IR 05000213/1985011

Insp Rept 50-213/85-11 on 850510-0614.No Violations or Unacceptable Conditions Noted.Major Areas Inspected:Plant Operations,Radiation Protection,Physical Security,Fire Protection,Emergency Plan Exercise & SEP Topics

ML20129D485
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	07/02/1985
From:	Mccabe E, Swetland P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20129D470	List: IR 05000213/1985011 ML20129D465
References
TASK-06-01, TASK-6-1, TASK-RR 50-213-85-11, IEIN-84-74, NUDOCS 8507160546
Download: ML20129D485 (7)

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U.S. NUCLEAR REGULATORY COMMISSION- DCS No. 50213-850516

REGION I

Report N /85-11 Docket N License N DPR-61

- Licensee: Connecticut Yankee Atomic Power Company P. O. Box 270 Hartford, CT 06101-Facility: Haddam Neck Plant, Haddam, Connecticut Inspection at: Haddam Neck Plant Inspection conducted: May 10 - June 14, 1985 Inspector: Ge. % Ode ,b 7/2/er Paul D. Swetland, Senior Resident Inspector Date Signed Approved by: & kOdA r /2 /Pr E. C. McCabe, Chief, Reactor Projects Date Signed Section 3B, Division of Reactor Projects Summary: Routine resident inspection (66 hours7.638889e-4 days <br />0.0183 hours <br />1.09127e-4 weeks <br />2.5113e-5 months <br />) of plant operations, radiation protection, physical security, fire protection, Emergency Plan exercise, previous inspection findings, events occurring during the inspection, Systematic Evaluation Program topics, and a special inspection of potential inter-systems loss of coolant accident configuration Seven open items were close No violations or unacceptable conditions were iden-Lifie ~

8507160546 850708 PDR ADOCK 05000213 G PDR

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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

-- Control Poin Intake Structure and Pump Building Control room instruments were observed for correlation between channels and for.conformance with Technical Specification requirements. The inspector

, observed various alarm conditions which had been received and acknowledge 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 inspecte Compliance with Radiation Work Permits and use of appropriate personnel monitoring devices were checked. Plant housekeeping controls were observed, including control 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

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were properly made and communicated ' equipment status / deficiencies. These 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 monitorin On June 7, 1985, the licensee conducted a limited site emergency drill to ex-ercise the station evacuation and-personnel accountability procedures. The drill, ' simulating loss of cooling and shielding of used fuel elements in the-spent fuel. pool, was initiated at 9:00 a.m. Site evacuation was ordered at

.9:02. The emergency response-force completed protected area accountability at 9:35 and' site accountability was finished at 9:42 a.m. Upon completion of these drill objectives, the exercise was terminated. The inspector veri-fled that observations related to the improper assignment of technical. support center accountability in procedure 1.5-26, Manager of Control Room Operations, were corrected by the licensee. No-further discrepancies were identifie .- Followup on Previous Inspection Findings During the course of the inspection, six NRC open items were reviewe The inspector found-licensee actions with regard to these areas to be sufficient to close these items. Details follow:

2.1 '(Closed) Followup Item (213/82-22-05) The licensee was to formalize restoration procedures for emergency operations ~ facility (EOF) distribu-tion loads which do not re-energize following a loss of power event.'

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The licensee determined that only EOF ventilation fans and dampers must ;

be realigned after the E0F emergency diesel generator restores powe '

The inspector reviewed the implementation of restoration responsibility and instructions-in revision 3 to procedure 1.5-19, E0F Actuation. No discrepancies'were identifie .2 (Closed) Unresolved Item (213/84-07-05) The NRC was to review repair work on the containment high range radiaticn monitors and licensee cor-rective actions for the identified design change control program defi-ciencies. The radiation monitor cable-to-detector connections were re-worked during the 1984 refueling outage using environmentally qualified- l connectors-and specified heat shrink insulation. The inspector identified no further discrepancies between this installation and the approved de-sign' change. Subsequent'to the identification of design change control deficiencies in this modification, further design control problems.were identified in other modifications, and NRC enforcement action was taken as detailed in NRC Region I Inspection Report 50-213/84-23. Licensee

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corrective actions include a comprehensive revision of the design change process and an ongoing independent review of that process and the imple-mentation of previous modifications. Implementation of those actions is being separately followed by NRC. This item is close .3 (Closed) . Unresolved Item (213/84-08-01) Subsequent to the identifica-tion of design deficiencies related to' post-accident sample system modi-fications, the licensee committed to review the design change control /

modification program to identify areas for improvement to prevent recur-rence of similar events. Other problems identified in the design change control areas reinforced the need for a comprehensive upgrade of this area. The licensee approved revised quality assurance procedures related to design control in November 1984. Implementation _of these new proce-dures was completed in February 1985. An independent review of the ef-

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fectiveness of-these measures is under way, and NRC is monitoring the progress of licensee activities as a followup'to escalated enforcement action in this area. This item is close .4 (Closed) Followup Item (213/84-14-02) The licensee was to provide _for-mal guidance to plant personnel regarding control of fire doors. In May

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1985, the licensee implemented administrative control procedure 1.0-29, Control of Fire Doors. This procedure defines the proper operation of fire door hardware and prescribes appropriate compensatory actions for inoperable fire doors. The' inspector verified that the control measures specified were consistent with Plant Technical Specification requirement The inspector had no further questions in this~ are ' 2.5 (Closed) ~ Violation (213/84-22-01) The licensee failed to incorporate safety-related voltage regulator settings in several procedures uscd to shut down the emergency diesel generators (EDGs).- Mis-setting the volt-age regulator could' prevent the EDG from assuming safeguards loads on demand. The licensee revised procedures 3.1-9, Total Loss of AC Power; 3.1-10, Partial Loss of AC Power; and 5.1-18 & 19, Loss of AC Testing

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with Core Cooling Actuation to include appropriate voltage regulator settings prior to EDG shut down. Licensee and NRC review of other volt-age sensitive components identified no similar discrepancies. To prevent recurrence of similar problems, the licensee has initiated'a safety-re-lated setpoint control program to define and control safety significant settings which are incorporated in the plant design. This program was implemented in October 1984 by administrative control Procedure 1.2-3.3, Setpoint Change Request. The inspector had no further questions in this are .6 (Closed) followup Item (213/80-BU-05) In order to assure that adequate vent capacity was available to prevent collapsing the refueling water i storage tank (RWST) during maximum pump-down conditions, a 14 inch tem-porar" vent was installed on the RWST manway. The licensee was to com-plete permanent modifications to the RWST vent and restore the manway to its original configuration. The licensee completed the installation of plant Design Change Request 532, implementing these modifications, in April 1985. The inspector reviewed the completed modification package and verified the satisfactory implementation of this change. This action completes NRC review of licensee commitments to IE Bulletin 80-0 . Followup on Systematic Evaluation Program Findings 3.1 Topic VI-1, Organic Materials and Post-Accident Chemistry (Action Item 4.21)

To control the pH of containment sump water used for post-accident re-circulation cooling, the licensee' committed to install baskets of tri-sodium phosphate (TSP) in the containment lower level. The licensee in-stalled these baskets during the 1984 refueling outage. The inspector reviewed the implementation of PDCR 618, TSP Baskets. Two 25 cubic foot TSP baskets were installed close to the containment sump suction. Peri-odic surveillance of the chemical effectiveness was implemented on June 7, 1985, by procedure 5.4-36, TSP Surveillance Requirements. The in- .

spector verified that operability and surveillance requirements for the l TSP baskets were included in draft Standard Technical Specifications for ;

the Haddam Neck Plant, which are to be submitted for approval in 198 The inspector had no further questions in this are . Review of Interfacing Systems Loss of Coolant Accidents (LOCAs)

Recently, events have occurred at several reactor facilities in which low pressure piping connected to the reactor coolant system has been overpressur-ized. The potential for loss of coolant accidents in these " interfacing" low pressure systems had been previously reviewed by NRC. Generic licensing ac-tion was implemented to reduce the risk of this event for pressurized water reactors during the period 1980-1981 because of a dominance of this accident sequence for PWRs in the Reactor Safety Study (WASH 1400). These recent overpressurization events have occurred at boiling water reactors and have resulted, in part, because of multiple component / personnel failures. During

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this inspection, the arrangement, maintenance, and testing of such interfacing systems were reviewed to verify the data used in previous evaluations of the interfacing.LOCA concern and to assess the potential for bypassing of system-components or interlocks' designed to protect against this accident sequenc The-inspector also interviewed selected operations and maintenance personnel to evaluate their understanding of the interfacing LOCA event and the effect

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of certain maintenance or test activities on the potential occurrence of this even . NRC has compiled summary descriptions of interfacing systems at many reactor facilitie This computer data base is used for NRC evaluations of certain accident sequences and was published in NRC NUREG/CR 2069, Summary Report on a. Survey of Light Water Reactor Safety Systems. As part of this inspection,-

- the accuracy of this data base was verified by review of plant drawings and walk down of accessible systems. Inspection findings were as follows:

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The descriptions of low pressure systems connected to the RCS in NUREG/

-CR-2069 were correct. Two interfacing systems not listed in NUREG/CR-2069 (the reactor coolant drain and letdown systems) would be difficult to overpressurize because of their large capacity relief protection which returns directly to the volume control tank with no downstream isolation valve NUREG/CR-2069 lists the design pressure of the low pressure safety in-jection system (LPSI) and the residual heat removal (RHR) system as 875 psig. This number is based on the piping design specification for the installed Class 601 piping at 200 degrees F. The licensee classes LPSI at 600 psi based on Class 601 pipe at 300-500 degrees F. For the RHR system, the design pressure is limited by the RHR heat exchanger design of 500 psi As a result of the 1980-81 NRC and licensee activities related to inter-facing LOCAs, several corrective measures were implemented to prevent potential LOCA events. High/ low pressure boundary valves are leak tested while the plant is shutdown for refueling and subsequent inservice test-ing of high/ low pressure boundary valves is' exempted. The boundary valves-remain closed during plant operation, thereby minimizing the potential for a LOCA. The safety injection boundary check valve leak tests have been incorporated into Technical Specifications 3.14 and Implementation of these tests was reviewed by NRC in inspection 50-213/

84-2 The inspector determined that routine operations and surveillance did not create increased potential for interfacing LOCAs because of the mul-tiple boundary valves which are routinely cycled only during cold shut-down conditions. The' refueling interval leak testing of these boundary valves increases confidence in the integrity of the high/ low pressure boundary. 'This maintenance of double isolation for low pressure systems reduced the checking of operability of the safety injection motor-oper-ated valves (MOVs). Tlase valves, which open to provide emergency core

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. 6 cooling, are no longer routinely tested monthly during plant operatio ' Subsequently, refueling outage checks have been performed with no in-operability identifie One surveillance which may increase the risk of an interfacing LOCA is the six-week interval protection system channel check of the safety in-jection actuation (SIA) logic. During these tests a partial trip signal develops in the SIA logic. Another, spurious trip input could actuate safety injection. If such were to occur, the safety injection MOVs would open leaving the check valve as a single boundary. As stated above, re-fueling interval check valve leak testing-has established confidence in this boundary. The licensee's surveillance procedures provide adequate precaution against inadvertent SIAs and operators and technicians were familiar with the high risk nature of these tests. The licensee has also committed to install pressure sensing interlocks in the safety injection MOV opening circuits to prevent the isolation valves from opening until RCS pressure has dropped below the design pressure of the safety injec-tion syste Maintenance of interfacing systems is conducted in accordance with the licensee's routine quality assurance program, including appropriate documentation, administrative and quality control, and post-maintenance testing. No special requirements are established for interfacing systems, however, the nature of these systems' required operating conditions and the frequency of valve operation and testing necessitates that major maintenance be conducted only during cold shutdown. Boundary leak test-ing is required prior to the following plant startup. As such, the potential for an interfacing LOCA caused by maintenance is low. The in-spector identified that safety injection MOV stroking could occur after tightening of the valve packing when necessar Operators were confident of the safety of this evolution based on the refueling interval testing of the boundary check valves and the need to reestablish operability of the MOV after this maintenanc The inspector also reviewed the licensee's evaluation of IE Information Notice 84-74, Isolation of the Reactor Coolant System from Low-Pressure Systems Out-side Containment, which reevaluated the interfacing LOCA event and found no further potential interfacing system pathways and no surveillance or preven-tive maintenance activities which compromise protective measures for this event. The inspector had no further questions in this are . Followup on Events Occurring During the Inspection 5.1 Licensee Event Reports (LERs)

The following LER was reviewed for clarity, accuracy of the description .

of the cause, and adaquacy of the corrective action. The inspector de-termined whether further information was required and whether there were

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generic implications. The inspector also verified that the reporting ( requirements of 10 CFR 50.73 and Station Procedures had been met, that

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appropriate corrective action had been taken and that the continued operation of the facility was conducted within Technical Specification Limit Multiple Dropped-Control Rods On May 16, 1985 two control rods (Nos. 30 and 33) dropped into the reac-tor core when their rod control group was withdrawn during a routine plant makeup (dilution) operation. Normal dropped rod indications in-cluding two rod bottom lights and rod position deviation ~and dropped rod annunciators were observed. The required automatic turbine runback was in progress. Operators manually tripped the plant by procedure, upon verification of multiple dropped rods. All' plant systems functioned properly:in response to the trip, and the plant was stablized in the. hot standby mode. . Subsequent troubleshooting of the rod drive mechanisms for rods 30 and 33 identified no abnormal operation. The licensee con--

-ducted limited rod exercise testing to verify satisfactory operation of the two rod No problems were found. ^The reactor was restarted on May 16 with all control rods functioning properly. The plant returned.to

. full power operation on May 18, after completing secondary chemistry

. cleanup holds at 5 and 25 percent powe During licensee ' review of rod control maintenance history, .it was deter-

. mined that previous rod failures of these and other. rods had occurred in 1969 and 1980. The licensee has conferred with utility and vendor representitives to determine whether'any common failure mechanism could be identified. No immediate corrective ~ action was identified, however, a plan of further rod testing has been developed and will be implemented during the next available plant maintenance outage. The inspector will

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follow the results of this test program during a subsequent inspection (IFI 213/85-11-01).

7. Exit Interview During this inspection, meetings were held with plant management to discuss the findings. No proprietary information related to this inspection was identifie ' '

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