ML17331A571
| ML17331A571 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 01/08/1981 |
| From: | Hunter R INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.1.1, TASK-1.C.5, TASK-1.C.6, TASK-1.D.1, TASK-2.K.2.13, TASK-TM AEP:NRC:0398, AEP:NRC:398, BEP:NRC:0398, BEP:NRC:398, NUDOCS 8101130217 | |
| Download: ML17331A571 (36) | |
Text
REGULATOP I>>FOR~ATION 0 ISTRI BUT ION TFl~'RIBS)
ACCESSI04 NOR:8101130217 DOC ~ DATE: Ri/01/08 NOTARIZED:
YFS FACIL:50 315 Donald C ~
Cook Nuclear Power Plantr Unit ii Indiana 5n 316 Donald CD CooK Nuclear power Plant> Unit 2i Indiana AUTH ~ NAHF
, AUTHOR AFFILIATION HUNTERrR,Se Indiana 8 >ichiaan Electric Co.
REC IP. >>A>E RFCIP IE>>T AFF ILIATION DENTONpH.RE Office of Nuclear Reactor Reaulationr Director
SUBJECT:
Forwards response to NRC 801031 1 tr re implementation of nos t-i~>T ID CODE/NA~E ACTION: -
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INDIANA IIt MICHIGAN ELECTRIC COMPANY P. O.
BOX 18 BOWLING GREEN STATION NEW YORK, N. Y. 10004 January 8,
1981 AEP:NRC:0398 Donald C.
Cook Nuclear Plant Unit Nos.
1 and 2
Docket Nos.
50-315 and 50-316 License Nos.
DPR-58 and DPR-74 Post-TMI Requirements (NUREG-0737)
Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.
C.
20555
Dear Mr. Denton:
The attachments to this letter provide our response to Mr. D.
G. Eisenhut's letter of October 31, 1980, which we received on November 6, 1980, concern-ing the implementation of the post-TMI requirements contained in NUREG-0737.
Our responses follow the same format of Enclosure 1 of NUREG-0737.
For those items of Enclosure 1 which are implemented under NUREG-0578 and are noted
'as "COMPLETE", no response is provided.
We have proceeded with the implementation of many of the post-TMI requirements based on the criteri'a of NUREG-0578.
In some instances a pro-posed revision to Regulatory Guide 1.97 was referenced for the design and qualification criteria of certain equipment.
- However, NUREG-0737 now requires us to meet the provisions of its Appendix 8 for design and qualification of some of the same equipment.
Ne are told in the applicable items that this is to be considered a
new requirement.
We do not believe that the issuance of nevi requirements at this late date is compatible a priori with your imple-mentation schedules and with our established engineering practices.
Appendix 8
of NUREG-0737 represents a substantial change in the design and qualification of some equipment to the extent that it may impact on the completion of items already underway.
Where this is the case, it is pointed out in the attachments to this letter.
Even more, our assessment of the impact of the design and qualification criteria of Appendix 8 is still under review.
Any additional problem areas-will be noted i'n supplemental correspondence to this letter.
cc:
attached Very truly yours, J
'4 R.
S. Hunter Vice President
Nr. Harold R. Denton
,AEP:NRC:0398 cc:
R.
C. Callen G. Charnoff John'.
Dolan R,
W; Jurgensen D.
V. Shaller - HrMgman NRC Region III.'es%dent Inspector at Cook Plant Bridgman
STATE OF NEW YOR.K COUNTY OF NEW'ORK R.
S. Hunter, Being duly sworn, deposes and says that he is the I
Vice President of Licensee Indiana 5 Michigan Electric Company, that he has read the foregoing response to the post-TMI requirements contained in NUREG-0737 and recognizes the contents thereof; and that said contents are true to the best of his knowledge and belief.
Subscribed and sworn to before me this ~~'ay of ~m.
, 19~i
&~7 e
Cue';i,n s.bent Cnunty Ccrlincrs',u liteci in t! w 'fork County
(: rser i~) v(I nn re i/dsch 4'ttC/
ATTACHMENT NO.
1 TO AEP:NRC:00398
RESPONSE
TO iTB1 I'.A.l. 1 SHIFT TECHNICAL ADVISOR; The Shift Technical Advisor Training Program as described below is complete with dedicated personnel on shift as of January 1,
1981.
The Shift Technical Advisor (STA) Training Program was developed to provide the necessary training and background for both the accident assessment function and an operating experience assessment function.
The initial train-ing program for these functions consisted of 0he following subjects in addition to extensive systems training;
- Reactor Physics, Cnemistry and Materials
- Reactor Thermodynamics, Fluid iMechanics, and Heat Trans er
- Electrical Engineering including Reactor Control Theory All subjects were taught at the college level by a local university, NRC approved
- vendors, and the on-site Training Department staff.
Only personnel with prior degrees in engineering or physical sciences are selected
- and, as such, their prior training in mathematics was determined sufficient to meet the requirements of i4lr, Denton's October 30, 1979 letter.
To ensure the STAs familiarity with desion, function, arrange-
- ment, and operation of plant systems as required by the October 30, 1979 letter, only applicants with prior power plant experience will be admitted into the STA program.
In addition, the extensive system training given in plant systems, including design and operation is consistent with the INPO reconmendations for systems training.
Transient and accident response training will be given through special lectures by the NSSS vendor and simulator training.
This training is con-sistent with the requirements of the Oc ober 30, 1979 letter.
Attachment I.A-1 shows a comparison of the content of the established D. C.
Cook Plant STA Training Program to the INPO guidelines contained in Appendix C of NUREG-0737.
All major areas of concern have been covered.
In some areas, the INPO'recommended contact hours (Section 6 of Appendix C) cannot be met due to time and resource cons raints.
- However, we eel that the content of all areas of concern are adequately covered to meet the training qualifications as specified in the October 30, 1979 NRC letter.
The STA candidates will attend a simulator training program annually where they will participate in plant evolutions to gain experience in situa-tion assessment and the necessary actions to mitigate the consequences of an accident.
An STA Requalification Program is designed to maintain a
continuous'egree of knowledge and proficiency as required by ANS 3. 1 is estab-lished for the Donald C.
Cook Nuclear Plant Units 1 and 2.
This program shall apply to all STA's, including STA's who perform such duties on an infrequent basis.
A site appointed Training Coordinator has been assigned to implement and administer this program.
A brief description of the requalification program is contained in Attachment I.A-2.
RESPONSE
TO ITEM I.A.1.3, SHIFT'MANNING:
1.
Limit Overtime-Plant Manager Standing Order PMS0.054 dated October 29, 1980 was issued limiting overtime for Reactor Operators, Senior Reactor Operators and Shift Technical Supervisors.
2.
Shift Staffing-Implemented as described in our November 7, 1980 letter (AEP:NRC:00450) in response to Eisenhut's letter of July 31, 1980 concerning interim criteria for shift staffing.
RESPONSE
TO ITEM I.A.3.1, UALIFICATIONS OF REACTOR OPERATORS:
Implemen ed as described in our September ll, 1980 letter (AEP:NRC:00395)
~
in response to Oenton's letter of March 28, 1980 concerning reactor operator qualifications program.
RESPONSE
TO ITEM I.C.1, SHORT-TERM ACCIDENT AND PROCEDURES RE'lIEM:
The Westinghouse Owners Group (MOG) of which Indiana 8 Michigan Electric Company
( IKMECo) is a member will submit by January 1, 1981, a detailed description of the program to comply with the requirements of Item I.C. l.
In addition, we have revised our natural circulation cooldown procedure to limit the cooldown rate to prevent the formation of bubbles in the reactor head area.
This revision is a result of our review of IE Circular No. 80-15.
RESPONSE
TO ITEM I.C. 5, FEEDBACK( OF OPERATING EXPERIENCE:
In accordance with our letter of June 20, 1980 (AEP:NRC:00419), which respondedto Eisenhut's May 7, 1980 letter concerning additional TMI-2 requirements, procedures for feedback of operating experience to plant staff have been reviewed.
Existing, in place plant procedures were found to ade-quately address all requirements for NRC reports and notices and Cook Plant Condition Reports as stated in Item I.C.5, "Procedures for Feedback of Operating Experience to Plant Staff".
A formal new procedure which meets the requirements stated in Item I.C.5 has been implemented at the Plant for the handling of non-NRC Notices, such as NSSS Vendor Hotic s, and IHPO Notices.
Consistent with tfie requirements for an 'Operating Experience Assess-ment'roup as outlined in Mr. Denton's October 30, 1979 letter to all operating nuclear plants, procedures for handling HRC, vendor and industry-related notices vill be modified to directly involve the
'STA in o,ice duty'n an immediate safety assessment function.
If it is found that the-notice contains information of significant importance that should not, wait for emphasis through the usual routing, this group wi 11 act promptly to take aporopriate action.
All NRC, vendors, indus.ry and internal Cook Plant Notices will be reviewed by the
'STA in of ice duty'or trend analysis.
RESPONSE
TO ITEM I.C.6 VERIFY CORRECT PERFORMANCE OF OPERATING ACTIVITIES This requirement, formally issued by NUREG-0737, has been reviewed for its impact on Cook Plant operating activities.
Me currently have in place an effective system of verifying the correct performance of operating activities, the essential elements of which are described below.
Conducting surveillance on the Reac.or Protection and Engineered Safe-guards System during plant operation requires Shift Supervisor or Operating Engineer approval prior to start.
Plant surveillance testing is in accordance with approved schedules and procedures.
Oata are reviewed by the Unit Supervisor, Operating Engineer or Shift Operating Engineer.
The daily mast r schedule for completion of required surveillance is reviewed by the Operating Engineer or Shift Operating Engineer to determine current status.
Clearance Permits to remove essential equipment from service are prepared by an Operating Engineer or by the Shift Operating Engineers.
To insure that the ControlRoom is aware of equipment status, the permits are sent to the Unit Supervisor who directs their implementation.
Return to service is in the reverse order of the above with the inal acceptance for system operation being made by the Operating Engineer or Shift Operating Engineer.
System lineup of the safety-related system requires independent verification of valve lineup after it has been out of service for maintenance or following surveillance testing.
RESPONSE
TO ITEM E.D.1",
CONTROL ROOM DESIGN REIEER An assessment of the requirements for implementation of this item will'e per formed when NUREG-0700 is issued and reviewed.
RESPONSE
TO ITEM I.D.2, PLANT SAFETY PARAMETER DISPLAY CONSOLE:
Me are in the process of implementing the Mestinghouse designed Plant Safety Status Display (PSSD) system in the control rooms of the Cook Nuclear Plant.
On May 27, 1980, AEPSC representatives met with Messrs.
- Hanauer, Mattson and other NRC staff members and presented to them a description of ihe PSSD to be installed in Cook Plant.
This equipment has been ordered since May 9, 1980 and we are pro-ceeding anead with the work required to complete its'nstallation and operation.
The PSSD is further described in Hestinghouse's submittal of klCAP-9725 to the NRC dated June 13, 1980 (NS-TMA-2261).
This system is an integral part of'the Westing-house Technical Support Center Complex and is designed with the application of Human Engineering principles.
RESPONSE
TO ITEM II.B.l REACTOR COOLANT SYSTEM VENTS The documentation required for this item will be provided by July 1, 1981.
Installation of this system will be completed by July 1, 1982.
This schedule complies with the requirements of NUREG-0737 for this item.
RESPONSE
TO ITEM II.B.2 DESIGN REVIEW OF PLANT SHIELDING AND ENVIRONMENTAL OUALIFICATION OF E UIPMENT:
This item has been implemented by the responses provided in our letters dated March 10, 1980 (AEP:HRC:00334B) and May 15, 1980 (AEP:NRC:00334D) regarding item 2.1.6.b of NUREG-0578.
In addition, the specific equipment qualifications are provided in our submittals in response to. IE Bulletin 79-018, dated March 7, 1980 (AEP:NRC:00356),
vtay 7, 1980 (AEP:NRC:00356A), June 5,
1980 (AEP:NRC:003568) and October 31, 1980 (AEP:NRC:00356C).
RESPONSE
TO ITEM II.B.3 POST ACCIDENT SAMPLING:
Design modifications, sampling equipment and analytical capability will be in place to satisfy this requirement by January 1, 1982, the NUREG-0737 given date.
RESPONSE
TO ITEM II.B.4 TRAINIHG FOR'MITIGATING CORE DAMAGE Me plan to use the Westinghouse program entitled "Mitigating Core Damage Training'Course"'.
This program will be initiated around April 1, 1981 and com-pleted-by October 1, 1981 in accordance with the requirements of this item.
RESPONSE
TO ITEM II.O.l, PERFORMANCE'TESTING OF RELIEF ANO SAFETY VALVES NUREG-0578 Item 2.1. 2:
As a participating member of the EPRI PWR Safety and Relief 'lalve Test
- Program, Indiana 8 Michigan Electric Company is complying with !the requirements of HUREG-0578, I'tern 2.1.2.
By letter dated Oecember 15,
- 1980, R.
C.
Youngdahl of Consumers Power Company has =provided the current partici-pating utility positions with regard to the clarifications of item II.O.1 of HUREG-0737.
RESPONSE
TO ITEN II.O.3 OIRECT INOICATION OF RELIEF AHO SAFETY 'IAL'lE POSITION:
This item has been implemented, in the Cook Plant and our compliance is documented in the HRC Safety Evaluation Report issued on March 20, 1980.
The human fac.ors analysis recommended in clarification item 6 will be included in the detailed control room design review 'to be performed in accordance with our response to item I.O.1 of NUREG-0737.
RESPONSE
TO ITEM II.E.l.l. AUXILIARY FEEDWATER SYSTEM EVALUATION:
The information required by this item concerning the AFW system flow design basis nas been submitted by our letter dated November 3, 1980 (AEP:HRC:00300C).
The NRC has closed this item with the issuance of the October 6, 1980 Safety Evaluation Repor" of the Cook Plant's AF!3 system reliability.
RESPONSE
TO ITEiM II.E.1.2. AUXILIARY FEEOWATER SYSTEM AUTOMATIC INITIATION ANO FLO!A I!VOICATION:
Part 1:
The APX system is automatically started at the Cook Plant and the initiation signals and circuits meet safety grade requirements.
Part 2:
The flow indication.system to be used at D.
C.
Cook Plant will be one auxiliary feedwater flow rate indicator and one narrow range steam generator level indicator for each steam generator Installation of environmentally qualified auxiliary feedwater flow rate transmitters will be compatible with the NUREG-0737 implementation schedu'le.
The information previously submitted for part 1 of this item and our letter of Decem6er 10, 1980 (AEP:NRC:00307EJ in response to S.
A. Yarga's letter of October 31, 1980 shows compliance with the specified requirem nts of part 2 of this i,tern.
Also, our submit.al of the Category A Lessons Learned Technical Specifications by letter dated December 10, 1980 (AEP:NRC:00449) is consistent with the above implementation of AFM flow indication.
RESPONSE
TO ITEM II.E.4.2 CONTAINMENT ISOLATION DEPENDABILITY:
D.
C.
Cook Plant is already in compliance with HRC positions 1 through 4
as documented in the HRC March 20, 1980 SER.
The Cook Plant design basis for the minimum pressure setpoint of con-tainment isolation is in compliance with NRC position 5.
For D.
C.
Cook Plant each unit has two levels of containment isolation identified as Phase A and Phase B.. Phase A isolation closes all lines penetrating the containment except essential lines sucn as Safety Injection and Containment Spray which are not isolated, and component cooling water to the reac or pumps and service water to the ventilation units which isolates on Phase B.
Phase A isolation is initiated by containment pressure high (1.1 psig),
any safety injection signal and manually.
The D. C.
Cook Plant Technical Specification for the
'high limit of containment internal pressure has as its Limiting Condition of Operation (LCO), +0.3 psig.
Therefore, the differential between the LCO and the Phase A isolation setpoints from containment pressure high (0.8 psig) is within the requirements of clarification item (6) for a margin of 1 psi.
Thus, the 1.1 psig setpoint is the minimum compatible with normal operat-ina conditions and no further action is required.
The requirements contained in NRC positions 6 and 7 have already been-implemented as the Cook Plant.
These two positions have been addressed as indicated in Attachment II.E-l.
RESPONSE
TO ITEM II.F. I ATTACHMENT 1 NOBLE GAS MONITOR:
The upgraded radiation monitoring system at D. C.
Cook Plant is being designed to provide the information required by both Attachments 1 and 3 of Item II.F. 1.
This system which is on order and being manufactured has been purchased in accordance with your earlier requirements of HUREG-0578, prior to issuance of NUREG-0737, based on the ccnmitment in our October 24, 1979 letter (AEP:NRC:00253).
The design of the system to meet the criteria specified in Attachments 1 and 3 is interrelated to the extent that separation of the documentation would be extremely dif,icuIt..
Schedule.relief..of the January.
1, 1981 date is therefore requested to provide the documentation for Attachment 1
on the same schedule as that required by Attachment 3, i.e.
by July '1, 1981 rather than January 1, 1981.
The portions of our upgraded radiation-monitoring system for compliance with this item will be installed by January 1,
1982 con-sistent with the requirements of Attachment 1.
The uograded radiation monitoring system at Cook Plant will meet the requirements of the NRC positions of Attachment 1 when reviewed in total but not on an individual monitor bases.
The system will provide the ranges for the release path as stipulated by Table II.F,1-1 under "Design 8asis Minimum Range".
However, the system normal operating range monitors are not capable of this extended range nor are the post-accident monitors capable of the ALARA*range"'equi'rements.'e wish to point out that clarification 4 regarding ALARA range requirements on the post-accident monitors conflicts with the requirements of clarification 1 and Table II.F.l-l.
RESPONSE
TO ITEM II.F. 1 ATTACHMENT 2, SAMPLING AND ANALYSIS OF PLANT EFFLUENTS:
The upgraded radiation monitoring system at 0.
C.
Cook Plant is being designed to provide information as required by Attachment 2.
The proposed Hi range noble gas. monitors will be able to sample particulates and radioiodines by absorpiton on a filter, followed by an onsite laboratory analysis.
The absorption of radio-iodines will be done on a Silver Zeolite filter.
Me believe that the technology currently available is limited to low temperature gases with low moisture content.
Consequently,,steam vents are not monitored for Iodine and particulates at the Cook Plant.
The 0, C.
Cook Plant will be equipped with facilities to analyze these filters consistent with clarifications 1 and 2.
The equipment. for implementation of this requirement is on order and is scheduled for delivery in late 1981.
Therefore, the first available date for installation is the refueling outages in 1982.
As such, an extension is re-quested beyond the implementation date of January 1,
1982 until these r fueling outages.
RESPONSE
TO ITEM II.F.1 ATtTACHi4lENT 3 CONTAINMENT HIGH RANGE RADIATION i(ONITOR:
As s.ated in the above response to Item II.F.1, Attachment 1, the radia-tion monitoring system upgrade for D.
C.
Cook Plant is being designed to provide the required information of both this attachment and Attachment 1.
- However, since Attachment 3 stipulates significant changes from previous documents, a
detailed review of the impact of these changes cannot be provided in this submittal.
Specifically, the impact of Appendix 8 to NUREG-0737 regarding Design and guaIification of Equipment is still under review.
If deviation of this Attachment's positions or clarifications are necessary, a detailed explanation of and justification for, the deviation will be provided by July 1,. 1981, along with the documentation required for final review.
This System which is on order and being manufactured has been purchased in accordance with your earlier requirements of HUREG-0578, prior to issuance of NUREG-0737.
The portions of our upgraded radiation monitoring system for ccmpliance with this item will be installed by January 1,
1982 consistent with the requirements of Attachment 3.
RESPONSE
TO ITEM II.F.I ATTACHMENT 4
- CONTAINMENT PRESSURE MONITOR:
This item has been implemented as per our letters of October 24, 1979 (AEP:NRC:00253),
and January 18, 1980 (AEP:NRC:00334).
RESPONSE
TO ITEM II.F.I, ATTACHMENT 5, CONTAINMEiVT MATER LE'/EL:
The containment water level monitoring system which was purchased for Cook Plant will be implemented 1n accol'dallce with our letters of October 24, 1979 (AEP:NRC:00253) and January 18, 1980 (AEP:NRC:00334).
The required in-stallation and documentation will be completed by January I, 1982.
RESPONSE
TO ITEii II.F.I ATTACHMEiVT 6 CONTAINMENT HYOROGEiV MOiVITOR:
Oesign modi,ications and monitoring equipment that will satisfy this requirement will be installed and operational by January I, 1982.
The required documentation will be submitted by January I, 1982.
RESPONSE
TO ITEM II.F.2 INSTRUMEiVTATION FOR OETECTION OF IiVAOEOUATE CORE COOLIiVG:
As stated in our previous submit.als addressing Item 2.1.3.b of NUREG-0578, letters dated January 18, 1980 (AEP:NRC:00334), Attachment 6 and March 10, 1980 (AEP:NRC:003348),
0.
C.
Cook Plant is implementing an inadequate core cooling (ICC) system.
The system is to be composed of reactor water level indication supplied by Mestinghouse Electric Corp.
supplemented as necessary by the equipment available to monitor margin to saturation of the reactor coolant system.
The position, clarification, at.achments and appendices now imposed by this item represent a major significant change to those previously required by other documents.
Mestinghouse.Electric Corp. is presently preparing Topical Reports to.respond to the documentation requirements of this item.
A tentative schedule for submittal of these reports by Mestinghouse to the iVRC is January I, 1981.
A detailed review by AEPSC cannot be performed by the January I, 1981 documentation deadline for this item.
Relief to April I, 1981 is therefore requested from the documentation required date of January I, 1981 to allow for detailed AEPSC review of these Mestinghouse reports.
The reactor water level system will be installed consistent with the implementation schedule of this i 'CBll ~
Our existing core exit thermocouple system does not meet the requirements of Attachment I to Item II.F.2.
RESPONSE
TO ITEM II.K.2.13, THERMAL'ECHANICALREPORT --- EFFECT OF HIGH-PRESSURE INJECTION QN VESSEL INTEGRITY FOR SMALL-BREAK LOSS-OF-COOLANT ACCIDENT NITH NO AUXILIARY FEEDWATER:
A proaram will be completed and documented to the NRC by January 1,
1982 by the ':lestinghouse Owners Group to completely address the NRC requirements of detailed analy'si's of 'th'e Dermal:mechahi'cal conditi'ons in the reactor vessel during recovery from small breaks with an extended loss of'all feedwater.
This program wi',
1 consist of analysis for generic iA PAR Plant groupings.
RESPONSE
TQ ITEM II.K.2.17 POTEiNTIAL FOR VOIDING IN THE REACTOR COOLANT SYSTEMS DURING TRANS IEiVTS:
The WOG is currently addressing the potential for void formation in the Reactor Coolant System (RCS) during natural circulation cooldown conditions, as described in 'ilestinghouse letter NS-TMA-2298 (T. M. Anderson, N. to P.S.
- Check, iVRC).
A report describing the results of this effort will be provided to the NRC before January 1,
1982.
RESPONSE
TQ ITEM II.K.2.19 SE UENTIAL AUXILIARY FEEDMATER FLOW ANALYSIS:
The transient analysis
- code, LOFTRAN, and the present small break evaluations analysis
- code, MFLASH, have both undergone benchmarking against plant information or experimental test facilities.
These codes under appropriate conditions have also been compared with each other The MOG will provide on a schedule consistent with requirement of item II.K.2.19 a report addressing the benchmarking of these codes.
RESPONSE
TO ITEMS I'I.K.3.1 - INSTALLATION AND TESTiNG OF AUTOMATIC POHER-OPERATED RELIEF VAL'lE ISOL'ATION SYSTEM!
AiVD II.K.3.2 -
REPORT ON OVERALL SAFEI Y EFFECT OF POMER-OPERATED REL'IEF'lAL'VE ISOL'ATION SYSTEM:
The
~AGOG is in tfie process of developing a report to address the iVRC concerns *of Item II.K.3.2.
How ver, due to the time-consuming process of data gathering, breakdown, and evaluation, this report is scheduled for'submittal to the NRC on March 1, 1981.
As required by the NRC, this report will be used to support a decision on the necessity of incorporating an automatic PORV isolation system as specified in Item II.K.3.1.
RESPONSE
TO ITPil II.K.3.5 AUTOMATIC TRIP OF REACTOR COOLANT PUMP OURING LOSS OF COOLANT ACCIOENT:
The MOG resolution of this issue has been to perform analyses using the riestinghouse small break evaluation model '(vlFLASH) to show that ample time is available for the operator to trip the reactor coolant pumps following cer-tain size small breaks, see HCAP-9584.
Tnis is in accordance with our letter of June 20, 1980 (AEP:NRC:00419) which responded to Mr. Eisenhut's May 7, 1980 letter concerning additional TMI-2 requirements.
In addition the Owners Group is supporting a best estimate s udy using the NOTRUMP computer code to demonstrate that tripping the reactor coolant pump at the worst trip time, after a small break will lead to acceptable results.
For both of these analysis efforts, the HOG is performing blind post-test predictions of loft experiment L3-6.
The input data and model to be used with ')FLASH on test L3-6 has been submitted to the staff on 12/1/80 (NS-TMA-2348),
the information to be used with NOTRUMP on test L3-6 will be submitted prior to performance of the test as stated in letter OG-45 dated 12/3/80.
The Loft prediction from both models will be submitted to the staff on February 15, 1981 given that the test is performed on schedule.
The best estimate study is scheduled for completion by April 1, 1981.
Sased on these
- studies, the
~rlOG believes that resolution of this issue will be achieved wi.hout any design modifications.
In the event that this is not the case, a schedule will be provided for potential modifications.
RESPONSE
TO ITEMS II.K.3.9 PIO CONTROLLERS II.K.3.10 PROPOSEO ANTICIPATORY TRIP MOOIFICATIONSi II.K.3.12 ANTICIPATORY TRIP ON TURSINE TRIP:
These items have been implemented per our June 20, 1980 letter (AEP:NRC:00419)
RESPONSE
TO ITEM II.K.3.17, EMERGENCY CORE COOLING SYSTEM OUTAGES:
The following information provides our report on ECC system outages committed to in our letter of June 20, 1980 QEP:NRC:00419) in response to this item.
Shown in the attached tables are outage times for Centrifugal Charging, Residual Heat Removal, Safety Injection, Containment Spray and the Emergency Oiesel Generator Systems while the reactor was in iXodes 1, 2, 3 or 4.
Separate listings are given for Units 1 and 2 s arting from initial critic-ality, 1/18/75 and 3/10/78 nespectively, as shown in Tables II.K.3-1 and'I.K.3-2 which are attached.
In some instances the exact length of time a component was out of service could not be determined.
For these cases it was assumed that it was out half of the Technical Specifications maximum allowable time.
These cases are marked with an asterisk.
Me believe that Technical Specifications for the ECC systems in Cook Plant are sufficiently restrictive to keep control of their availabil,ity.
No further actions are required for this item.
RESPONSE
TO ITEM II.K.3.25 EFFECT 'F LOSS OF AC PO!IER ON RCP SEALS:
The 0.
C.
Cook Plant is not susceptible to seal failure upon loss of off-site power.
The ',oss of of -site power results in a trip of the Reactor Coolant
- Pumps, but the Centrifugal Cnarging Pumos (CCP's) continue to supply seal injec-tion to the RCP's upon loss of off-site power.
The CCP's are powered from the Emergency Oiesel Generators.
RESPONSE
TO ITEMS II.K.3.30 SiilALL BREAK LOCA NETHOOS AND II.K.3.31. COHPLIANCE MITH 0 CFR 50.46:
These items have been implemented as per our letters of June 20, 1980 (AEP:HRC:00419) and October 1,
1980 (AEP:HRC:00398A).
RESPONSE
TO ITEiM III.A.1.2, UPGRAOE EMERGENCY
RESPONSE
FACILITIES:
IhMECo is presently reviewing its Emergency support centers versus the requirements of NUREG-0654 and the results or the review will be part of a re-vised and upgraded Facility Emergency Plan (EP) for the Oonald C.
Cook Nuclear Plant.
This plan will be submitted to the HRC and will address the upgraded emergency planning rules of 10 CFR 50.54 and Appendix E to 10 CFR 50 (Item III.A.2 below).
!le have submitted significant comments to the Comission on the draft of HUREG-0696 and we await issuance of the final criteria.
RESPONSE
TO ITEM III.A."2 IMPROVING LICENSEE EMERGENCY PREPAREDNESS-LONG TERM:
Revision 1 to NUREG-0654 was published in November, 1980.
Me are"in receipt of this revision and are in the process of revieuing and revising our EP, where necessary, to address the changed'criteria indicated in Pevision 1
to NUREG-0654.
Due to the time required for management review, final typing and printing we have s<<t the NRC a separate letter requesting an extension of the submittal date for this report to January 26, 1981. See our letter of
')ecember 3Q Ig8O(own.a<nC.O3O8C)
RESPONSE
TO ITEM III.D.3.3 IMPROVED INPLANT IODINE INSTRUMENTATION UNDER CCIDENT CONDITIONS:
The following cart mounted, continuous air monitors are available for use in an emergency that may involve airborne radioactivity concerns:
Two Nuclear Measurements Corporation, Model AM-221, which monitors air-borne particulate (beta scintillator detector) and radioiodine (Si iver Zeolite cartridge and single channel analyzer calibrated to 365 KeY).
These units are dedicated to emergency use in the Technical Support Center and either Control Room.
Two Eberline Model P!NG-1 airborne particulate and radioiodine monitor.
Radioiodine monitor, usually used with TEDA imoregnated
- charcoal, will also accept the Silver Zeolite car ridge.
The detector is con-nected to single channel analyzer calibrated to 365 KeY.
Thre Eoerline Model PING-lA airborne particulate,
- radiogas, and radioiodine monitor.
Radioiodine monitor, usually used uith TEDA impregnated charcoal, will also accept the Silver Zeolite cartridge.
The detector is a stabilized NaI detector connected to a two channel analyzer calibrated to 365 KeV with automatic Xe subtraction from the second channel.
All cart-mounted iodine detectors are in 3 inch lead shields.
In addition, there are available for use throughout the plant ten Eberline RAS-1 regulated air samplers, which accept either TEDA impregnated charcoal or Silver Zeolite cartridges.
In addition to the equipment normally available in the regular radio-chemistry counting facility, the following analysis equipment is available for analysis of the Silver Zeolite cartridges that might be used in an emergency:
A 4" x 4" NaI crystal connected to a Packard 1024 channel MCA is located in the low background counting facility.
-b.
In the basement assembly area there is a cartridge purge unit con-sisting of a T-size bottIe of dry nitrogen, regulator, cartridge
- holder, and associated piping to permit purging of Silver Zeolite or charcoal cartridges with dry nitrogen.
c.
Located in the basement assembly area is an Eberline MS-2 single channel
- analyzer, calibrated to 365 Ke'i, connected to a 2" x 2" NaI crystal in a 2i-" lead shield designed for counting in TEDA-charcoal or Silver Zeolite cartridges.
RESPONSE
TO ITEM III.0.3.4.
CONTROL ROOfl HABITABILITY:
The NRC position to assure that control room operators will be adequately protected against accidental release of toxic gases and radiation to operate and/or shut down the plant under design basis accident conditions is presently under review.
Our present schedule is to complete the yeview and evaluation and make the requir d submittal of information by February 2, 1981.
- Thus, we are requesting relief of the submittal deadline of January 1,
1981 to February 2,
1981 which will allow us sufficient time to complete our review efforts.
ATTACHMENT I.A-1 IiNDIANA 8 MICHIGAN ELECTRIC -vs-INPO STA INITIAL TRAINING PROGRAM
indiana 8 Hichi) an Electric IHPO Hath background assumed per IIIPO for engineering or pliys ical science degree.
Reactor Tlienry
- Ato)))fc and tlucleav I'liyslcs
- interacLInn of Badiatlun )arith Hatter
- Tl)e Fission Chain Reaction
- t{euLron Diffusion and Hnderation
- tluclear Reactor Tlieory (including tleutron Hiiltip'llcaLion Factors for a Ileterogeneous Reflected Tliermal Aeactn) )
- The Till)e"Dope))de)it Anacin) a.
Reactor Kinetics b ~
Colltrol Hechalli sll)s c.
Reactlv I ty I'eeilback d
I lss In))
I )'nduct I nlsnnlng e.
Core Cliaracteristlcs and Properties During L I fet$ n)e Eng lneering n)a tl)
- Ordinary Differential Equations
- Laplace Transforms Reactor Tlieory
- htn)))Ic and tluclear PI)ysics
- Reactor Statics
- Two Group Tlienry
- Dynan)lcs, Point Kinetics
- Reactivity Feedback
- Subcril:ical HulLlplication and I/lh Plots
- Esti)))ates Cvitical Position Calculations (Reactivity Balances)
- Slnitilown Havgin Calciilatlons
- Slnitdnwn Cool ing Require)))cuts
- Reactor Safety Considerations Aeacto)
Che)))is try
- Covroslon, Aeactlon Aates llucleav Hetaluvgy
- Crystallograpl)y/Pliase Diagrams
- llardenlng
- Response Lo Stress and Temperature
- P)opertles of fleactor Haterials
- RIrcaloy - Water Aeactlnn
- Tylies of Corrnsion Aeactor Cliemlstvy
- inorganic Clieml s try
- Covroslon, Reaction Bates Ih)clear Haterlal s
- Strength of Haterlals Beacl.ov Haterlal Pvopevties
- I'liase Diagrams
- Fuel Densiflcatlon TheA))odyl)al)iles/Fluid Oynan)ics/llea t Trans fer
- Laws of The))))odyna)))ics
- Equations nf State
- Steam Cycles/II'ficlency
- llevnniil1 I 's Eipiatinn
- FliiiilI'vlcLln)i and lleail l.oss Tliermo(lynamics/Fluid Dynamics/Ileat Trans fev
- Laws of The))))n)lyna)))ics
- Pvnpertles of 11ater/Steam
- Steam Cycles/Efficiency Oe)'no)Ill) s E)jua t in)1
- Fliild Frlctlnn anil lleail I.nss
Ili '
Therooolynami cs/Fluid Dynamics/Ilea t Trans fer (Cant)
- Compressible Flaw
- Incooqiressib'le Isentropic Flow
- Real Flovi Prableo~s
- Peop Characteristics
- Twn Phase Flow
- lnstrlooentaLIon
- Hetl)ods of Ileat Transfer
- Specific I!eat, Expansion, Viscosity
- Viscous Flow
- Cniobined Conduct;ion/Convection
- IIuci cate/Fl los Ooil ing
- Critical IleaL Flux IIIPD Tlieoondynamics/Fluid Dynamics/Ileat Transfer (Co<li).
- Elevation Ilead
- Pump and Systeo Characteristics
- Twn Phase FInw
- Hethods of Ileat Transfer
- Boiling Iteat Transfer
- Ileat Exchangers Electrical Sciences
- 4160 Volt Electrical Distribution
- Protective Belaying for Generators
- 600 VAC, 120 NC, and 250.-VDC Electrical D I s tribul,Ion
- Steam Generator Level Control
- I'eed I'Woe) Si>eed Control
- Pressur I zer I.evel/I'ressure Cnntrol
- Full Length Aod Control Nuclear Instreoentattan and Control
- Excore Iluclear Instnooentatlan Systeo)
- lncore IIucleal Instriooentation System Electrical Sciences
- Electronics
- Hntnrs, Generators, Trapsforu)ers, Switchgear
- Instruoientatton and Control Theory IlucIear Instrmoentation and Control
- Bad I a tion Detectors
- Aeac tor lns tnaoen ta t Ian
- Reactivity Control and Feedback A/P and Ileal t,h Physics
- Si te Emergency Plan and Io)ple~oenting I'rocedures
- AppIicabie Radiation Protection, Concepts Contained In 10 CFfl 20 and 10 CFB 100
- Aa~llologlcal Control Instructions
-,Aa<ltatian Hant t.nring System
- I'nrtable Itaitlatton Hanitorlng Instrumentatinns Plant. Speci fic Reactor Tectu>alogy
- AeacLor Cnre System Reactor Coolant System
- Pressurizer and Pressure Belief Systems
- Full I.engLh An>Lrnl Systeoi
- Residual IleaL Beoinva)
Systein
- Emergency Core Cooling Systeo>
- Iteacta>
I'rat.ection Syst.uo linc)ear-Aadiatinn Protection and Ileal th Physics
- Diologtcai Effects
- Aadlat tan Survey Instrumentatinn
- Shielding Pla>>t Specific Reactor Technology (Including Core I'hysics Data)
~~ll E1
- Plant Spec)fic Reactor Teclinology (Cont}
Plant Cliemlca'I and Corrosion Control
- Chemical and Yolunie Control and Ooron Hakeuli/Recovery Sys tern
- Pr)mary a>>d Seen<<davy Sampling Systein Plant Irist.rumentat lan and Control
-See Item //6, Elect;rical Sc)ences, a<<d Item k'I>
Nuclear Ins Lvlllnenta t)on and Control Plant Ha ter lais
- See ILeni A, 'N<<clear Hetalurgy'lant Thennocycl e
- Steam, Condensate, and Feed Systeins
- Steam Dump Systems
- Steam Generating anil Steam Generator Systenis Hanagement/Siipervisory
- Lo be covered on a select:ed personal basis afLev initial STA Tra)n)ng Program Plant Spec) f)c Reactor Techno]ogy
- Plant Chein)stry and Covvos)on Control
- Reactor Instrumentation and Control
- Apactor, Plant )later)als
- Reactor Plant Thevniocycle management/Superv) sory Sk) 1 ls
- Leadersh)p
- Interpersonal Conimunicat)on
- Hntlvation of Personnel
- Problems and Decis)onal Analysts
- Conw>and Responslb) lity and L)m)ts
- Stress
- lluman Oehav)ov Plant Systems
- Reactor Core System
- Reactor Coolant Systein
- Pressurizer and Pressure Ael)ef Systems
- Full Length Aod Cont,rol System (liicluding Aoil Posit)on Indicat)on}
- Chemical 5 Vol<<inc Control System and Boron Hakeup arid Recovery Systenis
- Aes)dual lleat. Aenioval Syst: em
- Emergency Core Cooling System
- Excore tlucleav Instr<<inc>>tat)on Systen
- I<<cove H<<clear Instr<<me<<tat)on Systein
- Reactor Pro tee L)on System Conta)nnienL SysLem
- Ice Condenser Syst: em Containine<<L Spray and Hydrogen Aecombiner System Plant Systems
- Reactov Coolant System
- Reactor Control I
- Reactor Coolant Inventory and Chemistry Control
- Aes)diial ileat Removal Systein
- Einergi.ncy Core Cool)ng System
- Einergency Cool lng )later
- tliiclear Instrumentat)on
- Reactor Protection Systein
- Containnient System (including Conta)nment Cool)ng)
- Co<<talninentllydrogen tlonltor)ng and Contro'I
Indiana IlHiclrioan Electric
~IN 0 (Cont) ll-Pl ant Systeors essential service water system non-essential service water system spent frrel pi t cooling and cl eanul) waste disposal systeor - liquid and gaseous only containroent ventilaLIorr systeor auxll lary lruildirrg and control room ventilation sys tears eorevgerrcy diesel generator system auxiliary feerlwater systeor corupressed air sysLem priorary water sys teor priurary gas systeor water fire protection system carlron <I!oxide fire protection system roiscellaneous fire protection system radiation rooni torirrg sys tero portal)le va(liat loll ololliIori rig 'lns tvrlolents steaor generating and stearo generator water level control sys teors
- steaor, condensate and feed.systeors steam duorp systeor 4160 KV electrical distribution system 600 VRC, 120 VRC, and 250 VDC electrical distribution system NOTE:
see incore instrumentation and reactor coolant systeors Plant Systems
- plant ventilation
- emergency electrical power
- auxiliary feedwater system
- eorergency control air
- radiation oronitoring system
- steam generator level control
- main steam, condensate, and feedwater
- non nuclear instrrooentatlon loose part oronitoring status monitoring {including computer)
- seismic monitoring
)2 - hdmirrlstr.ative Control
.- Beslronsibilities fov Safe Operation 5 Shrrtdown
- Equiproent Outages and Clearance Procedures
- Use of Procedures
- Plant Hodi fications
- Shi ft !Iellef Turnovev and Hannlng
- Contairrment Access
- Haintainirrg Cognizance of Plant Status
- Unit Int,erface Controls
- Plrysical Secrrr'ity
- ConLvol lhrour Recess
- I)rrLIr.s a>>d Iles!>nrrsilrllities of Lire STA
- IIadlologica1 Eoiergerrcy I'lan
- Co(ie of fe(ler'rl IIe<jrrlatiorrs (air!rvoirviate sections) hr!orirristrative Control
- Responsibilities for Safe Opevation 8 Shrrtdown Eqrllploellt OuLages and Clearance Procedures
- Use of I'rocedures
- Plant lhdifications
- Slrift IIelief Tuvnover and Hanning
- Contairooent Access
- Haintaining Cognizance of 'Plant Status
- Unit Interface Corrtrols
- I'lryslcal Secrrvlty
- Corrtvol Honor Recess lhrL!es arrrl Aes!rorrsibilities of tire STA
- IIa<llological furer gency Plarr Corle of fe(leval IIegrrlations (approlrriate sections) 4
Indiana 8 Hichi an Electric INPO (Cont) 12 - Administrative Controls Administrative Controls
- Plant Technical Speci ficatjons
( including bases)
- Plant Technical Specifications (including bases)-
- Radiological Control Instructions Ifadiological Control I>>structions 13 - Transient and Accident Analysis for Shift Technical Advisors with reference to the I:SAII and Plant Abnormal and Emergency Procedures 14 - tlormal, Transient; and Emergency Opera tions (Simula tor Training) 15 - General Operatj>>g Procedures
- tiOTE; covered in systems and reactor theory lectures Transient and Accident Analysis for Shift Technical Advisors with reference to the FSAII and Plant Abnormal a>>d Emergency Procedures Normal, Transient, and Emergency Operations (Simula tor Training)
General Operating Procedures Startup
- At power operations
- Sl>>itdnwn
- Xenon followjng while on standby ECP and S.D, margin calculation
ATTACHllENT I.A-2 STA REOUALIFICATION PROGRAM
The Shi,t Technical Advisor Requalification Program shall be con-ducted on an annual cycle basis.
The requali,ication program shall consist of:
1.
Formal classroom lectures Z.
On-the-job training (including simulator trainino) 3.
An annual eval,ua.ion 4.
Training documentation.
Lectures shall be conduc ed in the following "ar as with emphasis on identified weak or problem areas:
a.
Theory and Principles of Oper ation (includes Thenmdynamics, Heat Transfer and Fluid Flow) b.
Cwneral and Specific Plan" Ooeraiing Characterisacs c.
Plant Instmentation and Control Systems d.
Plant Protec ion Sys.ems e.
Engineered Safety Systems f.
Normal, Abnonral and Emergency Operatino Procedures g.
Radiation Con rol and Safe:y h.
Technic 1 Specifications i.. Applicable portions of Title 10, Chapter 1,
Code of Federal Regulations
%he use of.raining aids such as video~pcs or -,.ilms may be used in lieu of an instructor.
- However, no rare dan 50" of the lecture series shall be solely video" pe or film.
The annual lec ure series wi11 be of an estimated lenoth of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />, but in no case less than "0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.
Lectures shall be evenly spac d
throughout the period, taking infrequent operations such as refueling operations into account.
'Aritten quie~a will be administered a ~~r each lec ur topic -,or the valua ion of individual knowledae level and progress.
On-the-job training shall consist of:
a.
Sucervision and/or perfomance of control manipulations (simulator training}
b.
On-shi,t abnormal and emergency proc dure review c.
Keeoing abreast of all facility and procedur changes d.
Review of NRC, vendor, and indus ry related notices.
Each shif..echnicaI advisor shall, during the requalifica-.ion.rain-ing cycle, per.orm/supervise a minimum of plan control manipuivions which demonstrate his skill and/or familiarity wiD plant control systems.
Each shift.echnical advisor shall either dir ct or evalua
~ the ac.ivities of others or manipulate. the controls during these con-rol manipulations.
It shall be emphasid that the shi
~ technical advisor serve in his designated role during these control manipulations where possible.
As many of the following control manipulations as possible should be perforned during 'each requalification cycle.
Tne aster isked i am shall be performed annually.
<<1.
2.
3.
7.
8.
9.
10.
<<11.
12.
20.
21.
22-
"23.
24.
25.
Plant or reactor startuos o include a range that reac ivi y feedback from nuclear heat additi'on is noticeable and heatup rate is established.
Plant shutdown Manual control of steam generators and/or feedwater during surtup or shutdown.
Boration and/or dilution during power ope. ation Any significant (>10~) power changes in manual rod control Loss of coolant, including:
1.
Significant steam generator leaks.
2.
Inside and Outside primary containment.
3.
Large and small, including leak-rate determination.
'4.
Sa ura ed Reactor Coolan-response.
Loss of electrical power (and/or dearaded power sources).
Loss of core coolant flow/natural circulation.
Loss of Essential Servico Mater.
Loss of shutdown cooling.
Loss of Component. Cooling Syst m or cooling to an individual componen Loss of normal feedwater or normal,"-eecwater System failure.
Loss of all feedwater (normal and emergency).
Loss of protec ive sys-em channel.
Nspositioned control rod(s) (or rod drops).
Inability to drive control rods.
Conditions reguiring use of emergency boration.
Fuel cladding failure or high ac-ivity in r ac or coolant or offgas.
Tur'ne or generator trip.
i~lfunction of Automatic Contml System(s) which a=fee.
reacti vi s.g.
~'hl func ion of Reac.or Coolant Prwsure/Solus Con rol System.
Reactor Trio.
iNain s.earn line break (inside or outside containment).
Nuclear Instrumentation failure{s).
Required at leas anaual'i'y.
Even ifthe above manipulations are not ne ded to be accoaqlished by a simulator, each shi i technical advisor shall attend a training session at an appropriate simulator annually.
Abnonal and emergency procedures shall be reviewed by all shift technical advisors on a regularly scheduled basis as assigned by the Training Coordinator.
Tne proc dure review shall normally be accomplishea each shi t cycle by on-shift selfs udy.
Other ar as of interes.
may be included in the periodic r view assignment.
All abnormal and emergency plant operating procedures shall be reviewed at least annually.
All shi
~ technical advisors shall r view on a continuous basis all changes in facili> design, operating procedur s and the facility license.
Tne detewnation of the depth of r v e~ of any changes shal!
be made by 4
Training Coordinator or cognizant Cepartmnt Head.
Reviews shall be conducted by one of the followina methods:
1.
Formal training lectures, to be scheduled and conducted during requalification~1ectures.
2.
Individual review,.to be r ad by the individual auring h-;s normal work hours.
guestions to be directed to De Training Oepa~ent.
3.
Shift aroup discussion, to be conduct d on-shi by the Shi-Operating ~ngineer.
A11 shift technical advisors shall r ceive a,written examination annually to determine the effec"iveness of the overa11 requalific tion program and to define those ar as where additional ~hasis is reauired.
An overall grade average of less than 80 or any cateaory arade of. less, than 70" sha11 require the individual to be placed on an accelerated training proaram pr par d to correc the identified weakness.
Tne scope and duration of he accelerated training orogram shall be based upon managem nt evaluation in each instance it is required.
A permanent record sha'il be maintained for each shif. technical
~
advisor containing verifica ion of each proaram comleiion and the over-all grade scores for the annual written examination.
Tnis permanent record file shall be maintained for De lif of the facility.
ATTACHMENT II. E-1 Information for implementating positions 6 and 7 of item II.E.4.2 of NUREG-0737 in Cook Plant.
Our response to guestions 022.4 and 022.13 contained in Appendix g, Amendments 77 and 78 to the FSAR, submitted in July and October 1977 respectively.
This provided our response to Branch Technical Position CSB 6-4 including the impact on ECCS performance and an evaluation of the radiological consequences of a design basis accident during purge operation.
The response to guestion 022.4 and 022.13 apply to both Units of the Cook Plant.
2.
Our submittal of Oecember 29, 1977 on Unit 2 provided the procedure for the valve operability test.
The scope of this tes. procedure was reviewed with members of your staff prior to our submittal.
This applies to both Units 1
and 2 of the Cook Nuclear Plant.
3.
Our submittal of January 13, 1978 on Unit 2 provided the results of the in-situ purge valve ooerability tes performed on January 8, 1978.
This test was a pre-requisite for allowina unrestricted purging of the containment in accordance with our response to Containment Systems Branch guestion 022.4 of our FSAR.
The test results demonstrated that the purge valves are capable of closing against the dynamic forces of a design basis loss-of-coolant accident.
These results were submitted to the Commission in support of our Technical Specification change re-quest on Unit 2 to allow unrestricted purging of the containment.
This test, its results and the various supporting analyses we have performed address the concerns expressed in Mr, Eisenhut's letter of September 27, 1979 and no further action is required.
- his apolies to both Units 1 and 2 of the Cook Plant.
Our submittal of February 3, 1978 on Unit 2 provided supplemental in-
,ormation requested by your s.aff concerning the results of the valve operability test in support of our Technical Soecification change request.
This applies to both Units 1
and 2 of the Cook,Plant.
5.
Our submittal of April 27, 1978 on Unit 2 supplied analyses that demonstrate the operability of the lower compartment ourge system based on the test already performed.
The analysis provided shows that although lower compartment pressures might be higher than the test pressure, the pressures expected at the inboard containment isolation valves in the lower compartment purge and vent lines would be less than the pressure which existed during the valve operability test.
This is achieved by installing debris screens in the lower compartment purge systems which provide a high flow resistance.
This applies to both Units 1 and 2 of the Cook Plant.
6.
Our submittal of August ll, 1978 (AEP:NRC:00069) on Unit 2 provided additional information requested by your staff and applies to both Units of the Cook Plant.
7.
Our submittal of September ll, 1978 (AEP:NRC:00082) on Unit 2 provided sensitivity analyses of the resistance coefficients for the elbows and debris screens and the dependence on those coefficients of the resulting to~que and applies to both Units I and 2 of the Cook Plant.
8.
Gur submittal of January 4,
1979 (AEP:NRC:00114) on Units 1
and 2
provide our response to Mr. Schwencer's November 28, 1978 letter.
Allof the requests for additional information and justi,ication of unlimited purging were provided.
Additionally, we provided our review of the issue of overriding of safety actuation signals and the procedural steps taken to assure that operation of a bypass will not affect safety functions.
9.
Cur meeting with the NRC staff on May 31, 1979 to discuss the statu of review of the containment purge and related subjects.
The NRC staff informed us that we had a favorable writeoff as far as valve operability was concerned,
- However, we were told that the NRC staff required urther action from AEP on the issue of manual override of safety actuation signals.
10.
Our submittal of June 8, 19?9 (AEP:NRC:00114A),
applicable to Units 1
and 2 provided a descriptions of the modifications made to the reset/
block circuits and associated procedural changes required to meet the Commission's position as committed to at the May 31, 1979 meeting.
ll.
Gur submittal of June 29, 1979 (AEP:NRC:001148) on both Units 1 and 2
provided the additional information on he subject of unrestricted purging that was requested at the May 31, 1979 meeting.
12.
Our submittal of November 8, 19?9 (AEP:NRC:00295) provided our response to Mr.
D. Eisenhut's letter of September 27, 19?9 which dealt with con-tainment purging and venting during normal operation.
This submittal also provided the completion of our review of overriding of safety actuation signals.
13.
Gur submittal of Oecember 5, 1979 (AEP:NRC:00295A) provided further-information requested by the NRC staff concerning the monitoring of containment radiation in the context of the purge system operation.
14.
Our submittal of March 10, 1980 (AEP:NRC:00370) provided our response to Mr. 0. Eisenhut's letter of february 11, 1980 regarding the Com-mission's interim position concerning containment purging and venting during normal plant operation.
15, Our submittal of March 25, 1980 (AEP:NRC:002958) provided further information requested by the HRC staff concerning the control room isolation function and containment radiation monitors as a result of the on-going review of containment purging matters.
0 TABLE II.K.3 -
1 E
Date/Time 1-18-75 2-25-75 3-3-75
. 1-13-75 4-7-77 5-17-77 8-9-77 8-11-77 8-30-77 9-12-77 9 14-77 2-6-78 2-7-78 9-7-78 9-7-78 10-3-78 L'-7-78 11-10-78 1'18-78 LL-25-78 11-30-78'2-6-78 1-17-79 1-26-79 1-30-79 2-15-79 2-'-79 2-18-79 Ounatian 57 hrs 62 hrs 36 hrs*
57 hrs 67 hrs 36 hrs*
9 hrs 36 hrs 11 hrs 36 hrs*
5 hrs 3.3 hrs 3.5 hrs 3 hrs 15 hrs 4.25 hrs 8 hrs 5 hrs 8 hrs 12 hrs S.5 hrs 36 hrs>>
9.5 hrs 33.5 hrs 5.7 hrs 3.4 hrs Eauiament
.Initial Criticality "AB" Diesel Engine "CD" Diesel Engine "CD" Emergency Diesel "AB" Emergency Diesel East Centrifugal Charging Pumo "ast le CDII RHR Systan emergency Diesel Mest RHR Train IIABIe East East
'Rest Nest Emergency Diesel Generator Centrifugal Charging
- Pump, Centrifugal Charging Pump Centrifugal Charginc P mp Centrirugal Charging Pump RHR System Mest Centrifugal Charging Pump East Centrifugal Charging Pump Eas Centrifugal Charging Pump "AB" Diesel Generator
'Rest RHR "CD" Diesel "CD" Diesel "AB" Emergency Diesel "CD" Emergency Diesel North SI Pump South SI Pump Eas" CTS Pump "AB" Emergency Diesel jieeean Lubrication oil piping modifications Lube oil piping modifications leak Pump shaft severed Broken shaf leak repair wrist pin replacement motor modification wrist pin'repIac~ent maintenance tach. modifies ion tach. modification maintenance maintenance valve repair replace voltage regulator maintenanc maintenanc oil leak clearance on It~0-312 design change repair RHLZBM design change seal railure maintenance chancre oiI straine, repair gPIZ57 isolation vlv.
UNIT 1 TABLE II.K.3 -
1 4
Date/Time 2/27-2/28/79 2/27-2/28/79 2/27-2/28/79 2-23-79 3-(1-7)-79 3-l-79 3-3-79 3>>12-79 3-16-79 4-(5-10)-79
'-23-79 8-1-79 9-24-79 9-25-79 9-25-79 9-28-79 1-30-80
'-31-80 2-26-80 3-5-80 3-6-80 2.2 hrs 2.5 hrs 1.2 hrs 7.6 hrs 2.2 hrs 11.2 hrs 6.0 nrs 36 hrs~
36 hrs*
36 hrs*
5 hrs 2.6 hrs 4.5 hrs 1 hr 2 hrs 6.5 hrs 3.4 hrs 6 hrs 1.5 hrs 30.75 hrs Eaui omen t East and Mes RHR Pumps North and South SI Pumps East and Mest CTS Pump "AB" Diesel Mest Centrifugal Charginc Pump "CD" Diesel "AB" Diesel "CD" Emergency Diesel Gene. ator "AB" Emercency Diesel Gene~ tor "AB" Emergency Diesel North SI Pump Mes Centrirucal Charg ng Pumo South SI Pump North SI Pump East RHR Pump Mes RHR Pump "AB" Diesel "CD" Diesel "AB" Emergency Diesel East Centrifugal Charging Pumo East Centrifugal Charging
?ump Reason brkr inspection brkr inspec.ion brkr cleaning maintenance on ESM check vlvs.
clean brkr.
CD-work on ESM ck.vlv.
AB-work on ESM ck.rlv.
maint nance maintenanc ma ntenance repack ICN-250 maintenance maintenance maintenance malnteflaflc maintenance Per. ormanc
.as.- nc replace relays - RFC maintenance maintenance clean
<<CC change rotating assembly 3-26-80 3-30-80 3-21-80 3-24>>80 5.2 hrs 3.5 hrs 4.1 hrs 2.1 hrs 2.7 hl's 1.6 hrs East Centrifugal Charging Puma East Containment Spray Pump South Safety Injection Pump Mest RHR Pump Mest Containment Spray P mo Mest Centrifugal Charcirc Pumo clean CONO filter on lube oil system maintenance clean brkr.
clean brkr.
clean brkr.
clean brkr.
'Hii 1 TABLE II.K.3 -
1 Oaae/Tfme 3 25>>80 3-26>>80 3-27-80 4-l9%0 Onnaai on 2 hrs 9.0 hrs 9.3 hrs 1.75 hrs Eouicment East Centrifugal Charging Pumo Reaaan clean brkr.
"AB" Emergency Diesel East Centrifugal Charging Pump clean output brkrs.
change oil "CD" Emergency Diesel Generator clean output brkrs.
UNIT 2.
TABLE II.K.3 - 2 Oate/Time 3-10-78 4-16-78 5-15-78 7-11-78 7-18-78 9-11-78 9-12-78 9-19-78 10-3-78 10-11-78 10-lo-78 10-25-78 11-17-78 17 79 1-11-79 1-12-79 1-24-79 1-31-79 4-23-79 Ouratinn 10 hrs.
11.7 hrs 9.5 hrs 8.7 hrs 49.5 hrs 2 hrs 36 hrs*
16 hrs Oe5 hrs 1.75 hrs 18.9 hrs 4.2 hrs 2.1 hrs 8.3 hrs 17.3 hrs 7.7 hrs 21 hrs Eoui cment Initial Criticali.y Mest RHR Train "CD" Emergency Diesel "AS" Emergency Diesel "CD" Emergency Diesel "AB" Emergency Diesel "CD" Emergency Diesel Mest RHR Pumo East CTS Train "CD" Emergency Diesel "AB" Diesel CD':nercencj "AB" Emergency Diesel Generator E
RHR Pum M RHR Pump East Centrifugal Charing Pump "AB" Diesel Generator "AB" Emergency Diesel Generator "CD" Emergency Diesel Cenerator Reaaen'rroneously decIar&
inoperabl e (RH-104M Pos.
Indication at mid-position.
4 repair water leak on Air Cooler.
replace ESM Supply Spool Piece repair fuel in<ec ion Starting Air Ck. Vlv.
leaking through fuel oil le ks Repair brkr. Ieakace Valve ?roblem could not verify flow.hrougn CTS-:20E - Broken sha-,t ji"I0-27.
routine maintenance Failed to s art ~//ice.
Relay Calibration routire maintenance iubricat on lubrication oil change Repair ESM Supply VIv.
Repair
=" M-143 Repair ESM Supolv ck. viv.
7 1879 8-27-79 7.5 hrs 8 min "AB" Emergency Diesel Testing Co2 o
Diesel Rooms.
Emer.
"CD" Emergency Diesel Generator Repair leaking Ck. Vlv.
8-27>>/9 8 min.
"CD" Eme. gency Oiesel t esti ng Coi o
Diesel Roon s
Gate/Time 9-26-79 9-29-79 10-15-79 10-16-79
~ 1-18-80
.1-21-80 1.-30-80 1-31-80 2-1-80 2-4-80 e
Ouration 36 hrs*
7 hrs 9.1 hrs 11 hrs 8 hrs 13 hrs 7.2 hrs 7.3 hrs 6 hrs 2.0 hrs TABLE II.K.3'- 2 Eoui oment Reason East RHR Pump Mest Centrifugal Charging Pump routine maintenance repack coupling East Containment Spray Pump Mest Containment Spray Pump North SI Pumo South SI Pump Mest Centrifugal Charging Pump North SI Pump remove suction s raine remove suction s rainer reprove suction s.rainer remove suc.ion straine.
check suction straine.
Repair SV-98 on Oischarge He l'er "CQ" Emergency Oiesel Generator routine inspection "AB" Emergency Oiese'1 Generator routine inspection 2A-80 2-11-80 2-14-80 2-28-80 3-17-80 4-1-80
~10-80 4-15-80 4-16-80 12.5 hrs 39.2 hrs 9.2 hrs 7.9 hrs 7.1 hrs 10 hrs 57.5 hrs 5.9 hrs 1.5 hrs East RHR Heat Exchanger Mes RHR Pump 5 Heat Exchanger Mest RHR Pump East Centrifugal Charging Pump Mest CentrifugaI Charging Pump Mest Centrifugal Charging Pump Mest Centrifugal Charging Pump Mest Centri fugaI Charging Pump Mest Centrifugal Charging Pump Repair 'SV-104 Repack IHO-324 Repack ISO-322 remove suc.ion strainer Fitting leak oil supply line Replac gasket cn suction strainer Seal Failure Repair leak on
- CCM,
-.o He t Exch.
Output brkr. maint.