IR 05000213/1985013

Insp Rept 50-213/85-13 on 850627-0829.No Violations Noted. Eight Outstanding Insp Items Closed.Major Areas Inspected: Plant Operations,Radiation Protection,Physical Security,Fire Protection,Procedures,Tmi Action Plan Items & Part 21 Rept

ML20135D989
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	09/05/1985
From:	Mccabe E, Robertson J, Swetland P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20135D943	List: IR 05000213/1985013 ML20135D941
References
RTR-NUREG-1154, TASK-1.C.1, TASK-1.C.2, TASK-1.C.3, TASK-1.C.4, TASK-1.C.5, TASK-1.C.6, TASK-2.B.2, TASK-2.D.3, TASK-2.E.1.1, TASK-TM 50-213-85-13, GL-82-21, GL-85-13, IEB-79-06, IEB-79-25, IEB-79-27, IEB-80-03, IEB-80-06, IEB-80-20, IEB-80-21, IEIN-83-75, NUDOCS 8509160247
Download: ML20135D989 (20)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /85-13 Docket N License N OPR-61

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Licensee: Connecticut Yankee Atomic Power Company a P. O. Box 270 I

Hartford, CT 06101

.- Facility: Haddam Neck Plant, Haddam, Connecticut

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Inspection at: Haddam Neck Plant

Inspection conducted: June 27 - August 29, 1985 Inspector: & bed * , b 9ls'/PT Paul D. Swetland, Senior Resident Inspector Date

% L 6A, % 9Irler Jeffrey Robertson, Reactor Inspector Date Approved by: & bO , 9 ls Irr E. C. McCabe, Chief, Reactor Projects Section 3B Date Summary: Routine resident inspection (137 hours0.00159 days <br />0.0381 hours <br />2.265212e-4 weeks <br />5.21285e-5 months <br />) of plant operations, radiation protection, physical security, fire protection, procedures, previous inspection findings, events occurring during the inspection, IE Bulletins and Information Notices, Part 21 notification followup, and the status of THI Action Plan Items.

d No violations were identifie Eight outstanding inspection items were close Implementation of corrective actions for licensee identified violations related to overdue Licensee Event Reports and biennial procedure reviews (Detail 6) was verified. Follow-on inspection action was opened for verification of the final disposition of the combined Shift Supervisor / Shift Technical Advisor Position (Detail 8.a). -

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DETAILS 1. Review of Plant Operations The inspector observed plant operation during regular tours of the following plant areas:

-- Control Room -- Security Building

-- Primary Auxiliary Building -- Fence Line (Protected Area)

-- Vital Switchgear Room -- Yard Areas

-- Diesel Generator Rooms -- Turbine Building 1 -- Control Point -- Intake Structure and Pump Building Control room instruments were observed for correlation between channels and for conformance with Technical Specification requirements. The inspector ob-served various alarm conditions which had been received and acknowledge l Operator awareness and response to these conditions were reviewed. Control

) room and shift manning were compared to regulatory requirements. Posting and

] control of radiation and high radiation areas was inspected. Compliance with

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Radiation Work Permits and use of appropriate personnel monitoring devices

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were checked. Plant housekeeping controls were observed, including control

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and storage of flammable material and other potential safety hazards. The inspector also examined the' condition of various fire protection system During plant tours, logs and records were reviewed to determine if entries t were properly made and communicated equipment status / deficiencie These

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records included operating logs, turnover sheets, tagout and jumper logs, process computer printouts, and Plant Information Reports. The inspector observed selected aspects of plant security including access control, physical barriers, and personnel monitoring. No abnormal conditions were identifie . Followup on Previous Inspection Findings Three NRC open items were reviewed. Licensee actions were found to be suffi-l cient to close these item Details follow:

i 2.1 (Closed) Followup Item (213/82-07-01)

i The licensee was to complete an audit of Maintenance department confor-i mance to the requirements of ANSI N18.7, Administrative Controls and

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Quality Assurance for Nuclear Power Plants. Quality Assurance (QA) Audit No. A60220 was completed on March 14, 1985. This audit of maintenance i activities involved 175 audit hours by the corporate QA group and covered ,

the applicable sections of ANSI N18.7. Six audit findings were identi-fled pertaining to the maintenance of controlled documents and procedure ,

The licensee implemented corrective actions for each finding, and fol-lowup inspection has identified no further problems in these areas. This item is closed.

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2.2 (Closed) Unresolved Item (213/84-19-01)

The licensee was to evaluate and implement corrective action as necessary to resolve five open items pertaining to the degraded grid voltage issu The licensee completed an evaluation and implemented corrective action for each unresolved item as detailed in a letter to NRC Licensing dated February 14, 1985. As documented in an NRC Safety Evaluation Report dated July 2,1985, these actions were found by NRC to be sufficient to close the degraded grid voltage issue. The inspector verified that the corrective actions cited in the licensee's February 14 submittal were properly incorporated in revision 3 to procedure AOP 3.2-25, Low Voltage on Emergency Buses, and that operators were made aware of these change This item is close .3 (Closed) Followup Item (213/84-28-02)

The licensee was to develop procedures for detailed verification of safety-related instrumentation valve alignment prior to startup from any extended outage. On July 25, 1985, the licensee implemented procedure 9.2-29, Safety Class Instrument Inspection, Revision 0. This procedure requires verification during plant startup of all safety-related instru-mentation system valve lineups and provides detailed system checklists for accomplishing these checks. The inspector had no further questions in this are . Followup on IE Bulletins (IEBs) and Information Notices (ins)

3.1 Licensee action on the following IE Bulletins was reviewed for forwarding to appropriate management, licensee review for applicability, response timeliness, response appropriateness, response accuracy, corrective ac-tion commitments, and corrective action completio IEB 79-25 Failures of Westinghouse BFD Relays in Safety-Related Systems This bulletin addressed Westinghouse BFD relays that have the po-tential to stick in the energized position when the coil is de-en-ergized. Westinghouse recommended replacing the affected relays with a newer style relay identified as NBFD. The new NBFD relays were later found to exhibit marginal or unsatisfactory armature over-travel . The licensee was asked to determine if such relays were in use or planned for use in safety-related systems, and if so, perform the following: 1) identify the systems involved; 2) describe the specific function of the relays; and 3) provide plans for a test or replacement program which will assure the design performanc The licensee determined that only one relay, as described in IEB 79-25, is utilized at Haddam Neck. It is an NBFD relay which is normally de-energized and is actuated by a "High Containment Pres-

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sure" signal to shutdown Control Room and Radiation Laboratory air conditioning unit fans and Service Building ventilation supply fan The over-travel of this relay was measured and determined to be in the normal range specified by Westinghouse Technical Bulletin, NSD IB-79-05. The function of this relay is tested each refueling by surveillance procedure SUR 5.2-14. " Reactor Containment Pressure i Channel Calibration". Since this relay has acceptable travel, the licensee determined that additional testing of this relay is not i

necessar The inspector reviewed SUR 5.2-14 and verified that this procedure does test the relay operation on actuation of high con-tainment pressure. This test was completed satisfactorily during the 1984 refueling outage. The inspector had no further concerns in this are ' IEB 79-27 Loss of Non-Class-1E Instrumentation and Control Power Buses During Operation The licensee was requested by this bulletin to review the class 1E

., and non-class 1E buses supplying power to safety and non-safety re-1 lated instrumentation and control systems to determine if the loss

of any of these buses could affect the ability to achieve a cold shutdown condition. Also, the licensee was to prepare or review emergency procedures to be used by operators upon loss of power to

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any class 1E or non-class 1E bus supplying instrumentation and con-trol system The Haddam Neck Plant has two semi-vital buses fed by two sources via an automatic bus transfer. Each source is backed up by an emergency diesel generator. There are four vital buses fed by four j

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separate inverters. Two batteries supply two vital inverters apiec Loss of either semi-vital bus supply or loss of any vital bus is

I annunciated on the main control board. Bus voltage is also indi-cated on the back of the main control board.

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The inspector resiewed procedures for " Total Loss of Semi-Vital l

Power" (EOP 3.1-46) and " Partial Loss of DC" (EOP 3.1-49) to verify that they include: (1) the diagnostics / alarms / indicators / symptoms of a loss of vitsi or semi-vital bus; (2) the use of alternate in-dication and control power from other sources; and (3) the methods for restoring power to these buses. No inadequacies were identified.

! IEB 80-03 Loss of Charcoal From Standard Type II, 2 Inch, Tray Adsorber Cells This bulletin identified a problem on certain Type 'II adsorber cell The spacing between rivets securing the charcoal retention screen to the casing was too great to assure adequate contact between the casing and the screen, thus allowing charcoal to escape. The lic-ensee was required to determine if installed charcoal adsorber cells have the potential for a loss of charcoa In particular, the lic-

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ensee was to inspect adsorber cells and rivet spacing for separation of screen and cell housing or for adsorber cell or filter housing deformation causing loss of charcoa Licensee review of charcoal adsorber cell designs did not reveal any cells of the riveted type design. The licensee also performed a visual inspection to verify the integrity of the charcoal adsorber filter The inspector re-viewed the results of this inspection. No inadequacies were iden-tifie IEB 80-20 Failures of Westinghouse Type W-2 Spring Return to Neutral Control Switches This bulletin documented failures of Westinghouse Type W-2 spring return to neutral control switches, in that the closure of the neutral contacts became intermittent. Failure of the neutral con-tacts to close properly would prevent an automatic start function of the associated equipment. The licensee determined that these switches were used for the following components:

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Containment Air Recirculation Fans

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Service Water Pumps High

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Pressure Safety Injection Pumps Low

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Pressure Safety Injection Pumps Charging

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Pumps Emergency

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Diesel Generators Primary Water ,

Transfer Pumps

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In accordance with the Westinghouse Technical Bulletin and IEB 80-20, the licensee modified these control switches and indicating lights such that the indicating lights will confirm when both the control

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switch and its contacts are in the neutral or automatic start posi-

tion. The inspector verified the completion of these modifications in Plant Design Change Request No. 385. Indication of W2 switch continuity is verified during routine resident operational safety checks, and proper circuit operation has been satisfactorily demon-strated during refueling interval emergency core cooling automatic actuation tests (SUR 5.1-18 and 19). The inspector had no further questions in this are e. IEB 80-21 Valve Yokes Supplied by Malcolm Foundry Company, In The bulletin identified severe cracks in valve yokes that had been cast by the Malcolm Foundry Company. An analysis and evaluation concluded that the cracks were due to the yoke material not having the proper material properties. The Malcolm Foundry was no longer in business, and the licensee was asked to determine if any of the active valves in use or planned for use in safety-related systems have valve parts cast by Malcolm Foundr The licensee contacted .

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all suppliers of active valves in safety-related systems and deter-mined that no valves with potentially faulty material had been use No further action was required by this bulleti .2 Licensee action concerning IN 83-75, Improper Control Rod Manipulation, was reviewed to determine that the notice was forwarded to appropriate

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licensee personnel, that the stated concern was evaluated, and that necessary corrective actions were implemented. The inspector also re-viewed the licensee's response to the Institute of Nuclear Power Opera-tions Significant Operating Experience Report 84-2 which covered the same topic. The licensee's evaluation of this topic for Haddam Neck concluded in general that installed rod control equipment and operating procedures provide adequate control over control rod positioning. A need for clari-fication of the terms " substantial misalignment" and "significant time" (misaligned) was identified. The licensee incorporated quantitative definitions for these terms in procedure A0P 3.2-23, Malfunction of the

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Rod Control System. The inspector verified the correct implementation of these changes as well as the operator training rela +.ed to misposi-tioned control rods, procedure A0P 3.2-23 and strict procedural compli-i ance in general. The inspector had no further questions in this are . Followup on Events Occurring During the Inspection 4.1 Licensee Event Reports (LERs)

The following LERs were reviewed for clarity, accuracy of the description of cause, and adequacy of corrective action. The inspector determined whether further information was required and whether there were generic implications. The inspector also verified that the reporting require-

ments of 10 CFR 50.73 and Station Administrative and Operating Procedures had been met, that appropriate corrective action had been taken, and that >

the continued operation of the facility was cnnducted within Technical Specificatinn Limit *85-02 Reduced Containment Air Recirculation (CAR) Flow 85-03 Nonconservative Loss of Flow Setpoint -- Event detailed in Inspection Report 50-213/85-03

• 85-05 Auxiliary Feedwater (AFW) Control Valve Failures

• 85-10 Service Water Motor-0perated Isolation Valve Failures i

85-12 Fire Detection System Surveillance Interval Missed 85-14 Inoperable Fire Door 85-15 Spurious Load Runback

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85-16 Nuclear Instrumentation Dropped Rod 5etpoint Orift

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• 85-17 Post Loss of Coolant Accident (LOCA) Release Paths Outside Containment

• event detailed below 4.2 Reduced CAR Flow (LER 85-02)

During routine refueling interval measurement of CAR system flow rate on August 26, 1984, the licensee identified that 3 out of 4 CAR fan flow rates were less than the minimum acceptance criteria (50,000 CFM) speci-f fled in the surveillance procedure. The licensee concluded that, under worst case conditions, a 7% reduction in CAR flow could have existed as a result of this conditio The reduced fan flow was caused by movement of the flow adjustment levers. The licensee readjusted each fan and bolted the adjustment levers in place to prevent recurrence of this proble Subsequently, the CAR flow rate measurements were completed satisfactoril The licensee did not report this occurrence within 30 days of its iden-tification on August 26, 1985, because the design bases and safety an-alysis assumptions for CAR system operability had not been clearly docu-mented in Technical Specifications or other documents available onsit Upon further evaluation and review of the incident by the or. site review committee (PORC) and the corporate safety analysis group, the licensee determined that each combination of 3 CAR fans must provide 150,000 CFM of flow (50,000 CFM per fan as stated in the procedure) in order to meet

, the design bases for the CAR system. When this determination was made on December 28, 1984, the Licensea Event Report was properly written and submitted to NRC. The inspector reviewed the circumstances surrounding this event and two others (LERs 85-05 and 85-10) detailed below which were also not reported within 30 days. Each incident occurred during a protracted refueling outage and was initially classified as non-re-portable because of a lack of or misinterpretation of system design basis information onsite. Routine followup of each event was significantly delayed because of the volume of followup action generated during the outage. The PORC identified the reporting error during their interim review of actions resulting from these events. As a result of these later reports and other licensee-identified problems with event followup, the licensee implemented the following corrective actions: All plant events will be reviewed by site management at a collective meeting within one working day of tne occurrence. Also, corporate management will receive copies of internal plant event report (PIRs)

within a few days of the occurrence, The site engineering department will review each PI This review includes an evaluation of the effect of identified deficiencies on the system design functions and verifies the plant duty officer

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decision on event reportability. Also, this review establishes prompt contact between engineering and the appropriate repair or-ganization to insure that pertinent information on the cause of the failure is retaine To reduce the backlog and improve the quality of licensee review activities, improved guidance to plant members responsible for plant followup activities has been implemented. This guidance (Unit Policy 8.0) tasks reviewers with detailed root cause determination anu includes documentation criteria which should reduce interim review activities and shorten the overall evaluation process tim The licensee intends to create new staff positions to augment man-power for these and other activitie The licensee had previously initiated a long term program to collect and document system design and safety analysis bases informatio .Upon completion, this material will improve the onsite capability to prouptly assess component and system deficiencie The inspector verified the implementation of these commitment The three late LERs cited above constitute licensee-identified violations of 10 CFR 50.73, for which appropriate corrective action has been imple-mented. The effectiveness of licensee corrective action will be reviewed during a subsequent inspection (IFI 213/85-13-01).

4.3. AFW Control Valve Failures (LER 85-05)

During an AFW automatic actuation test on November 2, 1984, two of four AFW control valves failed to open. The plant was conducting routine testing during preparations for startup from a 3-month refueling outag These feedwater bypass valves are required to open automatically to pro-vide a flow path from the AFW ptm.ps to the steam generators. The valves did not operate because the control air actuation solenoid valves stuc Subsequent to the test, both solenoid valves cycled properly and no fur-ther bypass valve failures occurre The licensee inspected the solenoid valves and found no abnormal indications. Automatic actuation of AFW -

was retested satisfactorily. The licensee committed to perform follow-on testing of the solenoid valves. The inspector verified that these valves were satisfactorily tested during a plant shutdown in May 1985. On August 29, 1985, the licensee approved a plant procedure for periodically testing these valves during plant operation on a frequency to be deter-mined by test results (next test planned in September 1985). With the exception of the four month delay in reporting this event (detailed in Paragraph 4.2, above), the inspector had no further questions in this are .4 Service Water MOV Failures (LER 85-10)

During routine in-service testing cf MOVs on August 28, 1984, the non-essential service water header supply and return isolation valves failed to close. These twh valves isolate on a safety injection actuation sig-

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nal to assure an adequate flow rate of cooling water to the safety-re-lated loads. With both valves inoperable, and assuming no operator ac-tion and a single failure of one of two emergency diesel generators, service water flow to essential accident components would not have been adequat However, the valves could have been operated locally. Both valve operators were overhauled during the August-November 1984 refueling outage. Subsequent valve stroke tests were completed satisfactoril One valve failed due to a broken limit switch roto No similar failure modes or precursors were identified during overhaul of several other MOVs during this outage. The cause of the failure of the second service water MOV was not identified because the licensee's maintenace record of over-haul did not address the deficiencies found. The licensee concluded that the implementation of the improved MOV preventive maintenance program in 1984 and continued in-service testing would be adequate to prevent recurrence of these failures. As stated in Paragraph 4.2. above, the licensee has taken action to assure prompt assignment of followup re-sponsibility and root cause analysis for each significant failure. Both these actions, in part, focus supervisory attention on better documenta-tion of maintenance activities. As noted above, the effectiveness of these actions will be reviewed during subsequent NRC inspection At

present, the inspector has no further questions on this ite I

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4.5 Post-LOCA Release Paths Outside Containment (LER 85-17)

During a review of previous design changes in July 1985, the licensee identified two potential paths for release of post-accident radioactivity from containmen Post-LOCA safety injection recirculation may be routed from the l containment sump through the residual heat removal (RHR) pumps to the charging pumps and then into the reactor coolant system. RHR system leakage outside containment is monitored during plant opera-tion in order to limit offsite releases from this system during the post-accident recirculation phase. Charging system leakage had not been limited or monitored to the same specification (3 liters / hour)

> as the RHR syste Charging system leakage would be an additional unanalyzed release path when operated in the piggyback recirculation mod The licensee confirmed that combined charging and RHR systems'

leakage is below 3 liters / hour and has established routine surveil-lance of the leakage from both systems with a 3 liter / hour combined limi Another potential post-accident release path resulted from TMI ac-

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, tinn plan modifications completed in 1981. These modifications re-moved automatic safety injection / containment actuation to the iso-lation valves for reactor coolant pump (RCP) seal injection, return and cooling lines in order to preserve the capability for post-ac-cident RCP operation. These isolation valves remain remotely oper-able from the control room. Post-LOCA, if offsite power is lost, the charging pumps stop and are not automatically restarte If

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.1 s there were no operator action, the unisolated RCP seal return line would pressurize from seal leakage which is normally rou.ted to the charging pump suction. This line relieves the volume co'ntrol tank-(VCT) at 140 psig and will eventually fill the VCT and overflow to '

the waste treatment system. If no action is taken to stop the seal water leakage, a flow path for radioactive fission products to the ;

plant si.ack would eventually exist through the waste treatment sys-tem. The 'iicensee informed plant operators of this potential post-accident rehase path, including expected indications and appro- '

priate actions to stop the leakpg The inspector verified the licensee's implementation of charging system leakage monitoring and observed selected charging system component and piping locations for evidence of leakage. No abnormal conditions were identified. The inspector also verified that information concerning these scenarios had been routed to all operations personnel. These re- :

lease paths could exist only under certain post-LOCA conditions, and as- i sume no operator response to several availabl.e indications and alarm l However, the potential release paths were not considered in current off-site dose calculations which are part of the licensing basis for this

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plant. Results of these calculations are presently close to the 10 CFR "

100 thyroid dose limits. :Since this event involves issues previously reviewed by NRC Licensing (TMI Action Plan Itens II.E.4.2 and III.D.1.1) :

and could affect the bases for current offsite' dose calculations for this I license, the matter is being referred to'NRC Licensing. This item will l remain unresolved pending completion of NRC, review >and implementation

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of corrective action as necessary (UNR 213/05-13-02). i

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4.6 Spurious Load Runbacks (LER 85-19) '

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l On July 26, 1985, while reducing load for condenser cleaning, plant I operators adjusted gains on the power range nuclear! instruments (NIS) ;

in accordance with normal operating procedures. While adjusting NIS t Channel 33, a spurious power' spike occurred, and actuated the rod drop i protection circuits of the reactor protection system. The rod drop pro- i tection provides an. automatic main turbine load runback and blocks con- ;

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trol rod l withdrawal /when a sharp' change (5% decrease) is sensed by any

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1 of the'4 power range'NIS chantiels. There was no indication of an !

actual rod drop, so operators reset the rod drop protection circuit, preventing completion of the turbine load runback and restoring normal ,

rod control. On July 27, while increasing power after completion of _[

condenser cleaning, another spurious load runback occurred when the NIS ;

Channel 33 gain was being adjusted. The cause of the two spurious ac- ;

tuations of the rod'_ drop protection was determined to be a faulty gain adjust potentiometer inithe Channel 33 NIS drawer. The Channel 33 drawer -

wasireplaced with a spare' drawer and the faulty potentiometer'was re- ;

placed on July 29, 1985. A similar event (LER 85-15) involving the gain !

potentiometer on NIS Channel 34 occurred in June 1985. ; As a result of -

these events',' the licensee has scheduled replacement of gain potentio- ;

meters in the remaining NIS drawers during the next refueling outage in

January 1986. .The inspector'had no further questions'in this area.

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5. Review of Periodic and Special Reports Upon receipt, periodic and special reports submitted pursuant to Technical Specification 6.9 were reviewed. This review verified that the reported in-formation was valid and included the NRC required data; that test results and supporting information were consistent with design predictions and performance specifications; and that planned corrective actions were adequate for resolu-tion of any problems. The inspector also ascertained whether any reported information should be classified as an abnormal occurrence. The following periodic reports were reviewed:

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Monthly Operating Reports 85-05 through 85-07 These reports covered plant operation during the period May 1 through July 31, 198 . Followup on the Implementation of Biennial Procedure Review Technical Specification 6.8 and ANSI N18.7, Administrative Controls...for Nuclear Power Plants, require procedures affecting safety to be reviewed for accuracy and effectiveness at intervals no greater than 2 years. The licensee has recently implemented detailed procedure review guidelines in order to improve the detail and consistency of biennial procedure reviews. The in-spector reviewed these guidelines which have been incorporated in departmental instructions and which included checks for procedure content, format, clarity, workability, and applicability to the latest system configuration. The in-spector verified that these new guidelines were appropriately used for review of five recently completed procedure review NRC reviewed the status and tracking of biennial procedure review during In-spection 50-213/84-32. This inspection documented licensee corrective actions to reduce a backlog of overdue procedure reviews and prever.t its recurrenc The inspector reviewed the licensee's procedure review status tracking reports for June and August 1985. These reports show that, while the numerical back-log of reviews has been reduced, new overdue reviews have occurred. The in-spector brought this concern to licensee management attention on July 16 and August 8, 1985. The licensee committed to the following additional corrective actions: Backlog to be eliminated within 30 day Extension of scheduling window from 4 to 6 month Controlled routings issued 3 months prior to due date with no extensions permitte Due dates changed through early procedure review to more evenly distri-bute departmental workload _

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These actions were formalized in Station Superintendent memorandum 85-004 dated August 18, 1985. Since this occurrence was identified by the licensee's tracking system and corrective actions have been implemented, the inspector determined that this item would be carried as a licensee identified violation of the requirements cited above. The effectiveness of the licensce's correc-tive actions will be reviewed in a subsequent inspection (IFI 213/85-13-03).

7. Followup on a Part 21 Notification Regarding Diesel Generator (EDG) Lube Oil Modifications On May 30, 1985, the licensee received notification of a possible problem with recent EDG lube oil modifications implemented to alleviate concerns over tur-bocharger bearing pre-lubrication. This concern was the subject of a February 14, 1985, Morrison-Knudsen Company report to NRC in accordance with 10 CFR 2 In some installations, the arrangement and installation of a DC powered standby lube oil pump prevented the proper operation of the low lube oil pressure indication and alarm. Licensee review of this problem has determined that the DC pump modification was not used at this facility, and the proper functioning of the lube oil pressure indication has been verified. This evaluation is documented in a memorandum (EN 85-646) dated May 16, 198 The inspector verified that no DC lube oil pumps are used for EDGs and that the lube oil alarm and indicator function properly during routine surveillance testing of the EDG The inspector had no further questions in this are . Followup on TMI Action Plan Items I. A.1.1 - Shif t Technical Advisor (STA) (OPEN)

Licensees were to implement on-shift engineering expertise and describe the current STA qualification and training programs and the long term plan for maintenance or eventual elimination of the STA program. The licensee implemented an interim STA program and submitted the required program descriptions. This implementation was verified during NRC Region I Inspection 50-213/80-21. The interim STA program was approved by NRC by letters dated January 13 and March 31, 1982. STA requirements were subsequently incorporated in plant Technical Specifications, and licensee compliance has been verified during routine resident inspector opera-tional safety check The licensee has submitted and implemented his long term STA progra This program calls for training of shift supervisory personnel to meet licensee standards for equivalent engineering education and certification of these personnel for a combined shift supervisor (SR0)/STA positio The licensee informed NRC that this program would be implemented on shift on January 1, 1984, and completed that commitment. NRC approval of the long term STA program is awaiting final approval of the Commission Policy on Engineering Expertise On Shift. This remains open pending final dis-position of item by NRC Licensing (IFI 213/85-13-04).

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' I.A.1.2, I.C.3 and I.C.4 - Shift Supervisor Responsibilities and Control

Room Access (CLOSED)

j Shift supervisor responsibilities and the control of non-safety-related 1 duties were defined by the Vice President, Nuclear Operation in December 1979. Satisfactory implementation of these action plan items was docu-mented in NRC Inspection Report 50-213/80-21, and these licensee commit-ments were approved for meeting the TMI lessons-learned short-term re-quired actions by NRC Licensing on May 7, 1980. These policies are re-viewed annually at the vice president level, and the written policy statement is reissued to shift supervisors. The inspector verified that the latest policy statement, dated January 7,1985, and site administra-tive procedures ACP 1.0-3, Revision 2, Connecticut Yankee Organization Responsibility and Authority, and ADM 1.1-60, Revision 2, Access to Con-trol Room during Emergency Conditions, remain consistent with NRC guid-ance in NUREG 0578 item 2.2.1.a and 2.2.2.a. No inadequacies were iden-tified.

, I. A.1.3 - Shif t Manning (CLOSED)

The licensee was to provide administrative control and limitations over

the use of staff overtime and provide augmented licensed operator staff-

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ing on shift. The licensee implemented staff overtime controls in March 1981. Those controls were verified during NRC Inspection 50-213/81-03 and were approved by an NRC Licensing SER dated February 8,1982. Sub-l sequently, the Commission Policy on Staff Work Hours and NRC Generic Letter 82-21 provided further guidance on staff overtime and requested i licensee commitment to this standard and incorporation of overtime re-quirements in plant Technical Specifications. The inspector verified that the licensee has correctly implemented current NRC overtime guidance l in proceoure 1.1-116, Overtime Controls for CY personnel, except that deviation from certain of the guidelines may be approved by first line supervision rather than station management. In these cases, the Station Superintendent reviews the first line approval after the fac This ,

deviation had been previously accepted by NRC during the 1981-82 review  !

of this action plan item. The licensee has submitted a TS change request '

i to NRC incorporating the overtime limitation policy. The augmentation of shift staffing aspects of this item required adding a second licensed

, senior reactor operator to each operating shift. This requirement was  !

i subsequently codified in 10 CFR 50.54(m). The licensee met the rule im-plementation schedule of January 1,1984. NRC Licensing approved TS

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Amendment 61 on January 15, 1985, which modified TS staffing requirements to conform with 10 CFR 50.54(m). Licensee implementation of staffing l requirements is verified on a routine basis during resident inspector  :

operational safety verification checks. The inspector had no further  !

questions in this are !

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, I.A.2.1 and II.B.4 - Upgrade R0 and SR0 Training and Training for j Mitigating Core Damage (CLOSED)

The licensee was to upgrade initial operator and re qualification pro-grams and provide initial and follow-on training for mitigating core damage in accordance with NRC Licensing guidance issued March 28, 198 The licensee's implementation of these requirements was reviewed during NRC Inspections 50-213/81-12 and 82-19. Based on the results of these

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inspections and NRC Licensing review of the licensee training program

! in these areas, these items were formally resolved in an NRC Safety Evaluation Report dated December 15, 1982. Since then, several licensee

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candidates for R0 and SR0 licenses have successfully completed the re-vised NRC license examination process, indicating continued satisfactory operator training programs, and recent NRC review of operator requalifi-i cation programs have identified no problems with these TMI action items specifically (other concerns about operator requalification are being addressed separately). The inspector had no further questions on these

TMI action items.

} I.C.1 - Short Term Accident and Procedures Review (OPEN)

This action item involved an initial review of and revision of procedures for small break loss of coolant accidents (SBLOCAs) and subsequently a detailed re-analysis and procedure revision of all transients, accidents and inadequate core cooling situations. The licensee completed an in-itial review of SBLOCAs in 1980. The results of this analysis and the procedure changes implemented were reviewed during NRC Inspection 50-213/80-04. Further NRC guidance related follow-on analyses was issued on December 17, 1982, as NRC Generic Letter (GL) 82-33. Completion of this item was separated into three sub parts: (1) NRC review and ap-proval of owner group generic emergency response guidelines (ERGS);

(2) NRC review and approval of licensee procedure generation packages which delineate how the generic guidelines are tailored and implemented i as plant specific procedures; and (3) validation, training and implemen-i

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tation of revised plant specific emergency procedures. The Westinghouse generic ERGS were approved by NRC for implementation by GL 83-22 dated June 3, 1983. The licensee submitted a procedure generation package

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(PGP) on September 1, 1983, and provided addition information in this area by letter dated July 12, 1985. Licensee development of revised

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emergency operating procedures (EOPs) is ongoing. The licensee intends

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to delay implementation of revised E0Ps from May 1986 to September 1986 in order to complete validation and operator training on a plant-specific simulato The PGP and E0P implementation schedule remain under review l by NRC Licensing.

. I.C.2 - Shif t Turnover Procedures (CLOSED)

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, The licensee was to review and revise shift turnover procedures to assure

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valid and effective transfer of plant operating informatio Procedures

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ADM 1.1-45, Shift Relief and Turnover, and NOP 2.2-2, Steady State

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Operation and Surveillance, were reviewed by NRC inspection 50-213/80-21 and found satisfactory for completion of this item. Procedure ADM 1.1-45 has been superseded by Adminstrative Control Procedure 1.0-8, Shift Re-lief and Turnover. The inspector verified that the licensee commitments to this action item (I.C.2) continue to be implemented by the new pro-cedure. Adequate shift turnover is routinely verified during NRC resi-dent inspector operational safety checks. No inadequacies were identi-fied.

. I.C.5 - Feedback of Operatina Experience (CLOSED)

The licensee was to identify and implement organizational responsibili-ties for screening and dissemination of industry operating informatio The intent is that important operating experience be received by the

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plant staff in a timely manner and that operators not be burdened with

extraneous information which could detract from their job performanc This program was implemented in 1981 by procedure NEO 5.08, Operating Experience Assessment and Utilization. This item was verified during NRC Inspection 50-213/81-05. Subsequently, the licensee has expanded the role of the Nuclear Safety Engineering personnel who perform the operating information assessment function. The licensee's program for feedback of operating experience is currently contained in procedure NE0 2.06, Operating Experience Assessment and Utilization. The inspector verified that intent and function of this action plan item have been re-tained. The plant operating staff has access to important recent oper-ating experience without being distracted by unimportant information.

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The inspector had no further questions in this are h. I.C.6 - Verifying Correct Performance of Operatina Activities (CLOSED)

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The licensee was to implement a program for verification of correct alignment of safety-related systems. By letter dated February 8, 1982, NRC approved the licensee's committed program which includes a deviation from NRC guidance on this issue. Specifically, job supervisors, rather than operators, perform second verification checks. The licensee's im-plementation of this item was verified to be satisfactory during NRC In-spections 50-213/81-03 and 83-27. During routine resident inspections, no further inadequacies have been identified. The inspector had no fur-ther questions on this ite . II.B.2 - Plant Shielding (CLOSED)

This Action Plan item requires the licensee to perform a plant shielding design review and make necessary modifications to insure post-accident access to vital plant operating areas. The licensee's implementation of this item (with the exception of equipment environmental qualifica-tion) was approved by NRC Licensing in a safety evaluation report dated July 26, 198 NRC Inspection Reports 50-213/83-12 and 83-16 document satisfactory completion of all licensee commitments in this area. The environmental qualification aspects of this item will be addressed during NRC followup on licensee implementation of 10 CFR 50.49.

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i j j . II.D.3 - Direct Indicatior, of Relief and Safety Vahe Position (CLOSED) .

I The licensee was to assure positive indication of reactor coolant system <

1 relief / safety valve position indication by direct indication of valve i position or by appropriate downstream flow indication. The licensee in-stalled an acoustic flow indicator in the common tailpipe header for the

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two power-operated relief valves and the three pressurizer safety valves.

' When used with the stem mounted limit switch PORV position indicators i and the individual safety valve tailpipe temperature indicators, it is i possible to determine which' single relief / safety valve is open. This j commitment was approved by NRC Licensing on May 7, 1980, and the instal-a lation was verified during NRC Inspection 50-213/80-21. Equipment en-vironmental qualification issues will be reviewed during NRC followup of 10 CFR 50.49. The licensee has implemented periodic surveillance of

these components. The inspector identified no problems with continued

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operability of this feature and had no further questions on this ite :

I II.E.1.1 - Auxiliary Feedwater System Evaluation (CLOSED)

. NRC and the licensee were to complete an evaluation of the auxiliary J

feedwater (AFW) system including system reliability, a deterministic re-i view based on Standard Review Plan (SRP) 10.4.9 and re-evaluation of AFW

system flowrate design bases and criteria. These reviews were conducted j as documented in NRC Licensing Safety Evaluation Reports dated November 20, 1981 and November 9, 1982. Several plant modifications, including i

improved demineralized water storage. tank level indication and alarms, i~

installation of an AFW pump cross-connect isolation valve and redundant pump suction isolation valve, and permanent connection of the motor-

driven startup feedwater pump to the AFW system, were. completed to up-

! grade the reliability of the AFW system. The inspector verified the 1 installation of these modifications as committed to by the license l During the AFW system review, it was noted that the AFW system does not-i meet the SRP 10.4.9 criteria for system power supply diversit In its j SER dated November 9, 1982, the NRC approved this deviation noting the availability of the motor-driven feed pump, the time available should

, both turbine-driven pumps fail (30 minutes) for operator action to manu-ally lineup and start the motor-driven pump before steam generator dry-t out, and the relative reliability of the turbine-driven pumps during loss

of power type scenario The licensee supplemented this information on January 11, 1983, noting that af ter locally (outside the control room)

lining up the motor-driven pump, feed to the steam generators using this

! pump is controlled from the control room. The inspector verified that

, the use of the motor-driven feed pump during a loss of feedwater accident

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had been properly incorporated in procedures E0P 3,1-44, Abnormal Feed 1 Flow, Revision 12 and NOP 2.18-3, Electric AFW Pump Operation, Revision j l' A loss of feedwater incident recently occurred at the Davis-Besse plan NRC review of this incident resulted in several recommendations.for fol- i

lowup action These actions are documented in NRC Generic Letter 85-13 ,

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and NUREG 1154, Loss of Main and Auxiliary Feedwater at the Davis-Besse Plant on June 9, 1985. The following conclusion is extracted from NUREG i 1154 and will receive further evaluation by the NRC:

The availability of the electric motor-driven startup feedwater pump significantly improved the safety margin for the plant during the event. The capability to promptly place an electric motor-driven pump and associated valves for supplying auxiliary feedwater in service from the control room would have significantly increased the safety margin for the plant during the even This item remains open pending disposition of the Davis-Besse recommen-dations. (IFI 85-13-05) TMI action item aspects are, however, close . II.E.1.2 - AFW System Initiation and Flow (CLOSED)

In a two phase process, the licensee was to provide safety grade flow indication and automatic initiation of the AFW System. Phase I required

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rapid installation of control grade equipment. That was done in 198 This installation was verified during NRC Inspections 50-213/80-21 and 82-11. The licensee subsequently upgraded the system's qualification i to safety grade and received NRC Licensing approval for complete activa-l tion of these features in Amendment 44 to the Operating License, which included Limiting Conditions for Operation and Surveillance requirement License compliance with these requirements has been verified during routine resident inspection. Final NRC disposition of this action plan item was documented in an NRC SER dated October 5, 1985. This item is close m. II.E.3 - Emergency Power for Pressurizer Heaters (CLOSED)

The licensee was to provide redundant pressurizer heater capacity capable of being powered from emergency power sources and adequate to sustain natural circulation cooling. In addition, the licensee was to assure that these heaters do not overload emergency power capacity during safety injection system operation. The licensee has two redundant pressurizer backup heater groups (A and E) which are powered from vital buses in order to sustain natural circulation. These heaters are stripped from the emergency buses on a safety injection actuation and are reenergized by procedure when required for natural circulation. This system design was approved by NRC Licensing SER dated May 7, 1980, and the installation was verified during NRC Inspection 50-213/80-21. The inspector verified that the system design and operating procedures remain consistent with the requirements for this action plan item, n. II.E.4.1-DedicatedHydrogenPenetrations(CLOSED)

The licensee was to provide redundant capability to remove hydrogen from the post-accident containment atmosphere by dedicated penetrations for an external recombiner connection or by acceptable containment purge

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i i systems. The licensee had two systems capable of purging the containment i atmosphere through charcoal and particulate filters. These systems com-i bined to provida an acceptable alternative to the required single-failure l proof design. NRC Licensing approved the licensee's proposed alternative i

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in SERs dated May 7, 1980 and September 3, 1981. The inspector verified  ;

i that both purge paths remain available and that operating procedure j 2.13-4, Venting of Hydrogen from Containment Following LOCA, provides ,

j instructions for containment purging. The inspector had no further l

questions in this area.

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. II.E.4.2.1-4 - Diverse Containment Isolation (CLOSED)

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, The licensee was to provide diverse actuation signals to isolate all non-i essential containment penetrations and insure that no penetrations are '

! unisolated without direct operator action upon reset of the actuation j' signal. The licensee added a safety injection signal to the high con-tainment pressure actuation logic for all containment isolation valve In addition, containment isolation system (CIS) logic was changed to in-sure that all CIS valve position switches must be in the closed position i in order to reset the CIS logic. Implementation of these modifications i j were verified during NRC Inspection 50-213/80-21. IE Bulletin 80-06, i Engineered Safety Features Reset, also applied to this area. Licensee i compliance with the guidance of this Bulletin was reviewed during NRC

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Inspection 50-213/84-32. The inspector had no further questions in this area, i

p. II.E.4.2.5 - Containment Pressure Setpoint (CLOSED)

The licensee was to evaluate the containment pressure isolation setpoint  !

to assure that all non essential penetrations are isolated at a minimum pressure compatible with normal operating conditions. The licensee de-j! termined this minimum setpoint to be 4.7 psi. .The setpoint is imple-

! mented by procedure 5.2-55, High Containment Pressure Switch Channel l Calibration, and was verified during NRC Inspection 50-213/81-1 The [

j inspector verified that the current setpoint is the same as that approved  ;

by NRC Licensing ~SER dated August 5, 198 No inadequacies were iden-tified.

l q. II.E.4.2.6 and 7 - Containment Purge Valves (CLOSED)

i These items required the licensee to close and lock containment pump I valves which do not satisfy the' current operability criteria of Branch l Technical Position CSB 6- In addition, the licensee was to provide a trip signal to the purge valves from a qualified radiation detector.

j Plant Technical Specifications (TSs) for containment integrity at Haddam Neck required the purge valves to be closed whenever the reactor coolant temperature is above 200 F. Containment integrity procedures required i operators to lock these valves closed prior to exceeding 200 F. Since

the purge valves are not opened during plant operation, the radiation j trip signal was judged to be unnecessary. This approach was approved l

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by NRC Licensing SER dated October 4,1982 and verified during NRC In-spection 50-213/81-03. The inspector verified that these licensee com-mitments remain implemented by TS 3.11, Containment, and Procedure 2.13-5, Reactor Containment Control System Establishing Containment Integrit II.F.2.1 - Subcooled Margin Monitor (CLOSED)

The licensee was to provide a con inuous control room display of reactor coolant temperature or pressure margin to saturation. Additionally, backup manual methods to verify margin to saturation conditions in the primary system were to be available in the control room. The licensee installed a subcooled margin monitor (SCMM) on the main control boar This design was approved by NRC Licensing SER dated May 7, 1980. Veri-fication of this installation was performed during NRC Inspection 50-213/80-21. The operability of the SCMM is verified during routine resi-dent inspector operational safety checks. The inspector also verified that backup saturation monitoring curves and instructions are contained in procedure 3.1-4, Loss of Coolant Accident. No inadequacies were identifie II.G.1 - Power Supplies for Pressurizer Relief and Block Valves, and

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Pressurizer Level Indicators (CLOSED)

The licensee was to provide emergency power supplies to the pressurizer relief (PORV) and block valves, and level indications. The plant was designed with three pressurizer level indicators, each powered from a separate vital instrument bu Pressurizer relief and block valves are powered from motor control center (MCC) 5 which receives power from either emergency diesel generator through an automatic bus transfer de-vice. Each PORV and block valve receives power from the same sourc This arrangement was judged to be consistent with the overall plant de-sign philosophy and accepted by NRC Licensing SER dated May 7, 198 These design features were verified during NRC Inspection 50-213/80-2 The licensee has recently upgraded the capabilities of the PORVs and the control air system which operates them. The inspector verified that the approved power supply design scheme has been retained in these modifica-tions, which were made to improve PORV reliability. The inspector had no further questions in this area.

1 t. II.K.1.3 - IE Bulletin 79-06 Completion (CLOSED)

The licensee's implementation of commitments responding to IE Bulletin 79-06, Nuclear Incident at Three Mile Island, was reviewed during NRC Inspection 79-13. All outstanding items from that inspection have been closed and no further action is required, u. II.K.3.1 - Automatic Isolation of the Power-Operated Relief Valves (PORVs)

(CLOSED)

The licensee was to provide protection from a small break loss-of-coolant accident by automatically closing the PORV block valves should the PORVs fail to close on decreasing reactor coolant pressure. The licensee pro-

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vided manual override capability on the main control board to close the PORV block valves, but did not provide an automatic block valve closure syste By letter dated September 9, 1983, NRC Licensing concurred in the licensee's position that no PORV isolation system was warrante Consequently, this item was close II.K.3.9 - Proportional Integral Derivative (PID) Controller Modifications (CLOSED)

Haddam Neck does not have PID controllers for the pressurizer power-operated relief valve Therefore, this action plan item is not appli-cable to this licens III.D.1.1 - Primary Coolant Outside Containment (CLOSED)

The licensee was to implement immediate and long term actions to reduce leakage from systems outside containment that could contain highly radioactive fluids. The licensee provided satisfactory evidence that prompt actions had reduced applicable system leakage to as-low-as prac-tical levels. The licensee's continuing leak reduction program was ap-proved by NRC Licensing SER dated May 7, 1980 and reviewed during NRC Inspection 50-213/80-21. The inspector reviewed the continued implemen-tation of this leak reduction program under license procedure 5.7-24, Inservice Visual Examination. This program has been expanded since its initial implementation and maintains the commitments of this action plan ite . Meeting with Local Officials On July 17, 1985, the senior resident inspector met with the First Selectman of the Town of Haddam, Connecticu The meeting was requested by the inspec-tor to continue an open dialogue with local officials and to emphasize the availability of NRC personnel to the local community. During the meeting the inspector highlighted the NRC mission, the functional organization of the NRC, and its relationship to the local community. Mutual topics of interest, in-cluding emergency response and recent events at the Haddam Neck Plant, were also discussed. At the conclusion of the meeting, the inspector encouraged the selectman to contact the resident inspector anytime he or his constituents ,

have any questions or concerns regarding the Haddam Neck Plan . Exit Interview Ouring this inspection, meetings were held with plant management to discuss the findings. No proprietary information related to this inspection was identified.

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