IR 05000315/1981003

IE Insp Repts 50-315/81-03 & 50-316/81-03 on 810201-28. Noncompliance Noted:Failure to Submit Rept Required by Tech Specs & Inadequate Review & Maint of Operating Procedures

ML17331A734
Person / Time
Site:	Cook
Issue date:	05/15/1981
From:	Dubry N, Hayes D, Swanson E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML17326A900	List: IR 05000315/1981003 ML17326A899 ML17326A901
References
50-315-81-03, 50-315-81-3, 50-316-81-03, 50-316-81-3, NUDOCS 8106020541
Download: ML17331A734 (38)

Text

U.S.

NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT

REGION III

Report Nos. 50-315/81-03; 50-316/81-03 Docket Nos. 50-315; 50-316 License Nos.

DPR-58; DPR-74 Iicensee:

American Electric Power Service Corporation Indiana and Michigan Power Company 2 Broadway New York, NY 10004 Facility Name:

D.

C.

Cook Nuclear Plant - Units 1 and

Inspection At:

D.

C.

Cook Site, Bridgman, MI Inspection Conducted:

j Inspectors:

February 1-28, 1981

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Swanso Reactor Projects Section 1B Ins ection Summa Ins ection on Februa 1-28 1981 (Re ort Nos. 50-315/81-03.

50-316/81-03)

Areas Ins ected:

Routine, onsite regular and backshift inspection by the resident inspectors.

Areas inspected included Followup on Previous Inspection Findings, Operational Safety Verification, Maintenance Observa-tion, Surveillance Observation, Licensee Event Report Reviews, IE Bulletin Followup, IE Circular Followup, Onsite Review Committee Activities, Independent Inspection,,TMI Action Plan Followup, and Procedures Review.

The inspection involved a total of 158 inspector-hours onsite by two NRC inspectors, including 31 inspector-hours on the off-shifts.

Results:

Of the eleven areas inspected, no items of noncompliance or deviations were identified in eight areas.

Three items of noncompliance were identified in three areas; inadequate

CFR 50.59 reviews, failure to submit report as required by Technical Specifications, and inadequate review and maintenance of operating procedures.

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DETAILS Persons Contacted

• D. Shaller, Plant Manager-B. Svensson, Assistant Plant Manager

+E. Townley, Assistant Plant Manager

• R. Begor, Staff Assistant R. Keith, Operations Superintendent E. Smarella, Technical Superintendent R. Dudding, Maintenance Superintendent J. Stietzel, gA Supervisor The inspectors also conducted a number of interviews with operators, technicians, and maintenance personnel during the inspection.

• Denotes those present during the exit interview.

Followu on Previous Ins ection Findin s II (Closed)

Noncompliance (50-315/80-20-01)

Charging pump 1W not run for full 15 minutes required by Technical Specification 4.1.2.3.C:

The inspector verified that applicable Surveillance Test procedures now include the 15 minute acceptance criteria as described by the licensee's response dated February 23, 1981 (AEP:NRC 0516).

(Open) Noncompliance (50-315/80-19-01)

Fire barrier seals inoperable:

Immediate corrective action was found to be acceptable, however, the actions taken to prevent further recurrence appears to be inadequate based on the recent occurrence of a similar event on February 12, 1981, as detailed by the licensee's Condition Report 1-2<<81-055.

The inspector noted that though the Plant Manager Instruction was revised the bulk of plant employees were not instructed in their new respons-ibility.

(Closed)

Noncompliance (50-316/80-14-01)

Failure to report an event:

Entering action statement on ECCS accumulator was not reported.

The licensee's response of January 23, 1981, appears to adequately address the actions taken to prevent recurrence of a similar type.

The licensee's event report on the subject event, 80-39/03L-0 of March 4, 1981, satisfies the reporting requirements.

(Closed)

Noncompliance (50-315/80-11-01, 50-316/80-09-01)

Failure to complete a training review of design changes:

Although the licensee's response dated August 4, 1980 (AEP:NRC00448) detailed no procedural changes that would prevent the operator reviews from "falling through the cracks", it does appear to have been an isolated occurrenc i

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erational Safet Verification The inspector observed control room operations, reviewed applicable logs and conducted discussions with control room operators during the month of February.

The inspector verified the operability of selected emergency systems, reviewed tagout records and verified proper return to service of affected components.

Tours of the auxiliary building and turbine building conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, and excessive vibrations and to verify that maintenance requests had been initiated for equipment in need of maintenance.

The inspector by observation and direct inter-view verified that the physical security plan was being implemented in accordance with the station security plan.

The inspector observed plant housekeeping/cleanliness conditions and verified implementation of radiation protection controls.

The inspector also witnessed portions of the radioactive waste system controls associated with radwaste shipments and barreling.

These reviews and observations were conducted to verify that facility operations we e in conformance with the requirements-established under

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technical specifications, 10 CFR, and administrative procedures.

4.

Monthl Maintenance Observation Station maintenance activities of safety related systems and components listed below were observed/reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides and industry codes or standards and in conformance with technical specifications.

The following items were considered during this review:

the limiting conditions for operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable; functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used, were properly certified; radiological controls were implemented; and, fire prevention controls were implemented.

Work requests were reviewed to determine status of outstanding jobs and to assure that priority is assigned to safety related equipment maintenance which may affect system performance.

The following maintenance activities were observed/reviewed:

Reciprocating Charging Pump, Unit 1 Reactor Coolant Filter Change, Unit 1 Check Valve RH-108E, Unit 2

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No actual maintenance work took place on check valve RH"108, but the inspector verified that the affected train of the Residual Heat Removal System was returned to service properly.

Monthl Surveillance Observation The inspector observed technical specifications required surveillance testing on the determination of Moderator Temperature Coefficient (1-THP.4030.STP307)

and verified that testing was performed in accordance with adequate procedures, that test instrumentation was calibrated, that limiting conditions for operation were met, that removal and restoration of the affected components were accomplished, that, test results conformed with technical specifications and procedure requirements and were reviewed by personnel other" than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel.

The inspector also witnessed portions of the following test activities:

2-THP.4030.STP162, Gland Exhaust Vent Process Monitor (R-33);

-2-HP.4030.STPK63, Condenser Air.Ejecter"Gaseous-Process Min'tor (R-15);

and 2-THP.4030.STP181, Steam Generator Blowdown Treatment Liquid Process Minitor (R-24).

I,icensee Event Re orts Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to deter-mine that reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with technical specifications, 50-315/81-01/03L-0 Boron Injection Tank Below Specified Concentration 50"316/80-35/03L-0 50-316/80-35/03X-1 IE Bulletin Followu Rod Position Indication for Rod N-9 Inoperable Supplement For the IE Bulletins listed below the inspector verified that the written response was within the time period stated in the bulletin, that the written response included the information required to be reported, that the written response included adequate corrective action commitments based on information presented in the bulletin and the licensee's response, that licensee management forwarded copies of the written response to the appropriate onsite management representa-tives, that information discussed in the licensee's written response was accurate, and that corrective action taken by the licensee was as described in the written response.

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IEB 78-13 Failure in Source Heads of Kay-Ray, Inc. Gauges.

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The licensee did not have any of the subject gauges, but did discover a similar gauge that had been stored in a warehouse since its use in 1977. " It is in the process of being returned to the licensee holder (manufacturer) for the contained byproduct material.

8.

IE Circular Followu For the IE Circulars listed below, the inspector verified that the Circular was received by the licensee management, that a review for applicability was performed, and that if the circular were applicable to the facility, appropriate corrective actions were taken or were scheduled to be taken and that a review for applicability was performed and that correctiv actions are being taken in addressing TMI-2 Task Action Plan Item III.D.3.4 "Control Room Habitability".

IEC 80-03 Protection From Toxic Gas Hazards The inspector noted that the control room filter train is not redundant and apparently does. notmeet,the.single failure criterion...

(Inspector Followup Item 81-03-01)

9.

Onsite Review Committee A previously identified unresolved item (50-315/81-01-01; 50-316/81-01-01)

concerned the quality of review given to (a) License Event Reports (LER's)

and (b) changes to procedures, tests and experiments with the documenta-tion of the bases for the

CFR 50.59 determination of, whether or not the change constituted an unreviewed safety question (US/).

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The inspector found that a subcommittee review group had been established to give LER's a more in-depth technical screening.

The inspector will follow the results of this subcommittee and its effect on the adequacy of reported events.

The inspector also noted that besides the Technical Specifications there are no administrative procedures or charter existing which governs the review activities.

The only existing charter was effective January 7,

1975, and is not acknowledged as being effective though it has not been cancelled or replaced.

The inspector was informed that a revised procedure ia being prepared.

b.

IE Circular 80-18 concerning

CFR 50.59 changes contains a

discussion which reflects the generally accepted interpretation of the Code. It goes on to state that "For all cases requiring a written safety evaluation, the safety evaluation must set forth the bases and criteria used to determine that the proposed change does not involve an "unreviewed safety question".

A simple statement of the conclusion in itself is not sufficient".

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In addition, the circular exemplifies the point that "equipment important to safety" does not imply that only designated

"safety-related" equipment is of concern, but also systems and equipment which can have an effect on a safety system's performance or have radiological impact.

The plant review committee's responsibilities in this are are con-tained in Technical Specification 6.5.1.7b which states:

"The Plant Nuclear Safety Review Committee (PNSRC) shall:

Render determinations in writing with regard to whether or not each item considered under 6.5.1.6(a)

through (e).

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constitutes an unreviewed safety question."

Paragraph 6.5.1.6 includes:

(1)

Review of (1) all procedures required by Specification 6.8 and changes thereto, (2) any other proposed procedures or changes thereto as determined by the Plant Manager to affect nuclear safety.

(2)

Review of all proposed tests and experiments that affect nuclear safety.

(3)

Review of all proposed changes to Appendix "A" Technical Specifications.

(4)

Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety.

(5)

Investigation of all violations of the.Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Chairman of the NSDRC.

The review method used by the PNSRC is to complete the

"PLANT NUCLEAR SAFETY COMMITTEE REVIEW CHECKLIST" which: (1) asks if the item under review is a change to the Technical Specifications (TS) or bases and, if so, requires a change to the Technical Specifications to be approved, (2) asks if the proposal is safety related as identified in Appendix "B" of the FSAR, if not, further review is done, if it is, then a written evaluation is required.

This checklist is most often completed without referencing the guidance on the back side (many copies available were printed on only one side)

and only the Yes-No blocks were checked, thus not providing any documented basis for making the US/ determination.

During the review of two "Emergency" design changes, examples of inadequate review were obvious.

A modification was made to the Unit 1 ice condenser (RFC-Ol-1S19)

when the PNSRC review had determined that it was not a safety related structure, nor did it impact on a safety-related system component, or structure as identified in Appendix B of the FSAR.

The inspector verified

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that the subject insulated door frames are seismic Class I according to Appendix B and that they are identified in the plant's

"N" list as being safety-related.

A second modification (RFC 12-1814)

was made relocating four main steam valves.

The review checklist indicates that the modification is safety-related and documents the following:

"AEPSC-NY to examine proposed relocation of valves for any seismic effects on previous calculations".

A phone call record further documents that a conver-sation had taken place between the plant and the AEPSC piping section describing the modification and that the corporate licensee repre-sentative

"granted permission to proceed with this work".

There was no further documentation to reflect this other than that there did exist an US/ at the time the modification was made in the area of seismic qualification.

Numerous special tests and procedure changes were conducted for which either no safety evaluation was documented or the basis for the 10 CFR 50.59 determination was not documented.

Examples include the maintenance procedure for installing mechancal plugs in st am generator-tubes (HHP-SP-.011),-the-operat.ons procedu" for steam generator tube leak rate tests (OHP-SP.005),

the 48-hour endurance test of AFW pumps (THP-4030.SP.017),

and the Hydrogen Recombiner Functional Test (01.0HP-4030.SP.013).

The above items constitute noncompliance with the referenced section of the Technical Specifications and with 10 CFR 50.59 as set forth in Appendix A, Notice of Violation.

(81-03-02)

10.

Inde cadent Ins ection The inspectors reviewed Special Report Nos. SI-15 and SI-16.

SI-15 dated February 6, 1981, was noted as being overdue on February 4, 1981, by the inspectors.

It was determined that the licensee's equality Assurance Organization had provided a reminder of pending due reports but management is apparently not making good use of this tickler system.

The subject report (SI-15) was required to be submitted within 90 days of. the event which occurred on October 11, 1980, by Technical Specification 3.5.2 Action Statement 6:

"In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date."

The events were described in IE Inspection Reports 50-315/81-01 and 50-316/81-01.

The failure to'eport constitutes a violation of the above Specification and is detailed in Appendix A to the transmittal letter of this report.

(81-03-03)

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11.

TMI Action Plan Re uirements I.A.II Shift Technical Advisor (STA):

The inspector reviewed the licensee's training program, for STA's and found it to accur-ately describe the trairiing that was conducted by the Plant Training Department, Purdue University, Westinghouse, and General Physics.

The licensee has assigned a degreed engineer who has completed the 'initial training program to each shift and specified their responsibilities in various administrative directives.

During the review of initial training given to STA's, the inspector noted a lack of established criteria for determining that the STA's were retaining an acceptable level of knowledge.

In fact, several of the trainees received less than acceptable quiz grades with the most prevalent weak areas being Heat Transfer, Fluid Flow, and the Auxiliary Feedwater System.

Criteria has been established in the Requalification Program, and although not required, the inspectors feel that established criteria should have been required for the initial qualificati>n program.

~ In sugary,"the licensee.has put-forth-a commendabl

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effort in training, the STA's, but their training records appear to reflect weaknesses in the very areas where expertise was needed by the operating organization.

I.A.1.3 Shift Mannin

- Limit Overtime:

The inspectors reviewed the plant manager's Standing Order No.

.054 concerning overtime, the licensee's November 7, 1980 letter (AEP:NRC0450)

and the Division of Licensing's response dated January 21, 1981.

The inspector found that the licensee intends to reserve the right to exceed the recommended overtime limits through cognitive, documented management decisions in specific situations.

The licensee intends to be adequately staffed to meet the July 1, 1982 requirements without scheduled overtime.

The administra-tive guidance concerning limiting overtime was found to be adequate.

I.A.1.3 raded Trainin Pro ram and ualification for Reactor 0 erators and Senior Reactor erators:

The inspector verified that the revised training program submitted, July 25, 1980, has been implemented and utilized for one group of candidates for license requalification and a second group is due to be examined on March 9.

I.C.5 Procedure for eratin E

erience Feedback:

The Plant Manager has implemented his Standing Order Nos.

.052 and

.057 entitled

"Control of NRC Documents" and "Information Evaluation of Non-NRC Documents",

respectively.

These, along with the inte" gration of the Shift Technical Advisor into various related plant activities, including review of plant originated Condition

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Reports, appear to adequately address the Action Plan position for this item, although the administrative controls are somewhat vague.

I.C.6 Verification of eratin Activities:

As described in the licensee's January 8,

1981, response to the plant procedures governing equipment control during surveillance, maintenance and repair activities were reviewed and modified'to include the recommended verification activities.

The inspectors found that the supervisory operators are kept adequately informed of the status of equipment important to safety.

This item is considered closed.

II.E.1.2 Auxilia Feedwater S stem (AFW)

Part 1 - Automatic Initiation of AFW:

The inspector reviewed the licensee's January 8,

1981, response to this item (AEP:NRC:00398)

and several previous responses pertaining to this subject and detemined that, with the exception of the loss of main feedwater pumps, initiating signals, the auto-initiation -feature of AFW appears -to have been-constructed-to..- --. ---

meet safety-grad criteria.

The inspector was not able to verify in 'detail that this item meets the criteria of NUREG 0737 based on the licensee's response.

Part 2 - AFW Flow Indication:

NUREG-0737 clarification on this item requires as a minimum AFW flow rate indicator and

'ne wide-range steam generator level indicator for each steam generator or two flow rate indicators.

The licensee's January 8,

1981 response states that environmentally qualified feedwater flow rate transmitters will be installed to meet the requirements of this item.

The licensee apparently feels that the use of a narrow range steam generator water level indicator is more conservative as an operational restriction.

As with Part 1 of the licensee's response to this part was not of adequate detail to allow the inspector to determine accept-ability.

II.K.3.9 Both parts of this item are under review by a Nuclear Regulatory Commission contractor who has been provided the requested information to verify the adequacy of the licensee's design.

Pro ortional Inte ral Derivative (PID) Controller Modification:

The PID controller was modified in October, 1974, for Unit 1 and August, 1977, for Unit 2 to disable that portion of the controller which permitted the PORV's to operate automatic-ally below either their nominal setpoint or RCS operating pressure.

Also, the interlock bistable pressure setpoint was raised to 2350 in May, 1979, on both Units 1 and III.D.3.3 I roved In lant Iodine Instrumentation:

The equipment and procedures described in the licensee's response to this item (AEP:NRC0398 and 0398B of January 8 and 16, 1981) were found to be in place.

This equipment appears to meet the position described for this item.

12.

Procedure Review The inspectors conducted an examination to verify that the review and approval of station procedures were in accord and satisfied technical specifications.

This review included selected Administrative, General Plant Operating, Procedures,,

Startup, Operation, and Shutdown, Abnormal Conditions, Emergency, and Surveillance Procedures as found in the Master Vault'..

This inspection was done to verify procedures received a timely review; that temporary changes were proper and did not conflict with Technical Specifications, and that procedure changes were made to reflect license and Technical Specifications revisions.

Procedure changes were checked for conformance with 10 CFR 50.59(a)

and if applicable that the basis was documented and records retained.

Plant working files were also checked to verify procedures were current I

a.

Administrative Procedures Administrative procedures were reviewed for accord with Technical Specification 6.8.2.

Included was PMI 2010, "Plant Manager and Department Head Instructions, Procedures, and Associated Indexes Procedures were reviewed for:

required format, temporary change routings, cancellation requirements, distribution, and revision criteria and frequency requirements.

b.

erations De artment Procedures The inspectors reviewed the following Operations Department Proce-dures for agreement with regulations,. license conditions, Technical Specifications, ANSI Standards, Regulatory Guides, and administrative instructions committed by the licensee.

(1).

General Plant eration - Procedures 1-OHP"4021-001.001 (Rev. 8)

"Plant Heatup from Cold Shutdown to Hot Standby" 1-0HP-4022.001.001 (Rev. 3)

"Emergency Shutdown Including a Reactor Trip" Referred to:

1-0HP-4021-001.005 (Cancelled) referred to 1"OHP-4021.001.006 (Rev. 3) "Power Operation".

1-0HP-4021.001.006 (Rev.

1)

"Plant Shutdown from Minimum Load to Startup or Hot Standby Mode".

Items of nonconformance for above included:

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(a)

Failure to follow cancellation requirements, Procedure 1-OHP"4021.001.005.

(b)

Lack of,continuity in referencing sequence, Procedure 1-0HP-4022.001.001, Rev. 3.

(c)

Omission of procedures, Procedure 1-0HP-4022.001.005 Rev. 3.

(d)

Failure to conduct timely procedure reviews, Procedures 1-0HP-4022-001.001, Rev. 3, and 1-0HP-4021-001.003, Rev.

1.

(2)

Startu eration and Shutdown Procedures l-OHP-4021.002.001 (Rev. 7)

"Filling and Venting the Reactor Coolant System" 1-OHP-4021.002.003 (Rev.

4)

"Reactor Coolant Pump Operation" l-OHP-4021.002.005 (Rev. 3)

"Draining Reactor Coolant System" l-OHP-4021.002.006 (Rev. 3)

"Pressurizer Relief Tank Operation".

1-0HP-4021.028.011 (Rev.') "'"AuxiliaryBu'i1ding Ventialtion".

l-OHP-4021.051.001 (Rev. 2)

"Placing Steam Generators and Main Steam System in Service".

1-PHP-4021.051.003 (Rev.

3)

"To Remove From Service and Return to Service "A" Steam Generator Stop Valve, Dump Valve after Maintenance".

1-OHP-4021.051.005 (Rev. 0)

"Returning from and Returning to Service of the Moisture Separator Reheaters with the Unit Operating Between 50 and 100$ Power".

2-0HP-4021.008.001 (Rev.

1)

"Filling and Venting of the ECCS".

Referred to:

2-0HP-4021.007.002 and 2-0HP-4021.008.002 (Rev. 5)

"Placing ECCS in Standby Readiness".

2-0HP-4021.008.003 (Rev.

1)

"Charging and/or Decreasing Boran Concentration of Accumulators".

2-0HP-4021.019.001 (Rev. 0)

"Cancelled".

Referred to:

12-0HP-4021.019.001 (Rev.

4) 'illing, Venting and Placing in Service of the ESW System".

2"OHP-4021.020.001 (Rev.

2)

"Filling and Venting of NESW System".

2-0HP-4021.020.002 (Rev.

1)

"Operation of NESW During Normal Plant Operations".

Of the eleven Unit 1 electrical system procedures, l-OHP-4021.082.

.001 to.013; and ten Unit 2 electrical procedures, 2<<0HP--4021.

082.001 to.012; only l-OHP-4021.082.007 and 2-0HP-4021.082.

007, "Placing Batteries on Equalizing Charge" were current.

The remaining nineteen procedures were overdue for review.

Items of nonconformance for the above include:

(a)

Omission of continuity in referencing sequence, Procedure 2-0HP-4021.019.001, Rev.

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Incomplete distribution, procedures 1-0HP-4021.051.005, Rev. 0, and 2-0HP-4021.008.003, Rev. l.

(c)

Failure to conduct timely procedure reviews; Procedures 1-'OHP-4021.028.011, Rev. 2, l-OHP-4021.051.00,1 Rev. 2, 2-0HP-4021.020.002, Rev.

1, l-OHP-4021.082.001 thru.006 and

.008 thru.013, and 2-0HP-4021.082.001 thru.006 and

.008 thru.013, 1-0NP-4021.051, Rev. 0, 2-0HP-4021.

008.001, Rev.

1, 2-0HP-4021.019.001, Rev. 0.

(3)

Abnormal Procedures 1-0HP-4022.001.001 (See Previous Discussion)

1-0HP-4022.002.001 (Rev.

1)

"Malfunction of RCP,"

Referred to 1-0HP-4022.001.001 2-0HP<<4022.002.002

"Loss of RCP" (Cancelled)

2-0HP-4022.002.003

"Removal of a RCP from Service" (Cancelled)

1"OHP-4022.002.004 (Rev.

3) "Excessive Reactor Coolant Leakage" Referred to:

1-0HP-4022.016.003

"High Activity in CCW System" 1-0HP-4022.002.007 (Cancelled) refers to 1-0HP-4023.001.009 1-0HP-4022,002.008 (Cancelled)

refers.ta 1-.0HP-.4023.001.009..

,...

2-OHP-4022.001.001 (Rev.

1) "Emergency Shutdown Including Reactor Trip" Referred to 2-0HP-4021.001.005 2-0HP-4022.008.001 (Cancelled)

Replaced by 2-0HP-4023.001.002 2-OHP"4022.008.002 (Rev.

4) "Initiation of ECCS Recirculation Phase" Referred to 2-0HP-4022.009.001 2-0HP.-4022.008.003 (Rev. 2) "Termination of SI" Referred to:

2-0HP-4021.019.001 (See previous discussion),

2-0HP-4021.016.003 "Fill and Vent CCW" 2-OHP-4022.034.003

"Recovery from Spurious Containment Isolation, Phase

"A" 2-0HP-4022.008.004(.005)

"Initation of ECC Recirculation Phase with

"E" ("W") RHR pump out of service Referred to 2-0HP-4022.009.001 2-0HP-4022.008.006 (Rev. 2) "Initiation of ECC Recirculation with one entire Safety Train out of Service";

Referred to 2-0HP-4022.009.001 2-0HP-4022.019.002 (Rev.

2) "Operation of ESW System Following a IOCA" 2-0HP-4022.020.002 (Rev. 0) "Operation of NESW System Following a LOCA" Items of nonconformance identified in this area were:

(a)

Ommission of continuity in the referencing sequency, Proce-dure 2-0HP-4021.019.001.

,(b)

Failure to conduct timely procedure reviews; Procedures 1-0HP-4022.001.001, 1-0HP-4022.002.001, Rev.

1, 2-0HP-4022.001.001, Rev.

1, 2-0HP-4022.019.002, Rev. 2, and 2-0HP-4022.020.002, Rev.

0.

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(c)

Referencing non-existant procedures in the train of actions to be taken, Procedure 2-0HP-4022.001.001, Rev.

1.

(4)

Emer enc Procedures 1-OHP-4023.001.001

"Alternate Emergency Shutdown and Cooldown Procedure due to Loss of Normal and Preferred Alternate Method".

Referred to:

1-0HP-4022.001.001 (See previous discussion)

1-0HP-4021.017.002

"Starting RHR" 1-0HP-4021.005.004

"Normal Boration" l-OHP-4023.001.002 (Rev.

1)

"Emergency Procedures, Immediate Actions and Inadequate Diagnostics" Referred to:

1-0HP-4023.001.009

"Loss of Reactor Coolant" 1-0HP-4022.002.004

"Excessive Reactor Coolant Ieakage" 1-0HP-4022.008.003

"Termination of Safety Injection" 1-OHP"4023-001.011 (Rev.

3)

"Reactor Shutdown from Hot Standby Panel Ducts Control Room Inaccessability" Referred to:

1-0HP-4023.001.001 OHP-4023.001.012 (Rev.

0)

"Earthquake" Referred to:

- OHP-4021.032:001.."Test'and Iead -Deisel PMP-2080-EPP

~ 008 "Natural Emergencies" OHP-4023.001.013 (Rev.

0)

"Seiche" PMP 2080 EPP.008 (Rev.

1)

"Natural Emergencies" Referred to:

OHP-4023.001.013 and OHP-4023.011.012 l-OHP-4023.016.001 (Rev.

1) "Loss of CCW" Referred to:

1-0HP-4023.017.001

"Lossof RHR" 1-0HP-4023.019.001

"Loss of ESW" l-OHP-4023.019.001 (Rev.

0)

"Loss of ESW" l-OHP-4023.020.001 (Rev. 0)

"Loss of NESW" 1-0HP-4023.053.001 (Rev. 0)

"Ioss of Condenser Vaccuum" Referred to:

1-0HP-4022.053.001

"Partial Loss of Vacuum" 1-0HP-4022.001.001 1-0HP-4021.001.003 to 1-OHP"4021.001.004 2-0HP-4023.001.001 (Rev.

1)

"Alternate Emergency Shutdown Proce-dure due to Loss of Normal and Preferred Alternate Methods" Referred to:

2-0HP-4022.001.001 2-0HP-4021.017.002 2"OHP-4021.005.004 2-0HP-4023.001.002 (Rev.

1)

"Emergency Procedures

- Immediate Actions and Diagnostics" Referred to:

2-0HP-4023.001.004

- Steam Generator Tube Rupture 2-0HP-4023.001.005

- Steam Line Break 2-0HP.4023.001.009

- Loss of Reactor Coolant 2-0HP>>4022.002.004

- Excessive Reactor Coolant Leakage 2-0HP-4022.008.003

- Terminate Safety Injection" 2-0HP>>4023.001.011 (Rev.

1) "Reactor Shutdown from Hot Shutdown Panel to Control Room Inaccessibility" 2-0HP-4023.016.001 (Rev.

0)

"Loss of CCW" Referred to:

2-0HP-4023.017.001

- "Loss of RHR Cooldown" 2-0HP-4023.019.001 (Rev. 0) - "Loss of ESW" 13-

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Ahgp t

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= ~

2-0HP-4023.020.001 (Rev. 0) "Ioss of NESW" 2-0HP-4023.001 (Rev.

1) "Loss of Condenser Vacuum" Referred to:

2-0HP-4022.001.001 2-0HP-4021.001.003 2-0HP-4022.053.001 Nonconformance items in this area consisted of:

(a)

(b)

(c)

Referrals to incorrect on nonexistent procedures; procedures PMP2080EPP.008, Rev.

1, and 1-0HP-4023.053.001, Rev.

0.

Inproper distribution, procedure 1-0HP-4023.053.001, Rev. 0.

Failure to have working copies in the control room, procedure l-OHP-4023.053.001, Rev. 0.

(d)

(5)

Lack of timely procedure 1-0HP-4023.001.012, Rev.

1-0HP-4023.016.001, Rev.

1-0HP-4023.020.001, Rev.

2-Oiii-'4023. 001; 002,

'ev.'-0HP-4023.019.001,Rev.

2-0HP-4023.053.001, Rev.

Surveillance Procedures reviews; procedures 1-0HP-4023.001.001, 0, OHP-4023.001.013, Rev.0, 1, 1-0HP-4023.019.001, Rev.

0, 0, 1-0HP-4023.053.001, Rev. 0,

'1, 2-OHP-4023 01'6.001, Rev. '0,

',

2-0HP-4023.020.001, Rev. 0, and l.

12-0HP-4030.001.001 (Rev.

4) "Routine Plant Inspection Outside of Control Room" 1-0HP-4030.001.002 (Rev. 2) "Routine Containment Inspection" 1-0HP-4030.STP.006 (Cancelled)

"ESW Radiation Monitoring" 1-OHP-4030"STP.012 (Cancelled)

referred to OHP-4030.STP.007 1-0HP>>4030.STP.013 (Rev.

1) "Semi-Annual Electric Hydrogen Recombiner System Functional Test" 1-0HP-4030.STP.016 (Rev. 4)

"RCS Leak Test" 1-0HP-4030.STP.015 (Rev.

1) "Full Iength Control Rod Operability Test" 1-0HP-4030.STP.018 (Rev. 4) "Steam Generator Stop Valve Operability Surveillance Test" l-OHP-4030.STP.030 (Rev.

8) "Operations Daily and Shift Surveillance Checks" 2-0HP-4030.001.002 (Rev. 0) "Routine. Containment Inspection" 2-OHP"4030.STP.005 (Rev. 3)

"ECCS Operability Test" 2"OHP-4030.STP.008 (Rev.

2)

"ECCS Valve Operability Verification" 2-0HP-4030.STP.022 (Rev.

1)

"ESW System Test" 2-OHP-4030.STP.030 (Rev.

1) "Operations Daily and Shift Surveillance Checks" 2-0HP-4030.STP.031 (Rev.

1) "Operations Weekly Surveillance Checks" Items of nonconformance consist of:

(a)

Incomplete distribution, procedures 12-0HP-4030.001.001, Rev. 4, 1"OHP-4030.STP.015, Rev.

1, 1-0HP-4030.STP.030, Rev. 8, and 2-OHP-4030.STP.030, Rev.

1.

-14-

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V

(b)

Failure to follow cancellation requirements, procedure 1"OHP"4030.STP.006.

(c)

Failure to conduct timely reviews; procedures 1-0HP-4030.STP.006, 1-0HP-4030.STP.013, Rev.

1, 1-0HP-4030.STP.015, Rev.

1, 1-0HP-4030.STP.031, Rev. 0, 2-0HP-4030.001.002, Rev.

0, 2-OHP" 4030.STP.005,'ev.

3, 2-0HP-4030.STP.022, Rev.

1, 2-0HP-4030.STP.

030, Rev.

1, and 2-OHP-4030.STP.031, Rev.

1.

The above represents an item of noncompliance with multiple examples as detailed in Appendix A to the transmittal letter of this report.

(81-03-04)

It should be noted that this inspection effort reviewed only opera-tions procedures.

Of concern is that approximately 50/ of the procedures looked at had not undergone a timely procedure review as committed to by the licensee.

As discussed in paragraph nine, there were numerous revisions and temporary changes that received incomplete safety related evaluations (e.g. Filling and Venting ECCS; Filling, Venting, and Operation of ESW System, Loss of CCW).

The inspectors did observe that the licensee has recently started re-addressing this area and that efforts are underway to upgrade procedures governing the review process.

S ecial Procedures and Tem ora Procedure Chan es The inspector reviewed the Operations, Maintenance, and Technical Departments Special Procedures and attached Temporary Change Sheets which were found in the Document Control Vault to verify proper approvals and to identify any conflicts with Technical Specification requirements.

These procedures, as well as the other procedures addressed above exhibited numerous problems:

Temporary Change Sheets improperly categorized as not safety-related; special procedures dealing with tests were not reviewed to determine whether there were changes in tests described in the FSAR; basis not documented for 50.59 determinations.

These problems reflect adversely on perfor-mance of the PNSRC, review function which is discussed in paragraph nine.

It was also discovered that there was a change to the Technical Specifications which incorporated the "N" train battery modifications and apparently was not adequately reviewed.

During a verification review of procedures, a walk thru of remote alternate shutdown revealed that procedure 1-0HP-4023.001.001 and 2-0HP-4023.001.001

"Alternate Emergency Shutdown and Cooldown Procedure due to Loss of Normal and Preferred Alternate Methods" had not been revised to reflect the current power supply for the Turbine Driven Auxiliary Feedwater Pump.

This condition was found and corrected under the plant's condition report system.

Two observations concerning this

- 15-

Qa item are:

First, that although not specifically required by Technical Specifications (T.S.),

a post implementation review of procedures should be performed by the PNSRC following an amendment to the T.S.

or a modification to the facility.

Second, that periodic review of procedures by the plant organization is adequate to identify and correct deficiencies in a timely manner.

13.'xit Interview The inspector met with licensee representatives (denoted in Paragraph 1)

throughout the month and at the conclusion of the inspection on February 28, 1981, and summarized the scope and findings of the inspection activities.

The licensee acknowledged the inspection findings discussed in paragraphs 9,

10 and 12.

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-16-

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4p n0 UNITEDSTATES NUCLEAR REGULATORYCOMMISSION

REGION III

799 ROOSEVELT ROAD GCEN ELLYN,ILLINOIS60137 February 9,1981 Docket No.

50-315 Docket No.

50-316 American Electric Power Service Corporation Indiana and Michigan Power Company ATTN:

Mr. John E. Dolan Vice Chairman Engineering 2 Broadway New York, NY 10004 Gentlemen:

This refers to the enforcement conference held on January 13, 1981, at the Nuclear Regulatory Commission office at 799 Roosevelt Road in Glen Ellyn, Illinois, between Mr. James G. Keppler and members of his staff and Mr. J.

E. Dolan and members of his staff for the purpose of discussing

'ecent events of concern to the NRC in the operation of the D. C.

Cook Nuclear Plant.

The enclosed copy of our report of this enforcement conference identifies areas discussed during the meeting.

Pith respect to your continuing evalua-tions into the operability of the RHR system, please inform us as to your expected date of resolution of this problem.

No items of noncompliance with NRC requirements were identified as a result of this meeting.

In accordance with Section 2.790 of the NRC's "Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a copy of this letter and the enclosed inspection report will be placed in the NRC's Public Document Room, except as follows. If this report contains information that you or your contractors believe to be proprietary, you must apply in writing to this office, within twenty-five days of the date of this letter, to withhold such information from public disclosure.

The application must include a full statement of the reasons for which the information is considered proprietary, and'should be prepared so that proprietary information identified in the application is contained in an enclosure to the applicatio American Electric Power Service

- 2-Corporation We will gladly discuss any questions you have concerning this meeting.

Sincerely, James G. Keppler Director

Enclosure:

IE Inspection Report No. 50-315/81-02 and No.

50-316/81-02

REGION III==

Reports No. 50-315/81-02; 50-316/81-02 Docket Nos. 50-315; 50-316 Licenses No. DPR-58; DPR-74 Licensee:

American Electric Power Service Corporation Indiana and Michigan Power Company 2 Broadway New York, NY 10004 Facility Name:

D.

C.

Cook Nuclear Plant Units 1 and

Inspection At:

D.

C. Cook Site, Bridgman, MI Inspection Conducted:

Janua 13, 981 Report Prepared By:

D.

C. Boyd, Chief a

or ojects Section

Approved By:

.

. Heishman, Chief, Reactor Operations and Nuclear Support Branch, RIII A/e Fy Ins ection Summa Ins ection on Janua

1981 (Re orts No. 50-315/81-02'0-316/81-02)

Areas Ins ected:

An enforcement conference was held to discuss recent operational events, at the D. C.

Cook Plant, which are of concern to the NRC.

Results:

No items of noncompliance were identified as a result of this meetin DETAILS l.

Attendance American Electric Power Service Co oration Indiana and Michi an Electric Com an J.

E. Dolan, Vice Chairman Engineering R.

S. Hunter, Vice President A. S. Grimes, Staff Engineer S. J. Milioti, Assistant, Division Head Nuclear Engineering B. A. Svensson, Assistant Plant Manager R. S. Keith, Operations Superintendent I.E. Attendees J.

G. Keppler, Director, Region III A. B. Davis, Deputy Director, Region III R. F. Heishman, Chief, Reactor Operations and Nuclear Support Branch C. E. Norelius, Assistant to the Director D. C. Boyd, Chief, Reactor Projects, Section

K. R. Baker, Chief, Nuclear Support, Section

E. R. Swanson, Senior Resident Inspector, D. C.

Cook Plant 2.

Areas Discussed The Regional Director stated that the purpose of the meeting was to review two recent occurences at the D. C. Cook Plant that resulted in reduced margins of safety.

These involved valves in one containment spray system being inadvertantly locked closed and questionable operability

'of one RHR system over several months.

It was pointed out that these events appeared to constitute Severity I,evel III or Severity Level IV violations.

The licensee was requested to review the two events of concern and describe their corrective actions to prevent recurrence of simular types of events.

These discussions were as follows:

a ~

On August 4, 1980 the licensee reported (LER 80-32) that following maintenance on the East RHR Train, a surveillance test on the East RHR Pump indicated that the pump was inoperable.

The cause was attributed to be due to an inadequate filland vent procedure which left air entrapped in the suction lines to the pump.

Subsequent venting and testing on August 4, 1980 resulted in the pump being declared operable.

Followup of this event by NRC inspectors revealed that the East RHR Pump was again declared inoperable during the next two monthly surveillance tests because of entrapped air in the suction to the pump.

This raised the NRC concern that the venting had not been

adequately performed on three successive occasions, following the original maintenance on this RHR Train.

Thus, the possibility existed that this pump was inoperable for the entire period of time.

The licensee stated that they shared the NRC concern and had performed testing to assist in identifying the exact nature of the problem.

This testing indicates that the source of air in the system is not residual air from the initial maintenance performed on the train, but is air introduced into the system during a portion of the monthly surveillance testing of these systems which pulls a suction on a closed check valve to verify check valve operability.

Only one of the four pumps has been affected by this testing, and in each instance the air is vented from the system during the completion of the surveillance test.

Thus, according to the licensee, the pump is always left in an operable status.

The licensee also pointed out that each unit has two 100 percent RHR Pumps and that at no time has the operability of more than one pump at a time been in question.

The licensee stated that, their testing to determine how the air is being drawn into the suction of this pump is continuing and that an alternate method for verifying the operation of the check valve is being developed.

In the mean time the licensee stated that the final functional test portion of each of these monthly surveillance tests verifies that each RHR Pump is left in an operable status, and that they would perform weekly functional testing of these pumps to verify continued operability.

The enforcement considerations associated with this matter will be based on the results of the licensee's further evaluations as to the effect of the air on the operability of the pumps.

On December 14, 1980, while operating at full power, it was discovered that the containment upper and lower containment spray headers on the East Train in Unit 2 were isolated.

The redundant West Train of the containment spray system was found to be correctly aligned.

This event was reported to the NRC in LER 80-33.

The licensee could offer no excuse for this human error.

The sur-veillance test procedure was reviewed and found to be correct, if followed.

In this instance a qualified operator performed the valving associated with the performance of a surveillance test on this system.

At the conclusion of the surveillance test, performed on December 4,

1980, the operator signed, and returned to his supervisor, a check list which indicated that he had opened and locked open the two valves that were subsequently found to be locked in the closed position.

Appropriate disiplinary action has been taken against the operator.

The licensee state'd that since that time they have initiated a double verification of all safety system alignment changes'hus, following the re-alignment of any safety related system, a second qualified individual independently verifies that the system alignment is as require The licensee agreed that this human error did render one train of the containment spray system inoperable, but pointed out.that the second 100 percent redundant train was always operable.

Thus, the system performance was degraded but the safety function was not lost.

The NRC agreed with the above statement and concluded that this event constitutes a violation of category IV.

The Regional Director stated that, in view of past similar problems of this type, this meeting should be viewed as an Enforcement Conference and that similar Severity IV violations in the future would likely result in escalated enfoce-ment action.

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