ML17331A571

Forwards Response to NRC 801031 Ltr Re Implementation of post-TMI Requirements Contained in NUREG-0737

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 DOCKET ~

FACIL:50 315 Donald C ~ Cook Nuclear Power Plantr Unit ii Indiana 5n 316 Donald CD CooK Nuclear power Plant> Unit 2i Indiana 05000316 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 COPIES RED IP IFNT Cnr rES ID CODE/NA~E <TTR ENCL IO I;ODE/NA>E L TTR ENCL' ACTION: - 4'ALGA r S ~ 04 13 13 INTERNAL D/DIRrHUV FACQo 1 1 OIR Or'J nF L I C 1 IR2. Qo 2 uRC POR 02 1 1 1[ 0 0R AssEss pR in 0 REG F ILF 01 1 1 EXTERNAL: ACRS le LPDR 1 1 NSIC 1 1 JAg 1 8 tgg~

TOTAL NUDGER OF COPIES REQUIRED: LTTR 3 ENCL

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.

Very truly yours, J

'4 R. S. Hunter cc: attached 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 isI the 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 .

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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 of knowledge and proficiency as required by ANS 3. is estab-continuous'egree 1

lished for the Donald C. Cook Nuclear Plant Units and 2.

1 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 response to Oenton's letter of in our September ll, 1980 letter (AEP:NRC:00395)

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 an immediate safety assessment function. it If is found that the-notice contains information of significant importance that should not, wait for duty'n 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 of the requirements for implementation of this item will'eAn perassessment 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'ECHANICAL REPORT --- 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', consist of analysis for generic iA PAR Plant groupings.

1 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 which are attached. and'I.K.3-2 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 Eng lneering n)a tl) for engineering or pliys ical science - Ordinary Differential Equations degree. - Laplace Transforms Reactor Tlienry Reactor Tlieory

- Ato)))fc and tlucleav I'liyslcs - htn)))Ic and tluclear PI)ysics

- interacLInn of Badiatlun )arith Hatter - Reactor Statics

- Tl)e Fission Chain Reaction - Two Group Tlienry

- t{euLron Diffusion and Hnderation - Dynan)lcs, Point Kinetics

- tluclear Reactor Tlieory (including tleutron - Reactivity Feedback Hiil tip'llcaLion Factors for a Ileterogeneous Reflected Tliermal Aeactn) )

- The Till)e"Dope))de)it Anacin)

a. Reactor Kinetics b ~ Coll trol Hechalli sll)s

c. Reactlv I ty I'eeilback d I lss In)) I )'nduct I nlsnnlng

e. Core Cliaracteristlcs and Properties During LI fet$ n)e

- 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 Aeactor Cliemlstvy

- Covroslon, Aeactlon Aates - inorganic Clieml s try

- Covroslon, Reaction Bates llucleav Hetaluvgy Ih)cl ear Haterlal s

- Crystallograpl)y/Pliase Diagrams - Strength of Haterlals

- llardenlng Beacl.ov Haterlal Pvopevties

- Response Lo Stress and Temperature - I'liase Diagrams

- P)opertles of fleactor Haterials - Fuel Densiflcatlon

- RIrcaloy - Water Aeactlnn

- Tylies of Corrnsion TheA))odyl)al)iles/Fluid Oynan)ics/llea t Trans fer Tliermo(lynami cs/Fluid Dynamics/Ileat Trans fev

- Laws of The))))odyna)))ics - Laws of The))))n)lyna)))ics

- Equations nf State - Pvnpertles of 11ater/Steam

- Steam Cycles/II'ficlency - Steam Cycles/Efficiency

- llevnniil1 I 's Eipiatinn Oe)'no)Ill) t s E)jua in)1

- Fliiiil I'vlcLln)i and lleail l.oss - Fliild Frlctlnn anil lleail I.nss

Ili ' IIIPD Therooolynami cs/Fluid Dynamics/Ilea t Trans fer (Cant) Tlieoondynamics/Fluid Dynamics/Ileat Transfer (Co<li) .

- Compressible Flaw - Elevation Ilead

- Incooqiressib'le Isentropic Flow - Pump and Systeo Characteristics

- Real Flovi Prableo~s - Twn Phase FInw

- Peop Characteristics

- Twn Phase Flow

- lnstrlooentaLIon - Hethods of Ileat Transfer

- Hetl)ods of Ileat Transfer - Boiling Iteat Transfer

- Specific I!eat, Expansion, Viscosity - Ileat Exchangers

- Viscous Flow

- Cniobined Conduct;ion/Convection

- IIuci cate/Fl los Ooil ing

- Critical IleaL Flux Electrical Sciences Electrical Sciences

- 4160 Volt Electrical Distribution - Electronics

- Protective Belaying for Generators - Hntnrs, Generators, Trapsforu)ers, Switchgear

- 600 VAC, 120 NC, and 250.-VDC Electrical - Instruoientatton and Control Theory D I s tribul, I on

- 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 IlucI ear Instrmoentation and Control

- Excore Iluclear Instnooentatlan Systeo) - Bad I a ti on Detectors

- lncore IIucleal Instriooentation System - Aeac tor lns tnaoen ta t I an

- Reactivity Control and Feedback A/P and Ileal t,h Physics linc)ear- Aadiatinn Protection and Ileal th Physics

- Si te Emergency Plan and Io)ple~oenting I'rocedures - Diologtcai Effects

- AppIicabie Radiation Protection, Concepts - Aadlat tan Survey Instrumentatinn Contained In 10 CFfl 20 and 10 CFB 100

- Aa~llologlcal Control Instructions - Shielding

-,Aa<ltatian Hant t.nring System

- I'nrtable Itaitlatton Hanitorlng Instrumentatinns Plant. Speci f ic Reactor Tectu>alogy Pla>>t Specific Reactor Technology

- AeacLor Cnre System ' (Including Core I'hysics Data)

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

~ ~ll E1

- Plant Spec)fic Reactor Teclinology (Cont} Plant Spec) f)c Reactor Techno]ogy Plant Cliemlca'I and Corrosion Control - Plant Chein)stry and Covvos)on Control

- Chemical and Yolunie Control and Ooron Hakeuli/Recovery Sys tern

- Pr)mary a>>d Seen<<davy Sampling Systein - Reactor Instrumentation and Control Plant Irist.rumentat lan and Control - Apactor, Plant )later)als

-See Item //6, Elect;rical Sc)ences, a<<d Item k'I>

Nucl ear Ins Lvlllnen ta t) on and Control Plant Ha ter lais

- See ILeni A, 'N<<clear Hetalurgy'lant Thennocycl e - Reactor Plant Thevniocycle

- Steam, Condensate, and Feed Systeins

- Steam Dump Systems

- Steam Generating anil Steam Generator Systenis Hanagement/Siipervisory management/Superv) sory Sk) 1 l s

- Lo be covered on a select:ed personal - Leadersh)p basis afLev initial STA Tra)n)ng Program - 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 Plant Systems

- Reactor Core System

- Reactor Coolant Systein - Reactov Coolant System

- Pressurizer and Pressure Ael)ef Systems

- Full Length Aod Cont,rol System - Reactor Control (liicluding Aoil Posit)on Indicat)on} I

- Chemical 5 Vol<<inc Control System and Boron - Reactor Coolant Inventory and Chemistry Control Hakeup arid Recovery Systenis

- Aes)dual lleat. Aenioval Syst: em - Aes)diial ileat Removal Systein

- Emergency Core Cooling System - Einergi.ncy Core Cool)ng System

- Einergency Cool lng )later

- Excore tlucleav Instr<<inc>>tat)on Systen - tliiclear Instrumentat)on

- I<<cove H<<clear Instr<<me<<tat)on Systein

- Reactor Pro tee L) on System - Reactor Protection Systein Conta)nnienL SysLem - Containnient System (including Conta)nment Cool)ng)

- Ice Condenser Syst: em Containine<<L Spray and Hydrogen Aecombiner System - Co<<talninentllydrogen tlonltor)ng and Contro'I

Indiana IlHiclrioan Electric ~IN 0 (Cont) ll- Pl ant Systeors Plant Systems 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 - plant ventilation auxll lary lruildirrg and control room ventilation sys tears eorevgerrcy diesel generator system - emergency electrical power auxiliary feerlwater systeor - auxiliary feedwater system corupressed air sysLem - eorergency control air 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 - radiation oronitoring system portal)le va(liat loll ololli I ori rig 'lns tvrlolents steaor generating and stearo generator water level - steam generator level control control sys teors steaor, condensate and feed.systeors - main steam, condensate, and feedwater steam duorp systeor 4160 KV electrical distribution system 600 VRC, 120 VRC, and 250 VDC electrical distribution system NOTE: see incore instrumentation and reactor - non nuclear instrrooentatlon coolant systeors loose part oronitoring status monitoring {including computer)

- seismic monitoring

)2 - hdmirrlstr.ative Control hr!orirristrative Control

.- Beslronsibilities fov Safe Operation 5 Shrrtdown - Responsibilities for Safe Opevation 8 Shrrtdown

- Equiproent Outages and Clearance Procedures Eqrllploellt OuLages and Clearance Procedures

- Use of Procedures - Use of I'rocedures

- Plant Hodi fications - Plant lhdifications

- Shi ft !Iel lef Turnovev and Hannlng - Slrift IIelief Tuvnover and Hanning

- Contairrment Access - Contairooent Access

- Haintainirrg Cognizance of Plant Status - Haintaining Cognizance of 'Plant Status

- Unit Int,erface Controls - Unit Interface Corrtrols

- Plrysical Secrrr'ity - I'lryslcal Secrrvlty

- ConLvol lhrour Recess - Corrtvol Honor Recess

- I)rrLIr.s a>>d Iles!>nrrsilrllities of Lire STA lhrL!es arrrl Aes!rorrsibilities of tire STA

- IIadlologica1 Eoiergerrcy I'lan - IIa<llological furer gency Plarr

- Co(ie of fe(ler'rl IIe<jrrlatiorrs (air!rvoirviate sections) 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 Transient and Accident Analysis for Shift Technical Advisors with reference to the I:SAII Technical Advisors with reference to the FSAII and Plant Abnormal and Emergency Procedures and Plant Abnormal a>>d Emergency Procedures 14 - tlormal, Transient; and Emergency Normal, Transient, and Emergency Opera tions (Simula tor Training) Operations (Simula tor Training) 15 - General Operatj>>g Procedures General Operating Procedures

- tiOTE; covered in systems and reactor Startup theory lectures - 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. 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.

2. Plant shutdown

3. 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.

7. Loss of electrical power (and/or dearaded power sources).

8. Loss of core coolant flow/natural circulation.

9. Loss of condenser vacuum.

10. Loss of Essential Servico Mater.

<<11. Loss of shutdown cooling.

12. 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.

20. Tur'ne or generator trip.

21. i~lfunction of Automatic Contml System(s) which a=fee.

reacti vi s.g.

22- ~'hl func ion of Reac.or Coolant Prwsure/Solus Con rol System.

"23. Reactor Trio.

24. iNain s.earn line break (inside or outside containment).

25. Nuclear Instrumentation failure{s).

• Required at leas anaual'i'y.

Even if theeachabove manipulations are not ne ded to be shi i technical advisor shall attend accoaqlished by a simulator, 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 I I. 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.

TABLE II.K.3 - 1 0

E Date/Time Ounatian Eauiament jieeean 1-18-75 .Initial Criticality 2-25-75 57 hrs "AB" Diesel Engine Lubrication oil piping modifications 3-3-75 62 hrs "CD" Diesel Engine Lube oil piping modifications

. 1-13-75 36 hrs* RHR System leak 4-7-77 57 hrs Mest Centrifugal Charging Pump Pump shaft severed 5-17-77 67 hrs East Centrifugal Charging Pump Broken shaf 8-9-77 36 hrs* Eas Centrifugal Charging Pump leak repair 8-11-77 9 hrs "AB" Diesel Generator wrist pin replacement 8-30-77 36 hrs 'Rest RHR motor modification 9-12-77 11 hrs "CD" Diesel wrist pin'repIac~ent 9 14-77 36 hrs* "CD" Diesel maintenance 2-6-78 5 hrs "AB" Emergency Diesel tach. modifies ion 2-7-78 3.3 hrs "CD" Emergency Diesel tach. modification 9-7-78 3.5 hrs North SI Pump maintenance 9-7-78 3 hrs South SI Pump maintenance 10-3-78 15 hrs Eas" CTS Pump valve repair L'-7-78 4.25 hrs "AB" Emergency Diesel replace voltage regulator 11-10-78 8 hrs "CD" Emergency Diesel maintenanc 1'18-78 5 hrs "AB" Emergency Diesel maintenanc LL-25-78 8 hrs East Centrifugal Charging Pumo oil leak 12 hrs "ast RHR Systan clearance on It~0-312 11-30-78'2-6-78 II S.5 hrs le CD emergency Diesel design change 1-17-79 36 hrs>> Mest RHR Train repair RHLZBM 1-26-79 9.5 hrs IIABIe Emergency Diesel Generator design change 1-30-79 33.5 hrs East Centrifugal Charging Pump, seal railure 2-15-79 East Centrifugal Charging Pump maintenance 2-'-79 5.7 hrs 'Rest Centrifugal Charginc P mp chancre oiI straine, 2-18-79 3.4 hrs Nest Centrirugal Charging Pump repair gPIZ57 isolation vlv.

UNIT 1 4

TABLE II.K.3 - 1 Date/Time Eaui omen t Reason 2/27-2/28/79 2.2 hrs East and Mes RHR Pumps brkr inspection 2/27-2/28/79 2.5 hrs North and South SI Pumps brkr inspec.ion 2/27-2/28/79 1.2 hrs East and Mest CTS Pump brkr cleaning 2-23-79 7.6 hrs "AB" Diesel maintenance on ESM check vlvs.

3-(1-7)-79 2.2 hrs Mest Centrifugal Charginc Pump clean brkr.

3-l-79 11.2 hrs "CD" Diesel CD-work on ESM ck.vlv.

3-3-79 6.0 nrs "AB" Diesel AB-work on ESM ck.rlv.

3>>12-79 36 hrs~ "CD" Emergency Diesel Gene. ator maint nance 3-16-79 36 hrs* "AB" Emercency Diesel Gene~ tor maintenanc 4-(5-10)-79 36 hrs* "AB" Emergency Diesel ma ntenance

'-23-79 5 hrs North SI Pump repack ICN-250 8-1-79 2.6 hrs Mes Centrirucal Charg ng Pumo maintenance 9-24-79 4.5 hrs South SI Pump maintenance 9-25-79 1 hr North SI Pump maintenance 9-25-79 East RHR Pump malnteflaflc 9-28-79 2 hrs Mes RHR Pump maintenance 1-30-80 6.5 hrs "AB" Diesel Per. ormanc .as.- nc

'-31-80 3.4 hrs "CD" Diesel replace relays - RFC 2-26-80 6 hrs "AB" Emergency Diesel maintenance 3-5-80 1.5 hrs East Centrifugal Charging Pumo maintenance clean <<CC 3-6-80 30.75 hrs East Centrifugal Charging ?ump change rotating assembly 3-26-80 5.2 hrs East Centrifugal Charging Puma clean CONO filter on lube oil system 3-30-80 3.5 hrs East Containment Spray Pump maintenance 3-21-80 4.1 hrs South Safety Injection Pump clean brkr.

3-24>>80 2.1 hrs Mest RHR Pump clean brkr.

2.7 hl's Mest Containment Spray P mo clean brkr.

1.6 hrs Mest Centrifugal Charcirc Pumo clean brkr.

'Hii 1 TABLE II.K.3 - 1 Oaae/Tfme Onnaai on Eouicment Reaaan 3 25>>80 2 hrs East Centrifugal Charging Pumo clean brkr.

3-26>>80 9.0 hrs "CD" Emergency Diesel Generator clean output brkrs.

3-27-80 9.3 hrs "AB" Emergency Diesel clean output brkrs.

4-l9%0 1.75 hrs East Centrifugal Charging Pump change oil

UNIT 2.

TABLE II.K.3 - 2 Oate/Time Ouratinn Eoui cment 3-10-78 Initial Critical i.y Reaaen'rroneously 4-16-78 10 hrs. Mest RHR Train decIar&

inoperabl e (RH-104M Pos. Indication at mid-position.

4 5-15-78 "CD" Emergency Diesel repair water leak on Air Cooler.

7-11-78 11.7 hrs "AS" Emergency Diesel replace ESM Supply Spool Piece 7-18-78 9.5 hrs "CD" Emergency Diesel repair fuel in<ec ion 9-11-78 8.7 hrs "AB" Emergency Diesel Starting Air Ck. . Vlv.

leaking through 9-12-78 49.5 hrs "CD" Emergency Diesel fuel oil le ks 9-19-78 2 hrs Mest RHR Pumo Repair brkr. Ieakace 10-3-78 36 hrs* East CTS Train Valve ?roblem could not verify flow .hrougn CTS-:20E - Broken sha-,t ji"I0-27.

10-11-78 16 hrs "CD" Emergency Diesel routine maintenance 10-lo-78 Oe5 hrs "AB" Diesel Failed to s art ~//ice.

10-25-78 1.75 hrs CD':nercencj Relay Calibration 11-17-78 18.9 hrs "AB" Emergency Diesel Generator routire maintenance 17 79 4.2 hrs E RHR Pum iubricat on 1-11-79 2.1 hrs M RHR Pump lubrication 1-12-79 8.3 hrs East Centrifugal Charing Pump oil change 1-24-79 17.3 hrs "AB" Diesel Generator Repair ESM Supply VIv.

1-31-79 7.7 hrs "AB" Emergency Diesel Generator Repair ="

M-143 4-23-79 21 hrs "CD" Emergency Diesel Cenerator Repair ESM Supolv ck. viv.

7 1879 7.5 hrs "CD" Emergency Diesel Generator Repair leaking Ck. Vlv.

8-27-79 8 min "AB" Emergency Diesel Testing Co2 o Emer.

Diesel Rooms.

8-27>>/9 8 min. "CD" Eme. gency Oiesel t esti ng Co i o Diesel Roon s

e TABLE II.K.3'- 2 Gate/Time Ouration Eoui oment Reason 9-26-79 36 hrs* East RHR Pump routine maintenance 9-29-79 7 hrs Mest Centrifugal Charging Pump repack coupling 10-15-79 9.1 hrs "CQ" Emergency Oiesel Generator routine inspection 10-16-79 11 hrs "AB" Emergency Oiese'1 Generator routine inspection

~ 1-18-80 8 hrs East Containment Spray Pump remove suction s raine

.1-21-80 13 hrs Mest Containment Spray Pump remove suction s rainer 1.-30-80 7.2 hrs North SI Pumo reprove suction s.rainer 1-31-80 7.3 hrs South SI Pump remove suc.ion straine.

2-1-80 6 hrs Mest Centrifugal Charging Pump check suction straine.

2-4-80 2.0 hrs North SI Pump Repair SV-98 on Oischarge He l'er 2A-80 12.5 hrs East RHR Heat Exchanger Repair 'SV-104 2-11-80 39.2 hrs Mes RHR Pump 5 Heat Exchanger Repack IHO-324 2-14-80 9.2 hrs Mest RHR Pump Repack ISO-322 2-28-80 7.9 hrs East Centrifugal Charging Pump remove suc.ion strainer 3-17-80 7.1 hrs Mest CentrifugaI Charging Pump Fitting leak oil supply line 4-1-80 10 hrs Mest Centrifugal Charging Pump Replac gasket cn suction strainer

~10-80 57.5 hrs Mest Centrifugal Charging Pump Seal Failure 4-15-80 5.9 hrs Mest Centri fugaI Charging Pump Repair leak on CCM, -.o He t Exch.

4-16-80 1.5 hrs Mest Centrifugal Charging Pump Output brkr. maint.