 to NUREG-0737.
Section 3.0 is also consistent in organization with Table 3
(PWR Variables) of Regulatory Guide 1.97, Rev.
3 dated May 1983.
The schedule for each instrument indicates, as applicable, when the recommendations of Regulatory Guide 1.97, as described in this attachment, will be met.
In those instances in which the design of our post-accident monitoring (PAM) systems deviate from the guidance provided in Regulatory Guide 1.97, Revision 3, these deviations are explicitly identified in the Section 3.0 tables, consistent with the requirements of Section 6.2 of Supplement 1 to NUREG-0737.
A discussion of each deviation is contained in Section 2.1 and 2.2.
A summary of deviations is provided in Section 2.3.
The Section 2.3 summary also cross-references the deviations to Section 3.0 table entries.
2.1 DEVIATIONS RELATED TO LICENSED DESIGN This section provides a discussion of those areas in which we have identified deviations from the Regulatory Guide 1.97, Revision 3
recommendations where the deviations are primarily a result of the originally licensed design of the Cook Nuclear Plant.
A deviation number is assigned to each deviation identified; these deviation numbers are used in the Section 2.3 summary and the Section 3.0 table entries.
2.1.1 Deviation No.
DV-1 Environmental uglification As provided by 10 CFR 50.49(k), originally installed Cook Nuclear Plant Qualified instrumentation located in potentially harsh environment has been qualified in accordance with "Guidelines for Evaluating Qualification of Class 1E Electrical Equipment in Operating Reactors,"
November 1979 (DOR Guidelines).
Qualified equipment ordered after February 22, 1983, is to have environmental qualification in accordance with Category I of NUREG-0588 (i.e.,
IEEE Std. 323-1974) unless there are sound reasons to the contrary.
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Equipment located in a mild environment is not required to be environmentally qualified.
The above is consistent with the current licensing basis of the Cook Nuclear Plant.
- However, because we do not literally comply with the Reg.
Guide 1.97 recommendations for environmental qualification, we are noting herein our deviation from the Reg.
Guide 1.97 guidance on this issue.
2.1.2 Deviation No. DV-2'eismic uglification Seismically qualified equipment meets the provisions of Cook Nuclear Plant Updated FSAR p. 7.2-12 dated July 1982 (i.e., protection equipment is designed such that, for a design basis earthquake, the equipment will not lose its capability to perform its design objective, to shut the plant down and/or maintain the unit in a safe shutdown condition).
Reactor protection Instrumentation originally installed at the Cook Nuclear Plant was seismically tested by Westinghouse Electric Corporation as documented in WCAP-7397-L, "Topical Report Seismic Testing of Electrical and Control Equipment" dated January 1970.
No industry standards regarding seismic qualification existed at that time.
Consistent with our current licensed design, equipment ordered for Regulatory Guide 1.97 upgrading will be seismically qualified in accordance with IEEE Std. 344-1975 unless there are sound reasons to the contrary.
Seismic qualification for existing Category 1 circuits is provided from the sensor up to and including the channel isolation device (shown as a
"Signal Isolator" in Figure 2.1-1), typically installed in the reactor protection cabinets located in the control room area.
Our design does not provide for seismic qualification of equipment beyond the channel isolation device for existing Category 1
circuits; however, we have installed seismically qualified indicators and/or recorders on variables monitored by Category 1
instrumentation.
This was done because we believe that the primary indicators/recorders that provide direct Category 1 variable indication are the only credibly vulnerable equipment installed beyond the isolators that could provide ambiguous or misleading information due to a failure during a seismic event.
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ATTACHMENT TO AEP:NRC:0773AB Page 4
It should be noted that other equipment beyond the channel isolation device (such as control equipment) that directly provides information for the channel indication is of a similar design to existing equipment that is seismically qualified (such as reactor protection instrumentation).
As such, in spite of the fact that we do not take credit for seismic qualification, we believe that this equipment would not provide ambiguous or misleading information to the operator following a seismic event.
Our design does not provide seismic qualification of cables and equipment within the non-safety related portion of the PAM circuits.
Nevertheless non-safety related portions of the PAM channels that directly provide PAM indication are located in the control room, or control room vault, which are seismically qualified structures.
Cables and equipment associated with PAM circuits located outside cable vault areas (e.g.,
cables to valves, cables to Technical Support Center (TSC) I/O cabinets) may not be maintained entirely in a seismically qualified structure.
Since our current licensing basis, as described
- above, does not literally comply with the Reg.
Guide 1.97 recommendations for seismic qualification beyond the isolation device for existing
- circuits, we are noting herein our deviation on this issue.
Figure 2.1-1 provides an example of an existing PAM channel that is typical of our originally licensed plant design.
Consistent with the Cook Nuclear Plant licensing basis, PAM instrument channels or portions thereof scheduled to be upgraded or added per Reg.
Guide 1.97 and requiring seismic qualification will be installed to meet Reg.
Guide 1.97 Category 1 recommendations except as noted above and in the tables in Section III.
2.1.3 Deviation No. DV-3'alit Assurance The provisions of the Cook Nuclear Plant QA program as described in Updated FSAR Section 1.7 have been applied to the safety-related
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ATTACHMENT TO AEP:NRC:0773AB Page 5
portions of the PAM circuitry.
This program satisfies the requirements of 10 CFR 50, Appendix B.
The implementation of specific Regulatory Guides and ANSI Standards regarding quality assurance is consistent with the commitments of Cook Nuclear Plant FSAR Section 1.7, Appendix A, dated July 1988, which address all but one (Regulatory Guide 1.28) of the guidance documents referenced by Reg.
Guide 1.97, Rev.
3.
The Cook Nuclear Plant QA program has not necessarily been applied to non-safety related portions of the PAM system (see Fig. 2.1-1).
Since we do not literally comply with the Reg.
Guide 1.97, Rev.
3 quality assurance recommendations, we have noted herein our deviations from the Reg.
Guide guidance in this area.
2.1.4 Deviation No. DV-4'edundanc Instruments installed at the Cook Nuclear Plant to meet redundancy requirements have a minimum of two (2) redundant, electrically independent and physically separate channels up to and including any isolation device (shown as "Signal Isolator" in Figure 2.1-1) typically installed in the reactor protection cabinets located in the control room area.
For existing circuits, the isolation device is providing isolation between safety-related circuits and
,non-safety related circuits, which includes portions of the PAM circuitry as per our licensed design.
The display may be a common multi-pen recorder or a dual indicator.
Separation of safety-related circuits up to and including the isolation device is in accordance with Updated FSAR p. 7.2-4 dated July 1982 Figure 2.1-1 is an electrical schematic that shows a typical cable and hardware configuration for redundant PAM channels.
This illustration presents a typical configuration and is not meant to represent an actual circuit.
Exact circuit configuration may vary from the illustration.
This configuration reflects our originally licensed design.
This design was completed prior to the issuance of the NRC guidance regarding physical independence of electrical
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ATTACHMENT TO AEP:NRC:0773AB Page 6
systems (Regulatory Guide 1.75, September 1978) and the associated IEEE standard for separation of Class 1E equipment and circuits (IEEE Std. 384-1974).
Physical separation (of both electrical cabling and hardware) is maintained between the redundant reactor protection set channels.
The BOP interconnecting cabling between the reactor protection cabinets and the various PAM readout devices is routed through the cable vault area and then back to the control room.
The redundant PAM signal and power cables in the cable vault area and in the control room are not physically separated.
Also, the redundant PAM readout devices are not physically separated.
As previously stated, there are also some cases of redundant PAM signal cabling feeding a single readout device (recorder or computer).
At the request of NRC staff, an evaluation of the impact of this lack of physical separation in these areas was performed.
This evaluation identified two events that have the potential for compromising the integrity of the PAM system as presently installed.
These events are a fire in the cable vault area or a severe natural phenomenon.
In the case of a severe natural phenomenon, the only event of significance with regard to the cable would be an earthquake.
Even in the unlikely event, of a design-basis earthquake, the PAM system will continue to serve its intended function, after the appropriate system modifications have been completed.
In the event of a fire in the cable vault area, we would not expect to need the PAM instruments to follow the course of another accident, and the ability to bring the plant to a safety configuration would not be compromised.
This position is consistent with actions we have taken to ensure compliance with Appendix R to 10 CFR 50, "Fire Protection Program for Nuclear Power Facilities Operating prior to January 1, 1979."
Figure 2.1-1 also shows cables leaving the control room and vault areas and going out into other plant areas.
These areas in most cases will not offer the same missile and fire protection as offered
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ATTACHMENT TO AEP'NRC:0773AB Page 7
by the control room and vault areas.
The signals going to these remote areas are also not isolated from the PAM signals in the control room or cable vault areas.
The cables in mbst cases have been installed in tray and conduit which has been mounted to the same requirements as safety-related tray and conduit.
The cable is the same as that used for safety-related circuits.
The redundant PAM signal cables are not physically separated.
In some cases, cables carrying redundant signals also terminate in a cabinet or device that has not been designed to the same separation requirements as safety-related devices.
These cabinets or devices may not be seismically qualified.
Consistent with the current licensing basis of the Cook Nuclear Plant PAM instrument channels or portions thereof scheduled to be upgraded or added per Reg.
Guide 1.97 and requiring redundancy will be installed to meet Category 1 recommendations except as noted above and in the tables in Section III.
Since our current license design as described above, does not conform to the Reg.
Guide 1.97, Rev.
3 recommendations for physical separation beyond the isolation devices installed in PAM circuits, we are noting herein our deviations from the Reg.
Guide guidance on this issue.
2.1.5 Deviation No.
DV-5 Dis la Location The display will be indicated either inside the control room or at another location as permitted by Regulatory Guide 1.97.
In some instances, the analog recording of Category 1 variables is not directly provided, however the Technical Support Center computer does record these variables when analog recording is not provided.
Because we do not literally comply with the recording recommendations of Reg.
Guide 1.97 Rev.
3 for Category 1 variables, we are noting herein our deviation from the Reg.
Guide guidance on this issue.
Please note that this applies to analog variables only, Discrete variables such as breaker or valve positions are not recorded.
ATTACHMENT TO AEP'NRC:0773AB Page 8
2.2 DEVIATIONS IDENTIFIED FOR SPECIFIC VARIABLES 2.2.1 Deviation No. DV-6'ontainment Pressure Monitoring of containment pressure is currently provided by two Category 3 wide-range
(-5 to 36 psig) instruments and four Category 3 narrow-range
(-5 to 12 psig) instruments.
The design pressure of the Cook Nuclear Plant containments is 12 psig.
The four narrow-range instruments are scheduled to be upgraded to meet Reg.
Guide 1.97 Category 1 recommendations by the end of the 1987 refueling outage for Unit 1 (which has been completed) and by the end of the 1988 refueling outage for Unit 2.
The vide-range containment pressure instrumentation ranges were revised to meet the requirements of NUREG-0578 and NUREG-0737.
- However, these instruments are not powered by an emergency standby power source as recommended by Regulatory Guide 1.97, Rev.
3 for Category 1 instrumentation, and they do not meet the Category 1
separation criteria.
The wide-range instrumentation is, however, highly reliable, and as a result we believe it is unlikely that it would not be available if needed to monitor the course of an accident.
Further, for other than short-term individual compartment pressure
- peaks, the narrow-range instrumentation would span the range of pressure anticipated in our evaluation of loss-of-coolant type accidents.
The above justification is the basis for our deviation from the Regulatory Guide 1.97, Rev.
3 recommendation to provide Category 1 wide-range containment pressure instrumentation.
2.2.2 Deviation No. DV-7'ubcoolin Meter The saturation meter equipment was originally installed in accordance with the requirements of NUREG-0578.
In an SER dated March 20,
- 1980, the equipment installed to monitor degrees of subcooling was found to be acceptable (NRC letter, A. Schwencer to John E. Dolan, dated March 20, 1980).
As noted in that correspondence, the device installed was-a discrete digital monitor,
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ATTACHMENT TO AEP:NRC:0773AB Page 9
and the plant process computer was used in conjunction with this monitor to provide subcooling margin.
Additionally, as part of the NUREG-0737 Supplement 1 requirements, a subcooling margin is provided by our Technical Support Center computer.
We believe that these three instrument systems, which have been installed to be consistent with the requirements of their appropriate documents, are reliable.
As a result, we believe it is unlikely that they would not be available if needed to monitor the course of an accident.
Further, neither NUREG-0578, nor subsequently, NUREG-0737 required seismic qualification or redundancy for this instrumentation.
The above information is the basis for our deviation from the Regulatory Guide 1.97, Rev.
3 recommendation to provide Category 1
instrumentation for monitoring degrees of subcooling.
(This deviation is also discussed in AEP:NRC:0773S, Attachment 1, Item 3.3.3.)
2.2.3 Deviation No. DV-8'eactor Coolant S stem RCS Sam lin Boron Concentration Ran e
Primary coolant boron concentration can be measured in the range of 375 ppm to 10,000 ppm.
This range is based on PASS reactor coolant samples with a 1:1000 dilution.
The PASS would be used during and following loss-of-coolant accidents.
In the event of a LOCA, emergency boration and injection from the refueling water storage tank would occur and we would therefore expect a reactor coolant boron concentration substantially in excess of the low range of our PASS sample measurement capability.
The above is the basis for our deviation from the Regulatory Gui.de 1.97, Rev.
3 recommendation to provide the capability to measure boron concentration in PASS samples to the lower limit of 0 ppm.
(This deviation was also discussed in AEP:NRC:0773S, Attachment 1,
Item 3.3.2.)
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ATTACHMENT TO AEP:NRC:0773AB Page 10 2.2.4 Deviation No. DV-9'CS Sam lin adioactivit As stated in our original submittal (AEP:NRC:07730, dated October 15, 1985),
the primary coolant system radioactivity is not continuously monitored by in-line instrumentation.
Rather, periodic analysis of reactor coolant grab samples is provided to detect deterioration of fuel cladding.
Our post-accident sampling system provides a diluted grab sample that is analyzed by the gamma spectrum analyzer in a range of 1 uCi/ml to 10 Ci/ml.
This measurement range is consistent with the Reg.
Guide 1.97, Rev.
3 recommendations for this parameter.
The above information is the basis for our deviation from the Regulatory Guide 1.97, Rev.
3 recommendation for continuous monitoring of radioactivity in the reactor coolant system.
(This deviation was also discussed in AEP:NRC:0773S, Attachment 1, Item 3.3.5.) It should also be noted that Category 1 requirements for this system are only to be applicable to equipment that operates equipment installed in the portion of piping that is Seismic Class I.
Electrical equipment operating equipment installed in Seismic Class 3 piping is to meet Category 3 requirements.
2.2.5 Deviation No. DV-10'ench Tank Level We do not rely on the quench tank to perform any pose-pressurizer release function.
- However, we are providing the following information in response to the evaluation of our October 15,
- 1985, submittal (AEP:NRC:07730) performed by EG6G on behalf of the NRC.
The range of 74% of total tank volume originally submitted was not accurately stated to show the adequacy of the existing installation.
The correct range should have been stated as being from 7 inches above the tank bottom to 7 inches below the tank top.
This range includes coverage of the sparger.
With regard to the ability to quench a "design-basis" pressurizer
- release, as noted above we do
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ATTACHMENT TO AEP:NRC'0773AB Page 11 not rely on the quench tank to perform this function.
The quench tank is used during normal plant operation to contain pressurizer releases from routine pressurizer pressure adjustments and valve leakage.
In the case of a design-basis event that causes the PORVs and safety relief valves to lift, two rupture discs will burst before reaching the quench tank design pressure of 100 psi.
Subsequently, discharge through the quench tank into the containment sump will occur.
With regard to over pressurization, we do not understand the basis for the EG&G position that sufficient gas volume exists to accept pressurizer release without becoming over pressurized.
As noted
- above, over pressurization will not occur because rupture discs will burst and discharge into the containment before reaching the tank design pressure of 100 psig.
Normal water level is kept at between 80% and 84% of the instrument range with a high alarm at 84% and a low alarm at 79%.
As such, in-leakage from the relief discharge system can be adequately monitored.
The above information is the basis for our deviation from the Regulatory Guide 1.97 recommendation to monitor quench tank level from top to bottom of the tank.
(This deviation was also discussed in AEP:NRC:0773S, Attachment 1, Item 3.3.15.)
2.2.6 Deviation No. DV-ll' G Wide-Ran e Level On August 21,
- 1981, we submitted a letter (AEP:NRC:0300G) that documented discussions with NRR staff clarifying certain portions of an NRC SER (June 16, 1981) of the Cook Nuclear Plant auxiliary feedwater system.
In that letter it was confirmed that Regulatory Guide 1.97 recommendations for steam generator level instrumentation did not have to be implemented at that time, but that implementation would be addressed at some time in the future through the Regulatory Guide 1.97 compliance/commitment process.
The steam generator wide-range level indication is not required for post-accident monitoring and in fact has been deleted from our
ATTACHMENT TO AEP:NRC'0773AB Page 12 Technical Specifications for Units 1 and 2.
This was stated in our December 10, 1980, letter, which submitted a proposed amendment to our Technical Specifications (AEP:NRC:0449).
As stated in that
- letter, the reasons for deletion of steam generator wide-range level indication from the Technical Specifications are.'1) the S/G wide-range level indication does not perform any safety-related function and is not assumed operable in the various plant safety analyses; and (2) the S/G narrow-range instrumentation, which we believe fulfillspost-accident monitoring requirements, is environmentally and seismically qualified, powered from a Class lE source and has three redundant channels per S/G.
The S/G level indication is backed up by auxiliary feedwater flow instrumentation.
The S/G wide-range level instrumentation is powered from a Class 1E Bus power source, but all four channels are powered by the same source.
Since this is not in compliance with the Regulatory Guide, Rev.
3 recommendations, and based on the information given above, we are noting our deviation from the Regulatory Guide 1.97, Rev.
3 recommendation for S/G level instrumentation.
(This deviation was also discussed in AEP:NRC:0773S,, Item 3.3.16.)
2.2.7 Deviation No. DV-12'ondensate Stora e Tank CST Level CST level indication is currently provided in the control room through three Category 3 level-measuring devices.
One of these instruments is electrically operated, while the other two are pneumatic devices.
In addition, CST level can be read at the local turbine-driven auxiliary feedwater pump control panel.
We have also committed to provide additional CST level indication by adding one instrument channel to meet Category 1 requirements.
The new channel was installed during the 1987 refueling outage on Unit 1 and will be installed before the end of the 1988 refueling outage on Unit 2.
The CST is the initial source of water for the auxiliary feedwater (AFW) system, and provides sufficient volume to maintain the reactor
ATTACHMENT TO AEP:NRC:0773AB Page 13 coolant system in a hot standby condition for nine hours.
In the event that sufficient water is not available from the CST in one unit, operating procedures call for a cross-tie valve to be opened to supply feedwater from the CST in the other unit.
In the unlikely event that neither CST can supply sufficient AFW, procedures require transferring the supply source to the essential service water system (ESWS).
The water supply for the ESWS is Lake Michigan.
The number and diversity of instrumentation available to provide CST level monitoring, and the ultimate availability of Lake Michigan as a source of auxiliary feedwater, is the basis for our deviation from the Regulatory Guide 1.97 recommendation to provide more than one Category 1 level indication for the CST.
(This deviation was also discussed in AEP:NRC:0773S,, Item 1.)
2.2.8 Deviation No.
DV-13 Containment S ra Flow When operating normally, each containment spray pump will deliver 3200 gpm (design flow) at 490 ft TDH.
Figure 2.2-1 shows containment spray pump flow as a function of pump discharge pressure.
The Figure 2.2-1 curve indicates the expected range of operation for the containment spray pumps.
This operating range stems from consideration of pump suction head, containment pressure, and pump operating characteristics.
Routine surveillance of spray pump operation is performed to ensure that, if containment spray is
- required, the pumps will operate in the indicated area of the flow curve and hence provide the necessary flow to the containment spray system. 'he reactor operators can, therefore, verify proper containment spray flow by monitoring spray pump discharge pressure to confirm that it is within the expected range.
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ATTACHMENT TO AEP:NRC:0773AB Page 14 It should be noted that the upper containment spray flow instrumentation (IFI-330 and IFI-331) cited in our original submittal (AEP:NRC:07730, dated October 15, 1985) me'asures only the flow provided by the RHR pumps to the upper containment spray, not the flow from the containment spray pumps.
- However, the containment spray pumps, not the RHR pumps, are normally used to supply containment spray flow.
Also, please note that the flow range of 0-200 gpm for IFI-330 and -331 (for measurement of RHR pump flow to the upper containment spray) given in that submittal is incorrect.
The correct range is 0-2500 gpm.
The above is the basis for our deviation from the Regulatory Guide 1.97, Rev.
3 recommendation for containment spray flow instrumentation.
(This deviation was also discussed in AEP:NRC:0773S, Attachment 1, Item 3.3.17.)
2.2.9 Deviation No.
DV-14 Volume Control Tank VCT Level Because of the following actions that apply for normal, accident, and post-accident conditions, we believe VCT level indication beyond that currently provided is not required.
Upon receiving a hi-level alarm, flow into the VCT is automatically fully diverted into the hold-up tanks.
If a low-level alarm is reached, an alarm alerts the operator to restore level.
In the event this effort fails, an emergency lo-lo level alarm is sounded and the refueling water sequence is automatically initiated.
We believe that this range (0-70 inches) is adequate to safely monitor the operation of this tank.
In the unlikely event that VCT level indication is lost and the VCT becomes completely full, a safety relief valve (set at 75 psig) will open and the excess water will be discharged into the hold-up tanks.
The above is the basis for our deviation from the Regulatory Guide 1.97 recommendations to monitor volume control tank level from top to bottom of'he tank.
(This deviation was also discussed in AEP:NRC:0773S, Attachment 1, Item 3.3.19.)
ATTACHMENT TO AEP:NRC:0773AB Page 15 2.2.10 Deviation No. DV-15 Noble Gases and Vent Flow from Condenser Air Removal S stem Exhaust This instrumentation was recently (1985) upgraded by the addition of a high-range noble gas detector.
Based on our recent primary calibration analysis, the range of this instrumentation was determined to be 5.8 x 10 uCi/cc to 1.86 x 10 uCi/cc Xenon-133
-7 4
dose equivalent.
On July 23,
- 1986, a letter was sent to the NRC (AEP:NRC:0678Y) in which we stated that post-accident conditions would not result in steam jet air ejector exhaust noble gas concentration greater than 2 x 10 uCi/cc.
On this basis we 3
requested an exemption from the NUREG-0737,Section II.F.1-1 upper-range requirement 10 uCi/cc in favor of a more realistic 5
4.
upper range of 10 uCi/cc.
The above also is the basis for our deviation from the Regulatory Guide 1.97, Rev.
3 recommended range for this parameter.
It should also be noted that tag numbers SFR-1900 and SFR-2900, given in our original submittal (AEP:NRC:07730, dated October 15, 1985) are incorrect.
The correct tag numbers are SRA-1900 and SRA-2900.
(This information was also provided in AEP:NRC;0773S, Attachment 1, Item 3.3.21.)
2.2.11 Deviation No.
DV-16 Noble Gases from S
G Safet Relief Valves The range of 3 uCi/cc to 20 x 10 uCi/cc as provided in our original 5
submittal (AEP:NRC:07730) was based solely on the monitor's response to Xe-133 and not to the anticipated mixture of radioisotopes following a steam generator tube rupture.
The lower limit of 0.1 uCi/cc of XE-133 equivalent mixture can be measured.
As stated in our September 8,
1986, letter (AEP:NRC:0678Z),
when the anticipated mixture of radioisotopes for a steam generator tube rupture is used, the maximum concentration released in the main steam effluent from the S/G PORV is calculated to be 0.263 uCi/cc
" Xe-133 equivalent activity.
With respect to this upper range limit, an exemption from the NUREG-0737 requirement of 1000 uCi/cc was requested in the September 8,
1986, letter and a 100 uCi/cc value
~t
ATTACHMENT TO AEP:NRC:0773AB Page 16 proposed.
No response to our request has been received at this writing.
The above is a basis for our deviation from the Regulatory Guide 1.97, Rev.
3 recommended upper limit of measurement for this variable.
(This information was also provided in AEP:NRC:0773S,, Item 3.3.22.)
2.2 '2 Deviation No.
DV-17 RCS Sam lin
- Chloride Content Chloride content in undiluted samples 30 days after an accident is measured in a range of 0.01 to 20 ppm.
For diluted samples (1:1000 dilution) taken within 4 days of an accident, the range of measurement is 10 to 20,000 ppm.
The above is the basis of our deviation from the Regulatory Guide 1.97, Rev.
3 lower limit of 0 ppm for this variable.
(This information was also provided in AEP:NRC:0773S,, Item 3.3.24, No. 4.)
2.2.13 Deviation No. DV-18'ontainment Air -
H dro en Content An exemption from the requirement for taking hydrogen grab samples of containment air was granted via a letter from Youngblood (NRC) to Dolan (AEP) dated November 5, 1986.
This exemption is the basis for our deviation with respect to the Regulatory Guide 1.97, Rev.
3 recommendations for this parameter.
Ve do, however, perform continuous monitoring of containment air hydrogen content in the range of 0 to 30 volume percent (see Item C-10).
2.2.14 Deviation No.
DV-19 Containment Air - Ox en Content NUREG-0737 does not require sampling of containment air oxygen content.
As noted above,
- however, we do continuously monitor hydrogen content, which makes containment air oxygen content of less concern from the standpoint of potential hydrogen flammability or deflagration.
The above is the basis for our deviation from the Regulatory Guide 1.97, Rev.
3 recommendation to sample for containment air oxygen content.
(This deviation was also discussed in AEP:NRC:0773S, Item 3.3.24, No. 7.)
ATTACHMENT TO AEP'NRC:0773AB Page 17 2.2.15 Deviation No. DV-20 Indicatin Lam s Circuits that require the use of indicating lamps for position indications, status indication, etc., will be using existing General Electric ET16 indicating lamps for this function.
We have been advised by the manufacturer that these indicating lamps meet their (the manufacturer's) interpretation of IEEE-344-1975.
This indicating lamp is a seismically rugged commercial grade device for which comprehensive qualification is not available.
Since these lamps are purchased as standard commercial grade material and are not manufactured for a specific order, 10 CFR 21 can not be applied to these devices.
The above is the basis for our deviation from Regulatory Guide 1.97, Rev.
3 for this device.
2.2.16 Deviation No.
DV-21 Centrifu al Char in Pum CCP Flow Indication Our original submittal (AEP:NRC:07730) identified both the CCP flow and CCP motor breaker status as Type A variables.
This would require Category 1 instrumentation for monitoring these parameters.
The CCP breaker status indi.cation will meet the Regulatory Guide 1.97, Rev.
3 recommendati,ons for Category 1 instrumentation except as noted in the summary table, Entry A.28.
With regard to the CCP flow indication, it should be noted that our Emergency Operating Procedures require manual operator action based on indication of pump operation or flow.
The CCP breaker status indication and other parameters serve to verify pump operation.
The non Category 1
CCP flow indication can serve as a backup.
The above is the basis for our deviation from the Regulatory Guide 1.97, Rev.
3 recommendation to provide Category 1 instrumentation for CCP flow indication.
(This deviation was also discussed in AEP:NRC:0773S,, Item 2).
ATTACHMENT TO AEP:NRC'0773AB Page 18 2.2.17 Deviation No. DU-22 Safet In ection SI Pum Flow Indication Reasons similar to those stated in Section 2.2.16 above are the basis for our deviation from the Regulatory Guide 1.97, Rev.
3 recommendation to provide Category 1 instrumentation for SI pump flow indication.
The SI pump motor breaker status instrumentation will meet Category 1 requirements except as noted in the summary
- table, Entry A.29.
(This deviation was also discussed in.
AEP:NRC:0773S, Item 3.)
2.2.18 Deviation No. DU-23'adiation Ex osure Rate ERA-7303 through 7308, and ERA-8303 through 8308 have ranges of 0.01 to 1000 R/HR.
Monitors ERS-7401,
- 7403, 7404,
- 8401, and ERA-7507,
- 7601, 7603, and 7605 have ranges of.0001 to 10 R/HR.
Monitors ERA-8403,
- 7504, 7508,
- 7602, and 7604 have ranges of 0.001 to 10 R/HR.
These ranges are different from t'e Regulatory Guide 1.97,
-1
+4 Rev.
3 recommended range of 10 to 10 R/HR.
With the exception of ERA-7305,
- 7306, 8305, and 8306, the worst case maximum estimated accident dose rate in the areas where these monitors are to be installed is less than the upper range limits noted above.
The stated upper range limits are used to provide more accurate, useful information and to help prevent false "low fail" alarms.
In the case of the ERA-7305,
- 7306, 8305, and 8306, the worst case maximum estimated accident dose rate is 1730 R/HR, which exceeds the detector's upper range limit of 1000 R/HR.
However, within one (1) hour, this dose rate drops to 573 R/HR, which is well within the upper range limit.
Personnel entry into an area where exposure may exceed 1000 R/HR (indicated by a "high fail" status indication) is highly unlikely and the dose rate in these areas will quickly fall below the upper range limit of 1000 R/HR (at which time a quantitative indication will again be available).
Therefore, the range of 0.01 to 1000 R/HR for these detectors is adequate for the areas in which they are installed.
ATTACHMENT TO AEP'NRC:0773AB Page 19 The above discussion is the basis for our deviation from the Regulatory Guide 1.97, Rev.
3 recommended range for area radiation monitors installed in areas potentially requiring personnel access for servicing of equipment important to safety.
2.2.19 Deviation No.
DV-24 Valve Position Indication -
CM-250 We will upgrade the valve position limit switches on valves VCR-11 and VCR-21 to meet the environmental qualification requirements of 10 CFR 50.49 and Regulatory Guide 1.97, Rev.
3 recommendations.
The valve position limit switches for valve QCM-250, as well as the associated cable and terminations, are qualified in accordance with 10 CFR 50.49(k) except that they have not been qualified'for submergence.
The QCM-250 position indication limit switch is located below maximum flood level.
Although this is not completely consistent with the Regulatory Guide 1.97, Rev.
3 recommendations for equipment qualification, we do not believe any upgrading of the position indication limit switch is necessary.
This is due to the fact that QCM-250 is designed to close within 15 seconds of a containment isolation signal, which means that the valve will not become submerged before it performs its safety function.
In
- addition, once the valve is closed, it is extremely unlikely that it would change position due to its submergence.
Given these considerations, we believe that QCM-250 in its present
- status, without upgrading, adequately meets the intent of Regulatory Guide 1.97, Rev.
3 recommendations for achieving verifiable containment isolation and is the basis for our deviation from the Regulatory Guide 1.97, Rev.
3 recommendations for this parameter.
The planned schedule for upgrading VCR-11 and VCR-21 to meet 10 CFR 50.49 requirements calls for this work to be completed in both units by the end of the refueling outages presently scheduled for 1989 (Unit 1) and 1990 (Unit 2).
ATTACHMENT TO AEP:NRC:0773AB Page 20 2.2.20 Deviation No. DV-25 CCW Water Tem erature Indication CTR -410
-415
-420 and -425 If CCW water temperature is not available, adequate CCW cooling can be verified by monitoring CCW flow and RHR inlet and outlet temperatures, all of which are qualified (or planned to be qualified) for the intended purpose.
Therefore, because of the availability of suitable diverse indications, environmental qualification of instrumentation monitoring this variable is not required.
We are submitting a deviation with respect to Reg.
Guide 1.97, Rev.
3 recommendations for this variable for environmental qualification.
(Deviation No.
DV - 25) 2.3
SUMMARY
OF REGULATORY GUIDE 1.97 REV.
3 DEVIATIONS Table 2.3-1 provides a cross-referenced summary of the deviations from the Regulatory Guide 1.97, Rev.
3 recommendations that were identified and discussed in Sections F 1 and 2.2.
Table 2.3-1
.shows, by deviation number, a brief description of each deviation, where the affected variables can be found in the Section 3.0 summary tables and where each deviation is discussed in Section 2.0.
2.4 ADDITIONALCLARIFICATION OF SECTION 3.0
SUMMARY
TABLE INFORMATION The following information is provided in order to clarify certain entries contained in the Section 3.0 summary tables.
2.4.1 Sensor Location The "sensor(s) location(s)" information requested in Section 6.2(e) of Supplement 1 to NUREG-0737 is assumed to mean the parameter(s) monitored by the sensor and not the sensor's physical plant location.
This information is provided in the column labeled "variable."
ATTACHMENT TO AEP:NRC:0773AB Page 21 2.4.2 Power Source Instruments reported as conforming to the Regulatory Guide 1.97, Rev.
3 power source recommendations have their power derived from a 120V AC or 250V DC safety-related power source.
Some existing circuits use non-safety related cable from the source to the instrument power supply.
2.4.3 Schedule The implementation provided for each variable is the current, best estimate of the completion of the final configuration for the associated instrument including redundancy, final displays, power
- supplies, documentation of qualification and turnover to plant operations.
The schedules are based on anticipated delivery and plant outage schedules.
Equipment delivery delays, environmental qualification test difficulties, or other problems,
- however, may cause delays in these schedules beyond our reasonable controls Considering these factors, our overall target date for completion of work undertaken in response to Regulatory Guide 1.97, Rev.
3 is by the end of the Unit 2 refueling outage presently scheduled for 1990.
This completion date is also contingent on any changes resulting from the NRC evaluation of our deviations.
2.4.4 Chan es from October 1985 Submittal AEP:NRC:07730 Because of additional information obtained after our previous submittal in October 1985, discussions with the NRC Staff and its consultant (EG&G), and information developed as a result of our ongoing engineering/design work, there are numerous differences between our October 1985 submi.ttal and this updated submittal.
All item numbers in the tables in Section 3.0 have been retained, regardless of whether they are currently in use, in order to provide easy cross referencing between this and previous reports.
Most changes involve the updating of status of modifications, incorporation of deviation requests and answers to questions
ATTACHMENT TO AEP'NRC:0773AB Page 22 provided in our AEP:NRC:0773S documents, clarification of several
- items, and miscellaneous correction and revisions.
It is noted that although many variables presently listed are-measured with fully functional instrumentation, they are not considered operational in the sense that they meet Regulatory Guide 1.97, Rev.
3 recommendations.
In addition, it should be noted that instruments previously reported to the staff as meeting the requirements of NUREG-0737 or IE Bulletin 79-01B do not, in all
- cases, meet all of the requirements of Regulatory Guide 1.97, Rev.
3 as noted within this document.
3.0
SUMMARY
TABLES These tables represent the status of the Donald C.
Cook Nuclear Plant when we will have completed all of the recommendations associated with Regulatory Guide 1.97, Rev.
- 3. It does not, nor is it intended to, reflect the status of the plant as of the date of this letter.
The schedules provided in this enclosure for meeting Regulatory Guide 1.97, Rev.
3 recommendations are not intended to change any previous commitment regarding NUREG-0737 or 10 CFR 50.49, whose requirements may be different.
"A:"
A letter "A" in the columns labeled EQ, SQ, QA, SF or PS indicates that we have applied the guidance provided in Regulatory Guide 1.97, Rev.
3 subject to specific deviations noted in the Remarks/Action Req'd.
column.
"CMPLT" indicates that the action identified in our previous submittal AEP:NRC:07730 has been completed on Unit 1 or will be completed on Unit 2 prior to returning to service from the current steam generator repair and refueling outage.
"NA" letters mean Not Applicable; i.e.,
does not apply.
ACHMENT TO AEP:NRC:0773AB Page 23 TABLE 2.3-1 Summar of Deviation Re uests from Re ulator Guide 1.97 Rev.
3 Deviation No.
Where Identified In Section 3.0 Where Ex lained DV-1 DV-2 DV-3 DV-4 DU-5 DV-6 DV-7 DV-8 DV-9 DV-10 DV-11 DV-12 DV-13 DV-14 DV-15 DV-16 DV-17 DV-18 DV-19 DU-20 DV-21 DV-22 DU-23 DV-24 DV-25 Environmental Qualification Seismic Qualification Quality Assurance Redundancy Display Location Containment Pressure Subcooling Meter RCS Sampling/Boron Concentration Range RCS Sampling/Radioactivity Quench Tank Level S/G Wide Range Level CST Level Containment Spray Flow VCT Level Noble Gases and Vent Flow from Condenser Air Removal System Exhaust Noble Gases from S/G Safety Relief Valves RCS Sampling
- Chloride Content Containment Air - H2 Content Containment Air - 02 Content Indicating Lamps Centrifugal Charging Pump Flow Indication SI Pump Flow Indication Radiation Exposure Rate Position Indication - QCM-250 CCW Water Temperature Table Item C.2 Table Item D.13 Table Item D.15 Table Item D.19 Table Item D.20 Table Item D.26 Table Item E.3d Table Item E.3f Table Item Table Item Table Item Table Item and B.14 Table Item E.9d E.10a E.lob A.28, A.29 A.01 Table Item A.34 Table Item E.2 Table Item B.14 Table Item D.27 Various Various Various Various Various Table Items A.13/B.13 Table Item A.17 Table Item B.3 Section 2.1.1 Section 2.1.2 Section 2.1.3 Section 2.1.4 Section 2.1.5 Section 2.2.1 Section 2.2.2 Section 2.2.3 Section Section Section Section Section Section Section 2.2 '
2.2.5 2.2.6 2.2.7 2.2.8 2.2.9 2.2.10 Section 2.2.11 Section 2.2.12 Section 2.2.13 Section 2.2.14 Section 2.2.15 Section 2.2.16 Section 2.2.17 Section 2.2.18 Section 2.2.19 Section 2.2.20
ATTACHMENT TO AEPiNRC<0773AB Page ATT C H
N CH I SAFETY RELATED CABLE CH I NON-SAFETY RELATED CABLE CH II SAFETY RELATED CABLE CH II NON-SAFETY RELATED CABLE NON-SAFETY RELATED CABLE
<NO CHANNEL DESIGNATION)
CONTROL ROOM AND CABLE VAULT AREA CH I PROCESS TRANSMITTER (SAFETY RELATED)
PROCESS INPUTS FLOV, PRESSURF LEVE4 etc.
<SAFETY RE ATED>
REACTOR PROTECTION CABINETS SIGNAL ISOLATOR (I/I)
(SAFETY RELATED)
SOLID STATE PROT SYS (SSPS)
LOGIC CABINETS (SAFETY RELATED>
TO CIRCUIT BRKS, HOTORS; PUNPS, VALVES, RELAYS, etc.
COHTROL CABINETS CONPUTER INPUT CABINETS CONPU TER I/O CABINETS NOTD LOCATED IN AUL BLDG. HEAR COHTRQL ROON PROCESS
- INPUT, FLOV, PRESSURE LEVE4 etc.
REACTOR PROTECTIOH CABIHETS (NON-SAFETY RELATED>
SIGNAL ISOLATOR
(SAFETY RELATED>
(HOH-SAFETY RELATED)
L INDICATORS (HQN-SAFETY RELATED)
TO TSC CONPUTER (HQN-SAFETY RELATED)
TO PLANT CONPUTER (HOH-SAFETY RELATED)
TO RENQTE, NON-SAFETY RELATED Et>UIPHEHT Oe VALVES, etc>
(HQN-SAFETY RELATED)
TO CIRCUIT BRKS, HQTORS, PUNPS, VALVES, RELAYS, etc.
CH II PROCESS TRANSMITTER (SAFETY RELATED)
(SAFETY RELATED)
SOLID STATE PROT SYS <SSPS)
LOGIC CABINETS (SAFETY RELATED)
TYPICAL POST ACCIDENT MONITORING SIGNAL CABLING L HARDMARE LAYOUT F IGURE 2,1-i
t'l
DSGHARGE PRESSURE (r SK')
250 240 CONTAINMENTSPRAY PUNIP FLOW VS.
DISCHARGE PRESSURE Pi
~ ~
2s PJO o
IA 220 210 ANTlCPATED OPERATNG RANGE 190 0
1000 1500 2000 FLOW (GPM) 2500 3500
Page 26 Type A Var s:
Attachment to AEP:NRC:0773AB "those variables to be monitored that provide the p
y information required to permit the control room operator to take specific manually controlled actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for design basis accident events.
Primary information is information that is essential for the direct accomplishment of the specified safety funtions; it does not include those variables that are associated with contingency actions that may also be identified in written procedures."
Note:
These variables are plant-specific and based on review of the D.
C.
Cook Nuclear Plant Emergency Operating Procedures (EOP's) plus anticipated future changes to the EOP's.
The schedule and status for each instrument indicated is for when all of the applicable recommendations of Regulatory Guide 1.97, Rev 3 will be met.
Item No.
Purpose Variable Tag Nos.
Range E
S Q
S P
Display Q
Q A
F S
Location Remarks/Action Req'd U-1 U-2 Schedule Schedule A-1 Maintain Pressur-Centrifigal IFI-51,52, izer Level during Chg Pump Flow 53,54 S/G Tube rupture (CCP) 0-200 GPM A
A A
A A
Control Room No further action required CMPLT CMPLT Panel SIS See footnote (t)
(Deviation No.s DV-21 and DV-5)
A-2 Manual Trip of RC RCS Pressure NPS-121,122 0-3000psig A
A A
A A
Control Room No further action required CMPLT CMPLT Pumps based on RCS (wide range)
Panel RHR (Deviation No.s DV-1, DV-2, pressure DV-3, and DV-4)
A-3 A-4 NOT USED NOT USED A-6 NOT USED NOT USED A-7 Determination of required core exit temperature by S/G Pressure S/G Pressure MPP-210,211, 0-1200psig 212,220, 221,222, 230>231, 232,240, 241,242 A
A A
A A
Control Room No further action required CMPLT CMPLT Panel SG (Deviation Nos. DV-l, DV-2, DV-3, DV-4)
For Definite.on of "A" 'See Section 3.0
Attachment AEP:NRC:0773AB Page Item No.
Purpose Variable I
Tag Nos.
Range E
S Q
S P
Display A
F S
Location Remarks/Action Req'd U-1 U-2 Schedule Schedule A-8 Determination of Containment adverse containment Water Level NLA-320 NLI-321 599'-3" to 614 ft. el-evation (Containment Floor to max flood level)
A A
A A
A Control Room Replace transmitters.
Panel RHR See footnote (u)
(Deviation Nos. DV-1, DV-2 ~ DV-3,
& DV-5) 1 1989 re-CMPLT fueling outage A-9 Manual Reduction of S/G Level ECCS Flow Narrrow range (Secondary heat sink capability)
BLP-110,111, 112,120, 121,122, 130,131 I 132,140, 141,142 Prom below 1st stage separator to 2nd stage separator A
A A
A A
Control Room No further action required CMPLT Panel SG (Deviation No.s DV-1, DV-2, DV-3, DV-4)
CMPLT A-10 Manual Reduction of Pressurizer ECCS Flow Level NLP-151,152, 153 0-100X (96X of Total Volume)
A A
A A
A Control Room No further action required CMPLT CMPLT Panel PZR (Deviation No.s DV-1, DV-2, DV-3, and DV-4)
A-11 NOT USED A-12 Determination of adverse containment Containment Area Radiation Monitor High Range VRA-1310,
- 1410, (Unit 1)
- 2310, 2410 (Unit 2) 1 R/)R to 1X10 R/HR A
A A
A A
Control Room See footnote (v)
Panel RMS (Deviation Nos. DV-l, DV-2, DV-3, and DV-5)
NA NA A-13 Manually establish or trip containment spray Containment Pressure (Narrow range)
PPP-300,301, 302,303
-5 to +12 psig A
A A
A A
Control Room No further action required CMPLT CMPLT Panel SPY See Footnotes (d)
& (w)
(Deviation No. DV-6)
A-14 NOT USED For Definition of "A" See Section 3.0
Item No.
Attachment to AEP:NRC:0773AB urpose Page 28 Variable Tag Nos.
Range S
Q S
P Display Q
Q A
F S
Location Remarks/Action Req'd U-1 U-2 Schedule Schedule A-15 Manual Reduction of Auxiliary FFI-210,220, 0-2)0 ECCS Flow Feedwater Flow 230,240 x10 PPH (Secondary heat sink capability)
A A
A A
A Control Room No further action required Panel SG Redundancy provided by diverse variable - S/G narrow range level which is qualified (Deviation Noe. DV-1, DV-2, DV 3~
DV 4~
& DV 5)
CMPLT CMPLT A-16 Manual Transfer to RWST Level cold leg recircula-tion in low level in RWST ILS-950,951 essentially NA A
Top (bottom of overflow) to Bottom (100X of Total Volume)
A A
A Control Room No further action required CMPLT CMPLT Panel SPY (Deviation No.s DV-2, DV-3,
& DV-4)
A-17 Manual Trip or re-duction of Press-urizer Spray and ECCS Flow Degrees Sub-NA cooling 0-199'F sub-A NA cooling 0-199'F Superheat A
NA A
Control Room No further action required Panel BA See footnotes (b)
& (c) which apply beyond the isolating devices.
Also see footnote (x)
(Deviation No.s DV-7, DV-1, DV-3,
& DV-5)
CMPLT CMPLT A-18 A-19 A-20 A-21 A-22 A-23 A-24 NOT USED NOT USED NOT USED NOT USED NOT USED NOT USED NOT USED For Definition of "A" See Section 3.0
Attachment to AEP:NRC:0773AB Page 29 Item rpose No.
Variable Tag Nos.
Range E
g S
P Display A
F S
Location Remarks/Action Req'd U>>1 U-2 Schedule Schedule A-25 Manual Reduction Core Exit of ECCS Flow T/C's A-26 A-27 T/C 1-65 200-2300 F
A A
A A
A NOT USED NOT USED Control Room No further action required Panel FI (Deviation No.s DV-1, DV-2, (U-1) Panel
& DV-3)
RMS (U-2)
CMPLT CMPLT A-28 Manual trip of RCP's CCP Breaker Status Pump IE,IW 2E,2W OPEN/CLOSE NA A
A A
A Control Room Panel BA gualify or replace control room indicators with seismically qualified equipment See footnotes (d),(t)
& ()$)) (Deviation No.s DV-2, DV-3, DV-20 & DV-21)
End of 1988 End of 1988 A-29 Manual trip of RCP's A-30 A-31 A-32 A-33 SI Pump Pump 1N,1S Breaker Status 2N,2S OPEN/CLOSE NA A
A A
A Control Room Panel SIS NOT USED NOT USED NOT USED NOT USED gualify or replace control room indicators with seismically qualified equipment See footnotes (d), (y)
& (ggj) (Deviation No.s DV-2, DV-3, DV-20 & DV-22)
End of 1988 End of 1988 A-34 Manual Trip of RCPss Safety In)ec-IFI-260,266 0-800GPM tion Pump Flov NA A
A A
A Control Room No further action required CMPLT CMPLT Panel SIS See footnote (d)
& (y)
(Deviation No. DV-22 &
DV-5)
For Definition of "A" See Section 3.0
Item No.
/ /
/ ~
~
Attachment to AEP:NRC:0773AB
~
~
Purpose Page 30 Variable Tag Nos.
Range S
Q S
P Display A
P AS Location Remarks/Action Req'd Schedule Schedule A-35 S/G Blovdovn Radiation NOT USED - DELETED See footnote (z)
NA A-36 A-37 NOT USED NOT USED Por Definition of "A" See Section 3.0
Page Attachment to AEP:NRC:0773AB Type B Va es:
"those variables that provide information to indicat hether plant safety functions are (1) reactivity control,. (2) core cooling, (3) maintaining reactor coolant system integrity (including radioactive effluent control)."
Note:
The schedule and status applicable recommendations of Regulatory Guide 1.97, Rev 3 will be met.
are being accomplished.
Plant safety functions integrity, and (4) maintaining containment of each instrument is for when all of the Item Purpose No.
Variable Cat.
Tag Nos.
Range E
S Q
S P
Display A
F S
Location Remarks/Action Req'd U-1 U-2 Schedule Schedule B-1 Reactivity Neutron Control Flux 1
NE-21,23 10
-200Z A
power A
A A
A Control Room Panel NIS Install New, Cat.
1 1989 re-1990 re-Channel to provide fueling fueding Indication.
Upgrade outage outage 1 existing channel to meet Category 1 require-ments.
See footnote (aa)
(Deviations No.s DV-1, DV-2 & DV-3)
B-2 Control Rod Position 3
CA1-8, CBl-4 Full in or NA CC1-8, CD1-9 not full in SA1-8. SB1-8 NA NA NA NA Control Room None required Panel RC NA B-3 RCS Soluble.
Boron Con-centrate 3
NSX-101,103 375-2000 NA ppm NA NA NA NA NA See Item C-2 and footnotes (a)
& (bb)
(Deviation No. DV-8)
NA NA B-4 RCS Cold Leg Temp-erature 1
NTR-210,230, 0-700'F A
A A
A A
Control Room Panel DTU Relocate R/I and I/I converters to control room.
Upgrade cable to meet Category 1 require-ments.
See footnote (cc)
(Deviation No.s DV-1, DV-2, DV-3 & DV-4) 1989 re-1990 re-fueling fueling outage outage B-5 Core Cooling RCS Hot Leg Water Temperature 1
NTR-1100130, 0-700 F
A A
A A
A Control Room Panel DTU Relocate R/I and I/I converters to control room. Upgrade cable to meet Category 1 require-ments.
See footnote (cc)
(Deviation No.s DV-1, DV-2~ DV-3 & DV-4) 1989 re-1990 re-fueling fueling outage outage For Definition of "A" See Section 3.0
Attachmen to AEP:HRC:0773AB Page 32 Item No.
ose Variable Cat.
Tag Nos.
Range S
Q S
P Display A
F' Location Remarks/Action Req'd U-l U-2 Schedule Schedule B-6 RCS Cold 1
Leg Water Temperature See item B-4 B-7 B-8 RCS Pressure Core Exit Temperature See item A-2 1
See item A-25 B-9 B-10 Coolant Inventory Degrees of Subcooling NLI-110,111 120,121i 130,131 Top of head vent piping to bottom of vessel (100X of Volume)
A A
A A
A Control Room Upgrade Power Supply Panel SIS to Hot Leg Temperature Input.
See Item B-4 which provides an input to this variable (Devia-tion No.s DV-1, DV-2, DV-3 & DV-5)
See item A-17 1989 re-1990 re-fueling fueling outage outage B-11 Maintaining RCS Pressure RCS Integrity See item A-2 B-12 Containment 2
NLA-310 Sump Water HLI-311 Level 589'-6" to 599'-8" (Bottom of Sump to Containment Floor)
A NA A
NA NA Control Room See footnote (u)
& (dd)
Panel RHR (Deviation Ho.s DV-1 &
DV-3) 1989 re-CMPLT fueling outage B-13 Containment Pressure (Wide Range) 1 PPA-310,312
-5 to 36psig NA A
A A
A Control Room No further action Panel SPY required.
Also see Item A-13 for narrow range.
See footnotes (d) and (w) narrow range (Deviation No. DV-6)
NA NA For Definition of "A" See Section 3.0
Attachment to AEP:NRC:0773AB Page 33 Item se No.
Variable Cat.
Tag Nos.
Range S
Q S
P Display A
P
~
S Location Remarks/Action Req'd U-1 U-2 Schedule Schedule B-14 Naintaining Containment Integrity B-15 Containment Isolation Valve Posi-tion (ex-cluding check valves)
Containment Pressure 1
See listing in Attach-ment 1 to these tables CLOSED-NOT CLOSED A
A A
A A
Control Room Replace Limit switches Panels IV, with env
& seismically BA,SIS,SPY
. qualified devices as noted in Attachments 1
2 to these tables.
Qualify or replace control room indicators with seismically quali-fied equipment.
See foot-note (ee)
& ($/))
(Deviation No.s DV-20 DV-1, DV-2 & DV-3)
See items A-13 and B-13 1989 re-1990 re-fueling fueling outage outage For Definition of "A" See Section 3.0
Attachment to AEP:NRC:0773AB Page 34 Type C Va es:
"those variables that provide information to indicat e potential for being breached or the actual breach of the barriers to fission product releases.
The barriers are (1) fuel cladding, (2) primary coolant pressure
- boundary, and (3) containment."
Note:
The schedule and status for each instrument is for when all of the applicable recommendations of Regulatory Guide 1.97, Rev 3 will be met.
Item No.
Purpose Variable Cat.
Tag Nos.
Range E
S Q
S P
Display A
F S
Location Remarks/Action Req'd U-l U-2 Schedule Schedule C-l C-2 Fuel Cladding Core Exit Temperature Radioactive concentra-tion or Radiation Level in Circulating Primary Coolant 1
NSX-101,103 NA A
A A
A A
See item A-25 i No further action required.
See footnotes (h) and (ff)
(Deviation No.s DV-9, DV-1, DV-2, DV-3 6 DV-5)
CMPLT CMPLT C-3 Analysis of Primary Coolant (Gamma Spectrum)
See item C-2 and footnote (gg)
C-4 Reactor Cool-RCS Pres-ant Pressure sure Boundary See item A-2 C-5 Containment Pressure See items A-13 and B-13 C-6 Containment Sump Water Level See item B-12 C-7 Containment Area Radi-ation See item A-12 For Definition of "A" See Section 3.0
Attachment AEP:NRC:0773AB Page Item No.
ose Variable Cat.
Tag Nos.
Range E
S P
Display A
P S
Location Remarks/Action Req'd U-l U-2 Schedule Schedule C-8 Effluent 3
Radioactiv-ity-Noble gas Effluent from Con-denser Air Removal Sys-tem Exhaust SRA-1900 5.8E-7 to (Unit 1
1.86E4 SRA-2900 uCi/cc (Unit 2)
NA NA NA NA NA Control Room No action required CT-1 Control Terminal NA NA C-9 Containment RCS Pressure See item A-2 C-10 Containment 1
Hydrogen Concentra-tion ESR-1 thru 9
0-30 Volume X
NA A
A A
A Control Room See footnote (d)
Panel IV No action required (Deviation No.s DV-1, DV-2,
& DV-3)
NA NA C-11 Containment Pressure See items A-13 and B-13 C-12 Containment 2
Effluent Radioactiv-ity-Noble gases from identified release points VRS-1500, 5.8E-07 to NA NA A
(Unit 1) 1.86E4 uCi/cc VRS-2500, (Unit 2)
NA NA Control Room See footnotes (d) 6 (hh)
NA CT-1 (Deviation No. DV-3)
Terminal No Action Required NA For Definition of "A" See Section 3.0
Attachment to AHP:NRC:0773AB Page 36 Item pose No.
Variable Cat.
Tag Nos.
Range E
S P
Display Q
A F
. S Location Remarks/Action Req'd U-1 U-2 Schedule Schedule C-13 Effluent Radioactiv-ity-Noble Gases (from buildings or areas where penetra-tions and hatches are
- located, eg, secondary containment and AUX buildings that are in direct con-tact with primary containment SEE ITEM C-12 For Definition of "A" See Section 3.0
Attachment o AEP:NRC:0773AB Page 37 Type D Va es:
"those variables that provide information to indicat he operation of individual safety systems and other systems important to safety.
These variables are to help the operator make appropriate decisions in using the individual systems important to safety in mitigating the consequences of an accident."
Note:
The schedule and status for each. instrument is for when all of the applicable recommendations of Regulatory Guide 1.97, Rev 3 will be met.
Item Purpose No.
Variable Cat.
Tag Nos.
Range E
S g
S P
Display A
F S
Location Remarks(Action Req'd U-1 0-2 Schedule Schedule D-1 RHR System RHR System Flow 2
IFI-310,311, 320,321 0-1500 GPM A
1500-5000 GPM NA A
NA NA Control Room No Action Required Panel RHR See footnotes (jj) &
(mmm)
(Deviation No. DV-3)
NA NA D-2 RHR Heat Ex-change Out-let Temp 2
ITI-310,320 0-400 F
A NA A
NA NA Control Room Panel RHR No further action required (See footnote (ii)
(Deviation No.s DV-1, DV-3)
CMPLT CMPLT D-3a SI Systems Accumulator Tank Level 2
ILA-110,111 120,121, 130,131, 1401141 4.148 to A
120.8 in.
(wide range)
(52X of Total Volume) 104.15 to 120.8 in.
(narrow range)
(7.5X of Total Volume)
NA A
NA NA Control Room Panel SIS Replace the wide range 1989 re-1990 re-transmitter each tank fueling fueling with env. qual.
equipment outage outage See footnotes (iii) gualification shall apply to the wide range instruments only.
(Deviation No.s DV-1, 6
(DV-3)
D-3b Accumulator Tank Pres-sure 2
IPA-110,111, 0-800 psig A
120,121, 130,131, 140,141 NA A
NA NA Control Room Panel SIS Replace one transmitter(
1989 re-1990 re-tank with env. qualified fueling fueling equipment.
Qualification outage outage shall apply to only one instrument/tank.
(Deviation No.s DV-1, DV-3)
D-4 Accumulator Tank Isola-tion Valve Position 2
IM0-110,120, 130,140 Closed or NA Open NA A
NA NA Control Room None Required NA Panel SIS See footnote (r)
& (kk)
(Deviation No. 3)
NA For Definition of "A" See Section 3.0
Attachmen to AEP:NRC:0773AB Page 38 Item No.
pose Variable Cat.
Tag Nos.
Range Q
S P
Display A
F S
Location Remarks/Action Req'd U-l U-2 Schedule Schedule D-5 D-6 D<<7 D-8 Boric Acid Charging Flow Flow in HPI System Flow in LPI System RWST Level See item D-24 See item A-1 See item D-1 See item A-16 D-9 Primary Cool-RCP Status 3
QI,Q2,Q3,Q4 ant System 0-1200A NA NA NA NA NA Control Room No Action Required Panel RCP NA NA D"10 Primary Sys-tem Safety Relief Valve Positions or Flow Thru or pressure in Relief Valve Lines 2
QR-107 A,B, NA C,D A
NA A
NA NA Control Room See footnote (b)
Panel RC No Action Required (Deviation No.s DV-l, DV-3)
NA NA D-11 Pressurizer Level Pressurizer Heater Sta-tus 2
Group Al,A2, ON/OFF A3,C1,C2,C3 See item A-10 NA NA A
NA NA Control Room See footnotes (d) 6 (ll)
NA Panel PZR No Action Required (Deviation No. DV-3)
NA D-13 Quench Tank Level 3
NLA-351 7 inches above tank bottom to 7 inches below tank top NA NA NA NA NA Control Room None required Panel PZR See footnote (mm)
(Deviation No. DV-10)
NA NA D-14 Quench Tank 3
NTA-351 Temperature 50-750 F
NA NA NA NA NA Control Room No further action Panel PRZ required CMPLT CMPLT For Definition of "A" See Section 3.0
Page 39 Item No.
Variable Cat.
Tag Nos.
Range E
S P
Display A
F S.
Location Remarks/Action Req'd Schedule Schedule D-14a quench Tank Pressure 3
NPA-351
-10 to 100 NA psig NA NA NA NA Control Room None Required Panel PRZ NA NA D-15 Secondary Sys-S/G Level tern (Steam Generator) 1 BLI-110,120, 130,140 (wide range)
From 12" A
above tube sheet to sep-arators A
A A
A Control Room None required See Panel SG footnotes (i) 6 (oo)
(Deviation No. DV-11)
NA NA D-16a D-16b D-17 Safety/Re-lief Valve Positions Main Steam Flow Main Feed-water Flow 2
MFC-110 F 111, 0-4xl0 PPH A
6 120 '21, 130 '31, 140,141 3
FFC-210,211, 0-4x10 PPH NA 6
220,221, 230,231, 240,241 See Item D<<16b for alternate instrumentation NA A
NA NA Control Room None required Panel SG (Deviation No.s DV-1, DV-3)
NA NA NA NA Control Room None required Panel BA See footnote (pp)
NA NA D-18 Auxiliary Feed-Aux Feed-water System water Flow See item A-15 D-19 Condensate Storage Tank Level 1
CLI-113,114 CLR-110, 111
'\\
Essentially Top to Bot-tom (95K Total Volume)
A A
A A
A Control Room No further action Panel CP required.
See footnotes (qq) 6 (mmm)
(2 channels only)
(Deviation No. DV-12)
CMPLT CMPLT D-20 Containment Containment Cooling System Spray Flow 2
IFI-330,331 (Upper con-tainment) 0-2500GPM NA NA A
NA NA Control Room No Action Requ'ired.
Panel SPY See footnotes (d), (e),
and (rr)
(Deviation No.s DV-13, DV-3)
NA NA For Definition of "A" See Section 3 0
Attachment o AEP:NRC:0773AB Page 40 Item No.
ose Variable Cat.
Tag Nos.
Range E
S P
Display A
P S.
Location Remarks/Action Req'd U-1 U-2 Schedule Schedule D-21 D-22 Heat Removal by contain-ment Heat Re-moval System Containment Atmosphere Temperature 2
ETR-11,12,13, 0 to 14,15,16, 400'F 17,18,19, 20,21,22; 23,24,25, 26,27 A
NA A
NA NA Control Room Panel A-14 D. C. Cook Nuclear Plant NA Units 1&2 do not have a
Containment Heat Removal System, therefore this item does not apply 1
Replace six (6) tech spec 1989 re-related RTD's with envi-fueling ronmentally qualified outage equipment.
Increase range from -0 to 300'P to as specified. Qualification shall apply for the six (6)
T.S. related instr. only.
(Deviation No.s DV-1, DV-3)
NA 1990 re-fueling outage D-23 Containment 2
ITR-311,321 50 to Sump Water 400 P
Temperature A
NA A
NA NA Control Room Panel RHR Replace RTD's with quali-1989 re-fied equipment.
See fueling footnote (ss) outage (Deviation No.s DV-1, DV-3) 1990 re-fueling outage D-24 Chemical end Make up Volume Control Flow-In System 2
QFI-200 0-200 GPM NA NA A
NA NA Control Room Panel BA No Action Required See NA footnote (d)
(Deviation No. DV-3)
NA D-25 Letdown Flow 2
QFI-301 Out 0-200 GPM NA NA A
NA NA Control Room No Action Required Panel BA and See footnote (d)
HSD (Deviation DV-3)
NA NA D-26 Volume Con-trol Tank Level 2
OLC-451,452 Essential-ly top to bottom (65X of Total Volume)
NA NA A
NA NA Control Room Panel BA No Action Required See footnote (d),
and (tt)
(Deviation No.s DV-14
& DV-3)
NA NA D-27 Cooling Water System CCW water Temperature to ESF Sys-tem 2
CTR-410,415, 420,425 0-200'P A
NA A
NA NA Control Room No Action Required NA Panel ESW See footnote (d)
& (ill)
(Deviation No. DV-3 &
DV<<25)
NA For Definition of "A" See Section 3.Q
Attachment to AEP:NRC:0773AB Item se No.
Page 41 Variable Cat. Tag Nos.
Range E
Q S
P Display Q
Q A
F S.
Location Remarks/Action Req'd U-1 U-2 Schedule Schedule D-28 CCH Flow to ESF System 2
CFI-410,419, 0-10000GPM A
NA A
NA NA Control Room 420,429 0-6000GPM Panel CCM No Action Required See footnote (mmm)
(Deviation DV<<3)
NA NA D-29 Radwaste Systems High Level Radioactive Liquid Tank Level RLS-255,256 Essentially NA NA Top to Bot-tom (84X of Total Volume)
NA NA NA Panel HDG None required NA D-30 Radioactive Gas Holdup Tank Pres-sure 3
RPC-310,320, 330,340, 350,360, 370,380 0-225 psig NA NA NA NA NA Panel M)G No further action required CMPLT CMPLT D-31 Ventilation System Power Supplies Emergency Ventilation Damper Pos-ition Status of Standby Power and other Energy Sources Im-portant to Safety 2
VCR-201 thru Open-Closed A
NA 207 A
NA NA Control Room Panel IV VCR-207 on Control Room Panel SPY Replace Limit Switches VCR-201 6 202 with env.
qualified equipment.
Footnote (d) applies to VCR-203 thru 207.
See footnote (s)
(Deviation Nos. DV-l, DV-3) 1989 re-fueling outage 1990 re-fueling outage D-32a Diesel Cen 2
DGlAB Status DCICD 0-800A NA NA A
NA NA Control Room Panel SA See footnote (d)
No Action Required (Deviation No. DV-3)
NA NA D-32b 4KV Safety 2
Bus T11A, 0-150V Related Po-
- T11B, Power Systems Tllc Status T11D NA NA A
NA NA Control Room Panel SA See footnote (d)
No Action Required (Deviation No. DV-3)
NA NA For Definition of "A" See Section 3.0
Attachment to AEP:NRC:0773AB Page 42 Item ose No.
D-32c Variable Cat. Tag Nos.
Range 250VDC 2
Battery AB 0-300V Battery Power Battery CD System Status E
S P
Display R
0 A
F S
Location NA NA A
NA NA Control Room Panel SA Remarks/Action Req'd See footnote (d)
No Action Required (Deviation No. DV-3)
U-1 U-2 Schedule Schedule NA NA D-32d 120VAC Safe-2 Channel I,II, 0-150V ty Related III, IV Power System Status NA NA A
NA NA Control Room Panel SA See footnote (d)
No Action Required (Deviation No.> DV-3)
NA NA D-32e Instrument Air Status 2
XPI-100 XP!-50 XPI-20 XPI-85 0-150psig 0-100psig 0-60psig 0-160psig NA NA A
NA NA Control Room Panel SV These are mechanical devices - no electrical components No Action Required (Deviation No. DV-3)
NA NA For Definition of "A" See Section
~ F 0
Attachment to AEP:NRC:0773AB Page 43 yp
" hose variables to be monitored as required for use determining the magnitude of the release of radioactive materials d
continually assessing such releases."
Note:
The schedule and status for each instrument is for when all of the applicable recommendations of Regulatory Guide 1.97, Rev.
3 will be met.
Item No.
Purpose Variable Cat.
Tag Nos.
Range E
S Q
S P
Display A
F S
Location Remarks/Action Req'd U-l U-2 Schedule Schedule E-l Containment Radiation Containment Area Radia-tion High Range See item A-12 i E-2 Area Radiation Radiation 3
Exposure Rate (in-side build-ings or where areas of access are required to service equipment important to.
safety)
ERA-7303 See Foot-NA NA thru 7308 note (kkk)
ERA-8303 thru 8308 ERS-7401 ERA-7403, 7404 ERS-8401 ERA-&403, ERA-7504,
- 7507, 75Q8 ERA-7601 thru 7605 NA NA NA Control Room Install new monitors CRT See footnote (uu)
(kkk) (Deviation No.
DV-23)
By end of By end of 1989 19&9 E-3a Noble Gases and vent Flow Rate Containment 2
or Purge Effluent SEE ITEM E-3e E-3b Reactor Shield Building Annulus SEE ITEM E-3e E-3c Aux Building 2
SEE ITEM E-3e E-3d Condenser 2
Air Removal System Ex-haust SRA-190Q (Unit 1)
SRA-2900 (Unit 2)
SFR-401 5.8E-07 to 1.86E4 uCi/cc 0-250 scfm NA NA A
NA NA See footnotes (d),
(ww) and (xx)
(Deviation No.s DV-15
& DV-3)
NA NA For Definition of "A" See Section 3.0
At'tachme to AEP:NRC:0773AB Page 44 Item No.
rpose Variable Cat.
Tag Nos.
Range E
S Q
S P
Display Q
A F
Location Remarks/Action Req'd U-1 U-2 Schedule Schedule E-3e Common Plant Vent VRS-1500 (Unit 1)
VRS-2500 (Unit 2)
VRF-315 5.8E-07 to 1.86E4 uCi/cc 0-200K scfm NA NA A
NA NA Control Room None Required CT-1 Control See footnote (d)
Terminal (Deviation DV-3)
NA E-3f Vent from 2
S/G Safety Relief Valves MRA-1600 1700 (Unit 1)
NRA-2600 2700 (Unit 2) 0.01 to 100 uCi/cc NA NA A
NA NA Control Room Panel RMS None Required See footnote (yy)
(Deviation No.s DV-16 6 DV-3)
NA NA E-3g Other ident-2 ified re-lease points SRA-1800 (Unit 1)
SRA-2800 (Unit 2)
SFR-201 5.8E-07 to 1.86E4 uCi/cc 0-1000 scfm NA NA A
NA NA Control Room None Required Panel FI See footnote (d)
(Deviation No. DV-3)
NA NA E-4 Particulates and Halogens All identi-3 fied release points (ex-cept S/G Safety Re-lief valves and Conden-ser air re-moval System exhaust)
Sampling and onsite analysis See Item E-3e E-5a Environ Radia-tion and Radio-activity Airborne Radioacti-vity and Particulates sampling and analysis (portable) 3 NA 1E-9 to NA NA NA NA NA NA 1E-3 uCi/cc (minimum)
None Required NA NA For Definition of "A" See Section 3.0
~WE ~~fly%%tl<<+
Attachment t AEP:NRC:0773AB Page 45 Item No.
P e
Variable Cat.
Tag Nos.
Range E
Q S
P Display A
F S -
Location Remarks/Action Req'd U-1 U-2 Schedule Schedule E-5b Plant and 3
Environs Radiation (Portable)
NA Gamma 1.0E-3 to 1.0E4 R/hr.
Beta/low energy gama 1.0E-3 to 1.0E4 Rad/hr NA NA NA NA NA NA Completion of Calibration End of End of 89 89 E-5c Plant and Environs Radioacti-vity (Port-able)
NA Isotopic NA NA NA NA NA NA Analysis No further action required.
See footnote (zz)
CMPLT CMPLT E-6 E-7 Heteorology Wind Direc-tion Wind Speed 3
EFR-410,412, 0-360 413.414 3
EFR-400,404, 0-100 mph 402,403 NA NA NA NA NA Control Room None required Panel Flx and/or CRT NA NA NA NA NA Control Room None required Panel Flx and/or CRT NA NA NA NA E-8 Estimation of Atmos-pheric Stability 3
ETR-400,402,
-30 to 50'C NA NA NA NA NA Control Room None required 403 Panel Flx ETQ-401 and/or CRT NA NA E-9a Accident Samp-Gross Acti-ling Primary
-vity Coolant and Sump 3
NSX-101,103 1 uCi/ml to NA NA NA NA NA NA ESX-400 10 Ci/ml See Item C-2 See footnote (aaa)
NA NA E-9b Gamma Spec-trum 3
NSX-101,103 ESX-400 0.050 to 2.05 HeV Isotopic Analysis NA NA NA NA NA NA See item C-2 See footnote (bbb)
NA NA For Definition of "A" See Section 3.0
Attachment to AEP:NRC:0773AB Page 46 Item No.
ose Variable Cat.
Tag Nos.
Range S
Q S
P Display Q
A
. F S
Location Remarks/Action Req'd U-1 V-2 Schedule Schedule E-9c E-9d Boron Con-3 NSX-101,103 375 to 2000 NA NA NA NA NA NA tent ESX-400 ppm Chloride 3
NSX-101,103 0.01 to NA NA NA NA NA NA Content ESX-400 20 ppm See Item C-2 See footnote (ccc)
& (bb)
SEE ITEM C-2 (See footnote (ddd)
(Deviation No. DV-17)
NA NA E-9f Dissolved 3
H2 or total gas NSX-101,103 0-2000 cc/
NA NA NA NA NA NA ESX-400 kg SEE ITEM C-2 1
NA E-9g gissolved 2
3 NSX-101,103 0-20 ppm NA NA NA NA NA NA ESX"400 SEE ITEM C-2 NA NA'-9h pll NSX-101,103 ESX-400 1.0 to 13.0 NA NA NA NA NA NA pH SEE ITEM C-2 See footnote (eee)
NA NA E-10a Containment Air H2 Content, 3
ESX-001 NA NA NA NA NA NA NA None required See footnote (fff)
(Deviation No. DV-18)
NA NA E-10b 02 Content 3
NA NA NA NA NA NA NA NA None required See footnotes (q) and (ggg)
(Deviation No. DV-19)
NA NA E-10c Gamma Spec-3 ESX-001 trum 1 uCi/cc to NA NA NA NA NA NA 10 Ci/cc Isotopic Analysis None required See footnote (hhh)
NA For Definition of "A" See Section 3.0
ATTACH TO AEP I NRC'773Ab Page 47 ATTACHMENT NO.
1 TO TYPE B ABLES TABLE ITEM NO. B-14 CONTAINMENT ISOLATION VALVES Plant ID Channel Plant ID Channel YCR-20 YCR-21 XCR-100 XCR-101 XCR-102 XCR-103 GCR-301 Cctt-451 CCN-452 CClt-453 CCtt-454 CCN-458 CCM-459 ECR-31 ECR-32 ECR-33 ECR-35 KCR-36 VCR-901 VCR-905 VCR-909 VCR-913 VCR-900 VCR-904 VCR-908 VCR-912 VCR-902 VCR-903 VCR.906 VCR-907 VCR-910 VCR-911 VCR-914 VCR-915 VCR-921 VCR-925 VCR-929 VCR-933 VCR-920 VCR-924 VCR-928 Glycon Return fros Contsineent Glycon Return fros Contsinaent Cntrl Air to Cntnsent Islstion Ylvs Cntrl Air to Cntnsent Islstion Ylvs Cntrl Air to Cntnsent Ielstion Ylvs Cntrl Air to Cntnsent Islstion Ylvs M2 Supply to Pressurixer Relief TnM CCV fr RCPs Lvr Guide Bearing Coolr CCV fr RCPs Lvr Guide Bearing Coolr CCV iros RCPs Therssl Barriers CCV froa RCPs Therasl Barriers CCV to RCPs Oil Coolers snd Therasl CCV to RCPs Oil Coolers snd Therssl Barriers Contsinsent Air Monitor Contsinsent Air Monitor Contsinaent Air ttonitor Contsinsent Air Monitor Contsinsent Air Monitor NESV to CLY Unit 1
NES'V to CLY Unit 2 NESV to CLY Unit 3
'MESV to CLY Unit 4 NKSV to RCP CI.Y Unit 1
MESV to RCP Ct.Y Unit 2 NESV to RCP CLY Unit 3 MESV to RCP CLY Unit 4 MESV froa CLY Unit 1
MESV froa CLY Unit 1 MESV froa CLY Unit 2 MESV fros CLY Unit 2 NESV froa CLY Unit 3 NESV fros CLY Unit 3 NESV fros CLY Unit 4 NESV fros CLY Unit 4 NESV to CUY Unit 1 NESV to CUY Unit=2 NESV to CUY Unit 3 NESV to CUY Unit 4 IIESV to RCP CUY Unit 1
NESV to RCP CUY Unit 2
,NESV to RCP CUY Unit 3 VCR-932 VCR-922 VCR-923 VCR-926 VCR-927 VCR-930
'VCR-931 VCR-934 VCR-935 VCR-941 VCR-942 VCR-943 VCR-944 VCR-945 VCR 955 VCR-946 ltCR-956 VCR-947 VCR.957 VCR-948 VCR.958 VCR-951 VCR-952 VCR-953 VCR-954 VCR-960 VCR-961 VCR-964 VCR-965 VCR-962
'VCR-963 VCR-966 VCR-967 CCR-440
.CCR-441 NCII-221 NCM-231 CClt-430 CCM-431 CCN-432 CCtt-433 NESV to RCP CUY Unit 4 NESV froa CUY Unit 1
NESV froa CUY Unit 1
NESV froa.CUY Unit 2 NESV fros CUY Unit 2 NESV froa CUY Unit 3 MESV fros CUY Unit 3 NESV fros CUY Unit 4 NESV froa CUY Unit 4 NKSV to RCP 1 Air Cooler NKSV to RCP 2 Air Cooler NESV to RCP 3 Air Cooler NESV to RCP 4 Air Cooler
-NESV fros RCP 1 Air Cooler NESV froa RCP 1 Air Cooler NESV froa RCP 2 Air Cooler NESV froa RCP 2 Air Cooler NESV froa RCP 3 Air Cooler NKSV froa RCP 3 Air Cooler NESV fros RCP 4 Air Cooler NKSV froa RCP 4 Air Cooler MESV to RCP 1 Air Cooler NESV to RCP 2 Air Cooler NKSV to RCP 3 Air Cooler NESV to RCP 4 Air Cooler NESV to Instrsnt Rs Yntilstn Unite MES'V to Instrsnt Ra Yntilatn Units NESV to Instrsnt Rs Yntilstn Units MESV.to instrsnt Ra Yntilstn Units NESV fr Instrsnt Rs Yntilstn Units NKSV to Inatrsnt Rs Yntilstn Units NES'V to Instrsnt Ra Yntilatn Units MESV fr Instrsnt Ra Yntilstn Units CCV fros Nein Stesa Penetration CCV froa Main Stess Penetration Nein Stess to Auxiliary Feed Pusp ltsin Stess to Auxiliary Feed Pusp CCV to E Pressure Equslixstion Fsn CCV to E Pressure Equslixstion Fsn CCV to V Pressure Equslixstion Fsn CCV to V Pressure Equslixstion Fsn
'1
ATTACH>
TO AEP:NRC10773AB Page 48 ATTACllMLrNT NO.
1 TO TYI'Lx 9
'ADLL'S TABI E ITLM NO. 8-l4 CONTAINMENT ISOLATION VALVES Plant ID Channel Plant ID Channel CCR-A55 CCV to Reactor Supports CCR-A56 CCV froa Reactor Supports CCR-157 CCV froa Reactor Supports CCR-460 CCV fros Excess Letdovn Hx CCR-462 CCV to Excess Letdovn Hx DCR-201 RC Drain Tank to Vent Header
))Cl)-202 NC Drnln Tnnk tu Cnu Analyzer Ix:I)-2))'I IH: l)rnln Tn))k tn Vv))t Hu()dcl'CR-20'C Drain Tank to Oas Analyzer DCR-205 RC Drain Tank Suction Isolation DCR-206 RC Drain Tank Suction Isolation DCR-207 Nitropen Supply to RC Drain Tank DCR-301 S/0 Blovdovn Seeps Isolation DCR-302 S/0 Blovdovn Saaple Isolation DCR-303 S/0 Blovdovn Sasple Isolatfon DCR-301 S/0 Blovdovn Saaple Isolation DCR-310 S/0 If, l2, i3 6 IA Blovdovn Yalves OCR-320 S/0 ef, A2, e3 C i4 Blovdovn Valves DCR-.330 S/0 il, 82, A3 6 ii Blovdovn Valves DCR-340 S/0 tf, i2, l3 6 li Blovdovn Valves DCR-600 Containsent Susp to Vesta Holdup DCR-601 Containaent Susp to Vesta Holdup DCR-610 Ice Condenser Drain to Drain Header DCR-611 Ice Condenser Drain to Drain Header DCR-620 Cntnaent Ventilation Drain to Hldup DCR-621 Cntnsent Ventilation Drain to Hldu KCR-up ECR-41
-<f6 Containsent Liquid Saaplfng S
t 7
Containsent Liquid Sasplfng Systes ng ys ea KCR-496 Containsent Liquid Saapllnp S
t
-i97 Contafnaent Liquid Saaplfnp Systea ECR-5 KCR-535 Containeent Oae Saapli S
t np ya ea OCR-91
- 36 Contafnsent Oas Saapling Systea 9
Cont.
Desin.
Cleanup Vater Isol.
OCR-920 Cont. Basin.
Cleanup Vater Isol.
KCR-10 ECR-11 ECR-12 KCR-13 KCR-Ii ECR-15 ECR-16 KCR-17 KCR-18 ECR-19 ECR-20 ECR-21 ECR-22 ECR-2(
ECR-23 KCR-25 ECR-26 ECR-27 ECR-28 ECR-29 GCR-31(
ICR-5 ICR-6 HCR-251 NCR-252 HCR-253 HCR-25(
NCR-105 NCR-106 NCR-107 NCR-108 NCR-109 NCR-110 OCR-300 OCR-301 NCR-252 OCN-250 OCN-350 RCR-100 RCR-101 VCR-10 VCR-11 Cntnsent H2 Sasple 6 Return Valves Cntnsent H2 Sssple 6 Return Valves Cntnaent H2 Sasple 6 Return Valves Cntnsent H2 Sasple 6 Return Valves Cntnaent H2 Saaple 6 Return Valves Cntnsent H2 Sasple 6 Return Valves Cntnsent H2 Saaple 6 Return Valves Cntnsent H2 Sasple 6 Return Valves Cntnaent H2 Seeps 6 Return Valves Cntnsent H2 Saspl ~ 6 Return Valves
,Cntnsent H2 Saaple 6 Return Valves
)Cntnaent H2 Sasple 6 Return Valves
)Cntnaent H2 Sasple C Return Valves
)Cntnunt H2 Saaple 6 Return Valves iCntnsent H2 Saapl ~ 6 Return Valves
)Cntnaent H2 Saaple 6 Return Yalves
)Cntnsent H2 Saaple 6 Return Valves Cntnaent H2 )aspic 6 Return Valves Cntnsent H2 Saaple 6 Return Valves Cntnsent H2 Saaple 6 Return Valves iNftrogen Supply to Accusulators
~Accusulator Sasple Valves Accusulator Sasple Valves iSasple fros Hain Steas Lines
.Sasple fros Hain Steas l.ines Sasple fraa Hain Steas Lines Sasple fros Hain Steaa Lines Priaary Systes Hot Lep Saaple Prisary Systes Hot Leg Sasple iPressurlxer Liquid Saaple Pressurizer l.iquid Sasple Preseurixer Steas Saspl>>
Pressurlxer Steaa Saaple Letdavn Line Isolation Yalve
.Letdovn Line Isolation Valve
)Prsary Nkup H20 to Prsarzer Rlf Tnk
)RCP Seal H20 Return Isolation Valve iRCP Seal H20 Return Isolation Valve iPrssrxer Rlf Tnk to Gas Analyzer Prssrxer Rlf Tnk to Oaa Analyzer Olycon Supply to Containsent Olycon Supply to Containaent
ATTACHMENT TO AEP:NRC:0773AB Page 49 ATTACHMENT NO.
2 TO TYPE B VARIABLES TABLE ITEM NO. B-14 Most of the valves listed in Attachment No.
1 to Type B Variables Item No. B-14 are located outside of containment and are only subject to an HELB outside of containment.
Since they are not required to operate (per EOP's) in the event of an HELB, qualification is not required because they will be located in a mild environment should a Design Basis Event (DBE) occur where their use is required.
Redundant indication is also not required because the valves are backed up by a second, redundant valve, also listed in Attachment No.
1.
Exceptions to the above are listed as follows:
DCR-301,302,303,304,310,320,330,340, and XCR-100,101,102,103 will be upgraded by replacing the position indication limit switches with environmentally and seismically qualified equipment.
These valves are located outside of containment and are required to operate during an HELB.
The position indication limit switch for QCM-250 is located outside containment and is not qualified for a DBE inside of containment.
It is,
- however, backed up by QCM-350 which is located outside of containment and its position indication limit switch is qualified.
Should there be a
DBE inside containment and QCM-250 did not indicate appropriately, QCM-350 could be used to verify isolation.
In addition an operator could be dispatched to visually verify QCM-350 closure.
If a DBE outside of containment should
- occur, QCM-350 is qualified for the adverse environment generated and could be backed up by QCM-250 which would not be subjected to the harsh environment and therefore, would be expected to indicate appropriately.
See footnote (ee)
The position indication limit switches for VCR-11 and 21 are located inside containment and not qualified for a
DBE inside containment.
They
- are, however, backed up by VCR-10 and 20, respectively, which are located outside of containment.
Should there be a DBE inside containment and VCR-ll or 21 did not indicate appropriately, VCR-10 and 20 can be used to verify isolation.
In
- addition, an operator could be dispatched to visually verify closure.
DBE's outside of containment will not create a
harsh environment at VCR-10 and 20 locations.
See footnote (ee)
CCM-451 452 453,454,458,459; CCM-430,431,432,433; and MCM-221,231 are qualified for an HELB.
VCR-101 thru 107 and 201 thru 207 are not listed in Attachment 1
because their function is specifically listed in the Type D
Variables Table, Item D-31.
ATTACHMENT TO AEP:NRC:0773AB Page 50 4.0 FOOTNOTES DESIGNATING DEVIATIONS TO REGULATORY GUIDE 1.97, REV.
3 (a)
The automatic in)ection of boric acid into the RCS by the safety injection system following a postulated LOCA/HELB will be monitored and verified through the use of qualified instrumentation.
In addition, since all sources of water for the safety infection system (Accumulators, Boron Injection Tank and Refueling Water Storage Tanks) are required by Technical Specifications to contain boric acid solution of a minimum concentration, the proper operation of the ECCS ensures an adequate boron concentration in the reactor coolant to achieve and maintain the safe shutdown of the reactor core.
The RCS soluble boron content is not expected to change rapidly, if at all, following the initial borating during the SI phase of an accident.
Periodic analysis of RCS samples would detect any significant changes in boron concentration.
Instrumentation to continuously monitor RCS soluble boron concentration is not required since periodic analysis of RCS grab samples is adequate for verification of reactivity control.
This is a
deviation from the recommendations of Reg.
Guide 1.97 Rev.
3 for providing continuous RCS boron concentration indication (Deviation No. DV Also see footnote (bb)).
(b) Redundancy not required per NUREG 0737 requirements.
(c) Seismic qualification not required per NUREG 0737 requirements.
(d) All equipment when required to be used, is located in a mild environment, therefore environmental qualification is not required.
(e) Lack of lower containment spray flow monitoring instrumentation will not deter the operator's ability to determine adverse containment conditions.
adverse containment conditions can be monitored by looking at items such as containment pressure.
If we see containment pressure conditions different than what is
- expected, then we can confirm whether adverse containment conditions are due to a lack of spray flow by the monitoring of containment spray pump discharge pressure.
(f) DELETED (g)
DELETED (h)
Instrumentation to continuously monitor RCS radioactivity is not required.
See footnote (ff). Periodic analysis of RCS grab samples is adequate to detect deterioration of fuel cladding.
Indicative of an inadequate core cooling (ICC)
- event, fast deterioration of fuel cladding could be detected by sensing the ICC conditions through diverse instrumentation (i.e., RVLIS, CET's, TSAT meter).
(i)
Per SER issued June 16, 1981 concerning Auxiliary Feedwater System reliability, it is only required that qualified, S/G narrow range level indication be provided and backed up by qualified Auxiliary Feedwater Flow Indication.
S/G wide range level indication is not required to be environmentally qualified or powered from a Class IE power source.
ATTACHMENT TO AEP:NRC:0773AB Page 51 (j)
DELETED (k)
DELETED (l)
INTENTIONALLYLEFT BLANK (m)
DELETED (n)
DELETED (o)
INTENTIONALLYLEFT BLANK (p)
DELETED (q)
Not required per NUREG 0737 II.B.3 (r)
These valves are left normally open (safe position), the breakers racked out and they cannot change position.
Therefore, Environmental Qualification of position indication is not required.
(s)
Redundancy can be provided by VCR-101 thru 107 which are located inside containment.
VCR-201 thru 207 are located outside containment.
(t)
Our original submittal identified both the centrifugal charging pump (CCP) flow and CCP motor breaker status as type A variables.
This would require Category 1 instrumentation for monitoring these parameters.
The CCP breaker status indication will meet the Regulatory Guide 1.97, Rev.
3 recommendations for Category 1 instrumentation, except as noted in the Table under item No. A.28.
With regard to the CCP flow indication, it should be noted that our Emergency Operating Procedures require manual operator action based on indication of pump operation or flow.
The CCP breaker status indication and other parameters serve to verify pump operation.
The non Category 1
CCP flow indication can serve as a backup.
The above is a basis for our deviation from the Regulatory Guide 1.97, Rev 3
recommendation to provide Category 1 instrumentation for CCP flow indication.
(This deviation was discussed in AEP:NRC:0773S,, Item 2).
(u)
Transmitters to be replaced to improve accuracy.
See AEP:NRC:0836T dated July 16, 1987.
(v)
Credit was not originally taken for the seismic qualification of equipment as noted in AEP:NRC:07730.
We will take credit for seismic qualification of this equipment which has been established.
This response was also given in AEP:NRC:0773S Attachment 1, Item 3.3.7.
(w)
Monitoring of containment pressure is currently provided by two Category 3 wide-range
(-5 to 36 psig instruments) and four Category 3 narrow-range
(-5 to 12 psig) instruments.
The design pressure of the Cook Nuclear Plant containments is 12 psig.
The four narrow-range instruments are scheduled to be upgraded to meet Category 1 requirements by the end of the 1987 refueling outage for Unit 1
(which has been completed) and by the end of the 1988 refueling outage for Unit 2.
I i
ATTACHMENT TO AEP:NRC:0773AB Page 52 The wide-range containment pressure instrumentation ranges were revised to meet the requirements of NUREG-0578 and NUREG-0737.
These instruments are not powered by a emergency standby power source as recommended by Regulatory Guide 1.97, Rev.
3 for Category 1 instrumentation, and they do not meet the Category 1 separation criteria.
The wide-range instrumentation is, however, highly reliable, and as a result we believe it is unlikely that it would not be available if needed to monitor the course of an accident.
Further, it is our belief that for other than short-term individual compartment pressure
- peaks, the narrow-range instrumentation would span the range of pressure anticipated in our evaluation of loss-of-coolant-type'ccidents.
On the above
- basis, this is a deviation from the Regulatory Guide 1.97, Rev.
3 recommendation to provide Category 1
wide-range containment pressure instrumentation (Deviation No.
DV-6).
See response to AEP:NRC:0773S, Item 4.
(x)
The saturation meter equipment was originally installed in accordance with the requirements of NUREG-0578.
In an SER dated March 20,
- 1980, the equipment installed to monitor degrees of subcooling was found to be acceptable (NRC letter, A. Schwencer to John E. Dolan, dated March 20, 1980).
As noted in that correspondence, the device installed was a discrete digital
- monitor, and the plant process computer was used in conjunction with this monitor to, provide subcooling margin.
Additionally, as part of the NUREG-0737 Supplement 1 requirements, a subcooling margin curve is provided by our Technical Support Center computer.
We believe that these three instrument
- systems, which have been installed to be consistent with the requirements of their appropriate documents, are reliable.
As a result, we believe 't is unlikely that they would not be available if needed to monitor the course of an accident.
It should be noted that our original submittal providing status of Regulatory Guide 1.97, Rev.
3 compliance (AEP:NRC:07730, dated October 15, 1985) was intended to identify instrumentation currently in compliance with the Regulatory Guide; instrumentation not in compliance for which upgrading to the Regulatory Guide recommendations was planned; and instrumentation not in compliance for which justification for-a deviation from the Regulatory Guide recommendations was provided.
The instrumentation for monitoring degrees of subcooling falls into the latter category.
Based on the above information and justification, we again submit a
deviation from the Regulatory Guide 1.97, Rev.
3 recommendation to provide Category 1 instrumentation for monitoring degrees of subcooling (Deviation No. DV-7).
See AEP:NRC:0773S Attachment 1, Item 3.3.3.
(y)
For similar reasons to those stated in (t) above, we submit a deviation from the Regulatory Guide 1.97, Rev.
3 recommendation to provide Category 1
instrumentation for safety injection (SI) pump flow indication (Deviation No.
DV-22).
The SI pump motor breaker status instrumentation will meet Category 1 requirements except as noted in the Table under Item A.29.
This deviation was detailed in AEP:NRC:0773S Attachment 3, Item 3.
(z)
Our original submittal identified the steam generator blow down radia-tion indication as a
Type A variable.
- However, because of changes in our Emergency Operating Procedures made subsequent to our original submittal,
ATTACHMENT TO AEP:NRC:0773AB Page 53 manual operator action is no longer based on this variable.
We therefore request that steam generator blowdown radiation indication be deleted from our original list of Type A variables.
This response was given in AEP:NRC:0773S Attachment 3, Item 6.
(aa)
We will provide Neutron Flux Monitoring to comply with Reg Guide 1.97 Rev 3 recommendations, except as noted in the Table.
(bb) Primary coolant boron concentration can be measured in a range of 375 ppm to 10,000 ppm.
This range is based on PASS reactor coolant samples with a
1:1000 dilution.
The undiluted reactor coolant grab sample will be measured in the
.375 to 10 ppm range.
PASS would be used during and following loss-of-coolant accidents.
In the event of a
- LOCA, emergency boration and injection from the refueling water storage tank would occur and we would therefore expect a
reactor coolant boron concentration substantially in excess of the low range of our PASS sample measurement capability.
On the basis of the
- above, we submit a deviation from the Regulatory Guide 1.97, Rev.
3 recommendation to provide the capability to measure boron concentration in PASS samples to the lower limit of 0 ppm (Deviation No. 8).
This deviation was also detailed in AEP:NRC:0773S Attachment 1, Item 3.3.2.
(cc) Our original submittal indicated that we would replace the cold and hot leg RCS water temperature recorders with Category 1 instruments by the end of the 1987 refueling outages for Units 1
and 2.
- However, a
more detailed review of our current design has resulted in the identification of additional work (e.g., control room equipment and cable relocation, and installation of new emergency standby power sources) that needs to be performed beyond that identified at the time of our original submittal.
We therefore request that the completion dates for upgrading the recorders to meet Category 1 require-ments be changed to the 1989 refueling outage for Unit 1 and 1990 refueling outage for Unit 2.
This was noted in AEP:NRC:0773S Attachment 3, Item 5.
We have also since determined that only two hot leg and two cold leg channels are necessary to comply with the Reg.
Guide 1.97 recommendations and therefore we are only taking credit for the instruments noted in the table.
This item is pending NRC approval of an Appendix R change request.
(dd) Because of changes to the EOP's from the original
- October, 1985 submittal and planned changes to this instrumentation, we have reclassified this instrumentation as Category 2 per Reg.
Guide 1.97 guidance.
(ee)
We will upgrade the valve position limit switches on valves VCR-11 and VCR-21 to meet the environmental qualification requirements of 10 CFR 50.49 and Regulatory Guide 1.97, Rev.
3 recommendations.
The valve position limit switches for valve QCM-250, as 'well as the associated cable and terminations, are qualified in accordance with 10 CFR 50.49(k) except that they have not been qualified for submergence.
The QCM-250 position indication limit switch is located below maximum flood level.
Although this is not completely consistent with the Regulatory Guide 1.97, Rev.
3 recommendations for equipment qualification, we do not believe any upgrading of the position indication limit switch is necessary.
>1 t
ATTACHMENT TO AEP:NRC:0773AB Page This is due to the fact that QCM-250 is designed to close within 15 seconds of a containment isolation signal, which means that the valve will not become submerged before it performs its safety function.
In addition, once the valve is closed, it is extremely unlikely that it would change position due to its submergence.
Given these considerations, we believe that QCM-250 in its present
- status, without upgrading, adequately meets the intent of Regulatory Guide 1.97, Rev.
3 recommendations for achieving verifiable containment isolation.
We therefore submit a deviation from the recommendation of Reg.
Guide 1.97 Rev.
3 with respect to environment qualification for QCM-250.
(Deviation No.
DV-24).
The planned schedule for upgrading VCR-11 and VCR-21 to meet 10 CFR 50.49 requirements calls for this work to be completed in both units by the end of the refueling outages presently scheduled for 1989 (Unit 1) and 1990 (Unit 2).
This information was provided in AEP:NRC:0773S Attachment 1, Item 3.3.4.
(ff) As stated in our original submittal (AEP:NRC:07730, dated October 15, 1985) the primary coolant system radioactivity is not continuously monitored by in-line instrumentation.
- Rather, periodic analysis of reactor coolant grab samples is provided to detect deterioration of fuel cladding.
Our post-accident sampling system provides a
diluted grab sample which is analyzed by the gamma spectrum analyzer.
See our response to footnote (gg) below for the range of our gamma spectrum analyzer.
On the basis of the sampling capability described in footnote (gg) we are submitting a deviation from the Regulatory Guide 1.97, Rev.
3 recommendation for continuous monitoring of radioactivity in the reactor coolant system (Deviation No.
DV-9). It should also be noted that Category 1 requirements for this system are only to be applicable to equipment that operates equipment installed in the portion of piping that is Seismic Class I.
Electrical equipment operating equipment installed in Seismic Class 3 piping is to meet Category 3
requirements.
This information was provided in AEP:NRC:0773S Attachment 1,
Item 3.3.5.
(gg)
We believe the EG&G evaluation should have cited a range of 1 uCi/ml to 10 Ci/ml for this variable as per Table 3 of Regulatory Guide 1.97, Rev.
3.
Our range of measurement for gamma spectrum analysis of the diluted post-accident system grab samples of primary coolant is 1 uCi/ml to 10 Ci/ml.
This complies with the Regulatory Guide 1.97, Rev.
3 recommended range.
(hh) The instrumentation identified for this parameter in our original submittal was incorrect.
We initially identified our lower containment normal process radiation monitors (ERS-1300 and 1400 for Unit 1 and ERS-2300 and 2400 for Unit 2) for this parameter.
The instruments used to monitor this parameter are the unit vent radiation monitors (VRS-1500 for Unit 1 and VRS-2500 for Unit 2)..
The display location for these instruments is the control room CT-1 control terminal, not panel WDG.
All other information contained in our original submittal for this item remains the same.
This information was provided in AEP:NRC:0773S Attachment 3, Item 7.
'l L>
xii "p s 4g
ATTACHMENT TO AEP:NRC:0773AB Page 55 (ii) We incorrectly identified the instruments as ITR-311 and 321.
These are actually the RHR heat exchanger inlet temperature devices.
The outlet instrumentation tag numbers are ITI-310 and 320.
These devices have a range of 0-400'F and indicate locally and in the Technical Support Center computer terminal located in the Technical Support Center and the control room.
We believe this instrumentation meets the Regulatory Guide 1.97 recommendations for RHR heat exchanger outlet temperature measurement range.
This information was provided in AEP:NRC:0773S Attachment 1, Item 3.3.10.
(jj) The correct range for IFI-310 and 320 is 1500-5000 GPM.
We inadvert-ently reported the incorrect range to you in our previous submittal AEP:NRC:
07730.
We believe that this still complies with Reg.
Guide 1.97 Rev.
3 requirements for range for this variable.
(kk) To clarify our initial
- response, these motor-operated valves are normally left in the open position when the plant is operating in Mode 1 and Mode 2.
The circuit breakers are racked out and the valves
- cannot, there-
- fore, spuriously change position.
They can change position only as the result of deliberate operator action.
This information was provided in AEP:NRC:0773S Attachment 1, Item 3.3.12.
(ll) We presently have instrumentation installed that we believe meets Regulatory Guide 1.97, Rev.
3 recommendations.
Pressurizer heater current can be monitored by observing ammeters located on the pressurizer control panel-in the control room.
The range is 0-200 amps.
The pressurizer heaters are powered from the safety buses and therefore have the capability of being powered by the emergency power sources.
Automatic shedding of the pressurizer heaters following a blackout is provided to prevent overloading of the emergency power sources.
The operator can afterwards, at his discretion, manually energize the pressurizer
- heaters, taking care not to overload the emergency power sources.
This information was provided in AEP:NRC:0773S Attachment 1, Item 3.3.14.
(mm)
We do not rely on the quench tank to perform any post-pressurizer release function.
- However, we are providing the following information in response to the EG&G evaluation.
The range of 74% of total tank volume originally submitted was not accurately stated to show the adequacy of the existing installation.
The correct range should have been stated as being from 7 inches above the tank bottom to 7 inches below the tank top.
This range includes coverage of the sparger.
With regard to the ability to quench a "design-basis" pressurizer
- release, as noted above we do not rely on the quench tank to perform this function.
The quench tank is used during normal plant operation to contain pressurizer releases from routine pressurizer pressure adjustments and valve leakage.
In the case of a design-basis event that causes the PORVs and safety relief valves to lift, two rupture disks will burst before reaching the quench tank design pressure of 100 psi.
Subsequently discharge through the quench tank into the containment sump will occur.
With regard to overpressurization, we do not understand the basis for the EG&G position that sufficient gas volume exists to accept pressurizer release without becoming overpressurized.
As noted above, overpressurization will not
- occur, because rupture discs will burst and discharge into the containment before reaching the tank design pressure of 100 psig.
g,
ATTACHMENT TO AEP:NRC:0773AB Page 56
\\
Normal water level is kept at between 80X and 84X of the instrument range with a high alarm at 84X and a low alarm at 79X.
As such, in-leakage from the relief discharge system can be adequately monitored.
We therefore submit a deviation from the Regulatory Guide 1.97 recommendation to monitor quench tank level from top to bottom of tank (Deviation No. DV-10).
This information was provided in AEP:NRC:0773S Attachment 1, Item 3.3.15.
(oo)
On August 21, 1981 we submitted a letter (AEP:NRC:0300G) that documented discussions with NRR staff clarifying certain portions of an NRC SER (June 16, 1981) of the Cook Nuclear Plant auxiliary feedwater syst'm.
In that letter it was confirmed that Regulatory Guide 1.97 recommendations for steam generator level instrumentation did not have to be implemented at that time, but that implementation would be addressed at some time in the future through the Regulatory Guide 1.97 compliance/commitment process.
The steam generator wide-range level indication is not required for post-accident monitoring and in fact has been deleted from our Technical Specifications on Units 1
and 2.
This was stated in our December 10, 1980
- letter, which submitted a proposed amendment to our Technical Specifications (AEP:NRC:0449).
As stated in that letter, the reasons for deletion of steam generator wide-range level indication from the Technical Specifications are:
(1) the S/G vide-range level indication does not perform any safety-related function and is not assumed operable in the various plant safety analyses; and (2) the S/G narrow-range instrumentation, which we believe fulfills post-accident monitoring requirements, is environmentally and seismically, qualified, powered from a Class 1E source and has three redundant channels per S/G.
The S/G level indication is backed up by auxiliary feedwater flow instrumentation.
The S/G wide-range level instrumentation is powered from a
Class lE
- source, but all four channels are powered by the same source.
Since this is not in compliance with the Regulatory Guide 1.97, Rev.
3 recommendation, and based on the information given above, we are submitting a deviation from the Regulatory Guide 1.97, Rev.
3 recommendation for S/G level instrumentation.
(Deviation No.
DV-11)
This information was provided in AEP:NRC:0773S Attachment 1, Item 3.3.16.
(pp)
We incorrectly identified the range of instr~entation measuring this variable as 0-5x10 PPH.
It should read 0-4x10 PPH.
We believe that this still complies with Reg.
Guide 1.97 Rev.
3 requirements for range for this variable.
(qq)
I6MECo currently provides condensate storage tank (CST) level indication in the control room through three highly reliable Category 3 level-measuring devices.
One of these instruments is electrically operated, while the other two are pneumatic devices.
In addition, CST level can be read at the local turbine-driven auxiliary feedwater pump control panel.
I&MECo has also committed to provide additional CST level indication by adding an instrument channel to meet Category 1
requirements.
The new channel was installed during the 1987 refueling outage on Unit 1 and the 1988 refueling outage on Unit 2.
M gt Ci
>>M
ATTACHMENT TO AEP:NRC:0773AB Page 56 Normal water level is kept at between 80%
and 84X of the instrument range with a high alarm at 84% and a low alarm at 79X.
As such, in-leakage from the relief discharge system can be adequately monitored.
We therefore submit a deviation from the Regulatory Guide 1.97 recommendation to monitor quench tank level from top to bottom of tank (Deviation No. DV-10).
This information was provided in AEP:NRC:0773S Attachment 1, Item 3.3.15.
(oo)
On August 21, 1981 we submitted a letter (AEP:NRC:0300G) that documented discussions with NRR staff clarifying certain portions of an NRC SER (June 16, 1981) of the Cook Nuclear Plant auxiliary feedwater system.
In that letter it was confirmed that Regulatory Guide 1.97 recommendations for steam generator level instrumentation did not have to be implemented at that time, but that implementation would be addressed at some time in the future through the Regulatory Guide 1.97 compliance/commitment process.
The steam generator wide-range level indication is not required for post-accident monitoring and in fact has been deleted from our Technical Specifications on Units 1
and 2.
This was stated in our December 10, 1980 letter, which submitted a proposed amendment to our Technical Specifications (AEP:NRC:0449).
As stated in that letter, the reasons for deletion of steam generator vide-range level indication from the Technical Specifications are:
(1) the S/G wide-range level indication does not perform any safety-related function and is not assumed operable in the various plant safety analyses; and (2) the S/G narrow-range instrumentation, which we believe fulfills post-accident monitoring requirements, is environmentally and seismically qualified, powered from a Class 1E source and has three redundant channels per S/G.
The S/G level indication is backed up by auxiliary feedwater flow instrumentation.
The S/G vide-range level instrumentation is powered from a
Class 1E
- source, but all four channels are powered by the same source.
Since this is not in compliance with the Regulatory Guide 1.97, Rev.
3 recommendation, and based on the information given above, we are submitting a deviation from the Regulatory Guide 1.97, Rev.
3 recommendation for S/G level instr'umentation.
(Deviation No.
DV-11)
This information was provided in AEP:NRC:0773S Attachment 1, Item 3.3.16.
(pp)
We incorrectly identified the range of instrgmentation measuring this variable as 0-Sx10 PPH.
It should read 0-4x10 PPH.
We believe that this still complies with Reg.
Guide 1.97 Rev.
3 requirements for range for this variable.
(qq)
I&M currently provides condensate storage tank (CST) level indication in the control room through three highly reliable Category 3 level-measuring devices.
One of these instruments is electrically operated, while the other two are pneumatic devices.
In addition, CST level can be read at the local turbine-driven auxiliary feedwater pump control panel.
I&M has also committed to provide additional CST level indication by adding an instrument channel to meet Category 1
requirements.
The new channel was installed during the 1987 refueling outage on Unit 1
and the 1988 refueling outage on Unit 2.
l' WJ
ATTACHMENT TO AEP:NRC:0773AB Page 57 The CST is the initial source of water for the auxiliary feedwater (AFW)
- system, and provides sufficient volume to maintain the reactor coolant system in a hot standby condition for 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />.
In the event that sufficient water is not available from the CST in one unit, operating procedures call for a cross-tie valve to be opened to supply feedwater from the CST in the other unit.
In the unlikely event that neither CST can supply sufficient
- AFW, procedures require transferring the supply source to the essential service water system (ESWS).
The water supply for the ESWS is Lake Michigan.
In view of the number and diversity of instrumentation available to provide CST level monitoring, and the ultimate availability of Lake Michigan as a source of auxiliary feedwater, we are submitting a deviation from the Regulatory Guide 1.97 recommendation to provide more than one Category 1
level indication for the CST (Deviation No.
DV-12).
This information was provided in AEP:NRC:0773S Attachment 3, Item 1.
(rr) When operating
- normally, each containment spray pump will deliver 3200 gpm (design flow) at 490 ft. TDH.
We have attached to this response a curve showing containment spray pump flow as a function of pump discharge pressure (drawing No. HXP87055JW-1).
See Attachment No. 4.
The attached curve indicates the expected range of operation for the containment spray pumps.
This operating range stems from consideration of pump suction head, containment
- pressure, and pump operating characteristics.
Routine surveillance of spray pump operation is performed to ensure that, if containment spray is required, the pumps will operate in the indicated area of the flow curve and hence provide the necessary flow to the containment spray system.
The reactor operators
- can, therefore, verify proper containment spray flow by monitoring spray pump discharge pressure to confirm that it is within the expected range.
It should be noted that the upper containment spray flow instrumentation cited in our original submittal (AEP:NRC:07730, dated October 15, 1985
[IFI-330 and 331]) measures only the flow provided by the RHR pumps to the upper containment
- spray, not the flow from the containment spray pumps.
However, the containment spray pumps, not the RHR pumps, are normally used to supply containment spray flow.
- Also, please note that the flow range of 0-200 gpm for IFI-330 and 331 (for measurement of RHR pump flow to the upper containment spray) given in that submittal is incorrect.
The correct range is 0-2500 gpm.
Based on the above, we are noting a deviation from the Regulatory Guide 1.97, Rev.
3 recommendation for containment spray flow instrumentation (Deviation No.
DV-13).
This information was provided in AEP:NRC:0773S Attachment 1, Item 3.3.17.
(ss)
The RHR heat exchanger inlet temperature instrumentation will be upgraded to meet Regulatory Guide 1.97 Rev.
3 recommendations except as noted in the tables.
This response was given in AEP:NRC:0773S Attachment 1, Item 3.3.18.
4
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ATTACHMENT TO AEP:NRC:0773AB Page 58 (tt) Because of the following actions which apply for normal, accident, and post-accident conditions, we believe level indication beyond that currently provided is. not required.
Upon receiving a hi-level alarm, flow into the Volume Control Tank (VCT) is automatically fully diverted into the hold-up tanks.
If.",:a low-level alarm is reached, an alarm alerts the operator to restore level.
In the event this effort fails, an emergency lo-lo level alarm is sounded and the refueling water sequence is automatically initiated.
We believe that this range (0-70 inches) is adequate to safely monitor the operation of this tank.
In the unlikely event that VCT level indication is lost and the VCT becomes completely full, a safety relief valve (set at 75 psig) will open and the excess water will be discharged into the hold-up tanks.
We therefore are submitting a deviation from the Regulatory Guide 1.97 recommendations to monitor Volume Control Tank Level from top to bottom (Deviation No.
DV-14).
This information was provided in AEP:NRC:0773S Attachment 1, Item 3.3.19.
(uu) An analysis previously performed for one of the radiation exposure rate monitors (VRC-301) listed in our original submittal (AEP:NRC:0(720, dated Oc)ober 15, 1985) showed that the exposure rate range of 10 mR/hr to 10 mR/hr was adequate to monitor plant operation in the area in which this monitor is installed.
As part of a general upgrade of area radiation monitors at the Cook Nuclear
- Plant, monitor numbers NRA-340 and RRA-332 will be replaced.
As previously noted in our submittal AEP:NRC:0773S, Item 3.3.9, concurrent with the upgrade activities, analyses of the type mentioned above was to be performed to determine what range of exposure rate measurement is appropriate for the monitors in the areas where they are to be installed.
We expected that the analysis would show that a
range less than that recommended by Regulatory Guide 1.97, Rev.
3 would be adequate to safely monitor plant operations in the areas where these monitors are to be installed.
This analysis has been completed and the results are explained in footnote (kkk) herein.
The upgrading program is scheduled f'r completion at both Unit 1 and Unit 2 by the end of 1989.
(VV) INTENTIONALLYLEFT BLANK (ww) The range for SFR-401 was incorrectly identified as 0-2000 SCFM.
The correct range for measurement of this variable is 0-250 SCFM.
We believe that this still complies with the Reg.
Guide 1.97 Rev.
3 requirements for range for this variable.
(xx) This instrumentation was recently (1985) upgraded. by the addition of a high-range noble gas detector.
Based on our recent primary calibratioy
- analysis, the range gf this instrumentation was determined to be 5.8 x 10 uCi/cc to 1.86 x 10 uCi/cc Xenon-133 dose equivalent.
On July 23, 1986 a
letter was sent to the NRC (AEP:NRC:0678Y) in which we stated that post-accident conditions would not result in steam jet air ejector exhaust noble gas concentration greater than 2
x 10 uCi/cc.
On this basis we requested an exeqytion from the NUREG-0737,Section II.F.l-l upper-rangy requirement of 10 uCi/cc in favor of a more realistic upper range of 10 uCi/cc.
We therefore are submitting a deviation from the Regulatory Guide 1.97, Rev.
3 recommendation for this parameter (Deviation No. DV-15).
I
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ATTACHMENT TO AEP:NRC:0773AB Page 59 It should also be noted that the tag numbers SFR-1900 and SFR-2900 given in our original submittal are incorrect.
The correct tag numbers are SRA-1900 and SRA-2900.
This information was provided in AEP:NRC:0773S Attachment 1, Item 3.3.21.
(yy) The range of 3 uCi/cc to 20 x 10 uCi/cc as provided in our submittal 5
was based solely on the monitor's response to Xe-133 and not to the anticipated mixtuie of radioisotopes following a
steam generator tube rupture.
The lower limit of O.l, uCi/cc of Xe-133 equivalent mixture can be measured.
As stated in our September 8,
1986 letter (AEP:NRC:0678Z),
when the anticipated mixture of radioisotopes for a steam generator tube rupture is used, the maximum concentration is calculated to be 0.263 uCi/cc Xe-133 equivalent activity.
With respect to this upper range limit, an exemption from the NUREG-0737 requirement of 1000 uCi/cc was requested in the September 8,
1986 letter and a
100 uCi/cc value proposed.
No response to our request has been received at this writing.
We are submitting the same upper limit deviation from the Regulatory Guide 1.97, Rev.
3 guidelines (Deviation No.
DV-16).
This information was provided in AEP:NRC:0773S Attachment 1,
Item 3.3.22.
(zz) It was planned to have available by the end of 1988 a portable gamma-ray spectroscopy system providing the capab'ility for field analysis of plant and environs radioactivity.
This equipment was shipped to the Donald C.
Cook Nuclear Plant on July 30, 1987 and is now available for use.
Information related to this variable was provided in AEP:NRC:0773S Attachment 1,
Item 3.3.23
'aaa) The capability to measure gross activity in the range of 1 uCi/ml - 10 Ci/ml is available; however, this measurement is not normally used to assess core damage.
- Rather, our initial core damage assessment is done through gamma spectrum analysis of primary coolant.
This method provides an isotopic analysis as well as giving an indication of total primary coolant activity.
This information was provided in AEP:NRC:0773S Attachment 1,
Item 3.3;24 No. l.
(bbb)
Gamma spectrum isotopic analysis is performed in an energy range of 0.050-2.05 MeV.
This information was provided in AEP:NRC:0773S Attachment 1,
Item 3.3.24 No. 2.
(ccc)
Boron content is measured in the range of 375-10,000 ppm (see footnote (bb))
This information was provided in AEP:NRC:0773S Attachment 1,
Item 3.3.24 No. 3.
(ddd)
Chloride content in undiluted samples 30 days after an accident is measured in a range of 0.01 to 20 ppm.
For diluted samples 1:1000 taken within 4 days of an accident, the range of measurement is 10 to 20,000 ppm.
We are noting a deviation from the Regulatory Guide 1.97, Rev.
3 lower limit of 0
ppm (Deviation No.
DV-17).
This information was provided in AEP:NRC:0773S Attachment 1, Item 3.3.24 No. 4.
(eee)
Our range of measurement of pH is 1 to 13.
Our original submittal showed a range of 5 to 8 which we believed to be the range of interest for this parameter.
This information was provided in AEP:NRC:0773S Attachment 1,
No. 5.
II I L
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ATTACHMENT TO AEP:NRC:0773AB Page 60 (fff)
An exemption from the requirement for taking hydrogen grab samples of containment air was granted via letter from Youngblood (NRC) to Dolan (AEP) dated November 5,
1986.
Therefore, we are also submitting a
similar deviation with respect to Reg.
Guide 1.97 Rev.
3 (Deviation No. DV-18).
We do,
- however, perform continuous monitoring of containment air hydrogen content in the range of 0
to 30 volume percent (see Item C-10)
This information was provided in AEP:NRC:0773S Item 3.3.24 No. 6.
(ggg)
NUREG-0737 does not require sampling of containment air oxygen content.
As noted
- above, however, we do continuously monitor hydrogen content which mades containment air oxygen content of less concern from the standpoint of potential hydrogen flamability or deflagration.
We therefore are submitting a
deviation from the Regulatory Guide 1.97, Rev.
3 recommendation to sample for containment air oxygen content (Deviation No.
DV-19).
This information was provided in AEP:NRC:0773S Item 3.3.24 No. 7.
(hhh)
We do not understand the EGGG request to provide containment'ir gamma spectrum "capacity."
Regulatory Guide 1.97, Rev.
3 recommends that the capability be provided to perform an isotopic analysis of containment air.
As recommended by Regulatory Guide 1.97, Rev.
3, gamma spectroscopy techniques are used to provide an isotopic analysis of noble gases in containment air.
This gamma spectrum isotopic analysis is performed in a
energy range of 0.050 to 2.05 MeV using a Canberra series 85 multichannel analyzer with a Digital PDP 11/24 computer and either a germanium (lithium-drifted) (Ge[Li]) or high-purity germanium (HPGe) detector.
This information was provided in AEP:NRC:0773S Attachment 1, Item 3.3.24 No. 8.
(iii) Actual measured range was noted incorrectly in our previous submittal.
Per our response in AEP:NRC:0773S Attachment 1,
Item 3.3.11 we have made a
minor correction to the actual measured range value.
We believe that we still comply with the requirements for range for this variable.
(jjj)
Circuits which require the use of indicating lamps for position indications, status indication, etc. will be using existing General Electric ET16 Indicating Lamps for this function.
We have been advised by the manufacturer, that these indicating lamps meet their (the manufacturer's) interpretation of IEEE-344-1975.
This indicating lamp is a
seismically rugged commercial grade device for which comprehensive qualification is not available.
Since these lamps are purchased as standard commercial grade material and are not manufactured for a specific order, 10CFR21 can not be applied to these devices.
Therefore, we are submitting a deviation from Reg.
Guide 1.97 for this device so we may continue to use and purchase it and its parts for use in monitoring Post Accident conditions (Deviation No. DV-20).
(kkk)
ERA-7303 thru 7308, and ERA-8303 thru 8308 have ranges of 0.01 to 1000 R/HR.
ERS-7401,
- 7403, 7404,
- 8401, ERA-7507,
- 7601, 7603, 7605 have ranges of
.0001 to 10 R/HR.
ERA-8403,
- 7504, 7508,
- 7602, 7604 have ranges of 0.001 tg
-1 10 R/HR.
These are different than the recommended range of 10 to 10 R/HR and we therefore are submitting a deviation for this variable in regards to range (Deviation No.
DV-23).
The justification for this request is as follows.
With the exception of ERA-7305,
- 7306, 8305, and
- 8306, the worst case maximum estimated accident dose rate is less than the upper range limit noted above.
The lower upper range limit is used to provide more accurate, useful information and to help prevent false "low fail" alarms.
'I Ab
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ATTACHMENT TO AEP:NRC:0773AB Page 61 In the case of ERA-7305,
- 7306, 8305,
- 8306, the worst case maximum estimated accident dose rate is 1730 R/HR which exceeds the upper range limit of 1000 R/HR.
However, within one (1) hour, this drops to 573 R/HR which is well within the upper range limit.
We believe that because personnel entry in an area where-'exposure may exceed 1000 R/HR (indicated by a "high fail" status indication) is highly unlikely and because the dose rate will quickly fall below the upper range limit of 1000 R/HR (at which time a quantitative indication will again be available) the range of 0.01 to 1000 R/HR is adequate.
Again, using this range will provide more useful information and help prevent false "low fail" alarms.
(ill) If CCW water temperature is not available, adequate CCW cooling can be verified by monitoring CCW flow and RHR Inlet
& Outlet temperatures, all of which are qualified (or planned to be qualified) for the intended purpose.
Therefore, because of the availability of suitable diverse indications, environmental qualification of instrumentation monitoring this variable is not required.
We are submitting a deviation with respect to Reg.
Guide 1.97, Rev.
3 recommendations for this variable for environmental qualification, (Deviation No. DV-25).
(mmm)
All equipment when required to be
- used, is located in a
mild environment except for cables serving the following instruments.
For these instruments the cables pass through harsh environment areas.
Equipment qualification is not required except for these cables.
The instruments served by these cables are:
2-IFI-310, 2-IFI-311, 2-IFI-320, 2-IFI-321, 2-CFI-419, 2-CLI-114, and 1-CLI-114.
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