ML20133G549

Notice of Violation from Insp on 850617-28

ML20133G549
Person / Time
Site:	Millstone
Issue date:	10/11/1985
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20133G523	List: ML20133G519 ML20133G549 ML20133G579
References
50-245-85-15, NUDOCS 8510160024
Download: ML20133G549 (10)

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r APPENDIX A NOTICE OF VIOLATION Northeast Nuclear Energy Company Docket No. 50-245 Millstone Nuclear Power Station, Unit 1 License No. DPR-21 j As a result of the inspection conducted on June 17-28, 1985, and in accordance

.with the NRC Enforcement Policy (10 CFR 2, Appendix C), published in the Fed-eral Register on March 8,1984 (49 FR 8583) the following violations were iden-tified:

1 A. Technical Specification 6.8.1 states in part: " Written procedures shall be established, implemented and maintained covering activities...recom-mended in Appendix "A" of Regulatory Guide 1.33, February,1978.. ." This appendix requires that maintenance that can affect the performance of safety related equipment should be properly performed in accordance with

written procedures appropriate to the circumstance.

Contrary to the above requirements, two examples of a violation were identified as follows

1. As of June 28, 1985, written calibration procedures were not established for the calibration of safety related pressure switches PS-2-16, PS-2-20, PS-2-54 and PS-2-56 which provide low suction pressure, start prohibit, interlocks for the safety-related Feedwater Coolant Injection (FWCI) pumps A and B.

2. On June 18, 1985, it was identified that the self-aligning rod end bushing on a mechanical snubber, pipe support ICHR-8, of the safety-related isolation condenser had slipped partially out of the rod end bushing housing, resulting in an inoperable isolation condenser system until repairs were made. A subsequent review of procedure MP 739.6, " Mechanical Snubber Visual Inspection" revealed that the pro-cedure was incapable of identifying the misaligned rod end bushing.

, The above two examples constitute collectively a Severity Level IV violation. (Supplement I)

8. Technical Specification 16.12.2 states that locked doors shall be provided to prevent unauthorized entry into each high radiation area in which the intensity of radiation is greater than 1000 mrem /hr. The licensees admi-nistrative procedure (SHP) 4906, " Posting Radiological Controlled Areas",

, Section 8.4 requires that all high radiation areas with general area dose rates greater than 1000 mrem /hr shall have all entrances locked or shall be continuously quarded,to prevent unauthorized entry into those areas.

8510160024 851011 ,

i PDR ADOCK 05000245 G PDR 0FFICIAL RECORD COPY CAT MILLI 85 0001.0.3 10/10/85 l

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Appendix A 2 Contrary to the above requirement, on June 27, 1985 at approximately 1:00 PM, the door to the Scram Discharge Volume (SDV) high radiation area had been left unlocked and unattended. This area was posted with signs on the door "High Radiation Area, RWP Required for Entry" and " Door Must Be Locked At All Times". A survey of the area disclosed " hot spots" of 2000 mrem /hr. Further, if a reactor scram occurred the general area radiation would be greater than 1000 mrtm/hr.

This is a Severity Level IV violation. (Supplement I)

Pursuant to the provisions of 10 CFR 2.201, Northeast Nuclear Energy Company is hereby required to submit to this office within thirty days of the date of the letter which transmitted this Notice, a written statement or explanation in reply, including: (1) the corrective steps which have been taken and the results achieved; (2) corrective steps which will be taken to avoid further violations; and (3) the date when full compliance will be achieved. Where good cause is shown, consideration will be given to extending this response time.

l 0FFICIAL RECORD COPY CAT MILL 1 85 0001.1.0 10/10/85 l

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TABLE OF CONTENTS EXECUTIVE

SUMMARY

DETAILS 1.0 Persons Contacted

2.0 Scope of Inspection

' t i 2.1 Objectives 2.2 Inspection Items i 2.3 Conduct of Inspection 3.0 Recovery of Offsite Power 4.0 Safety / Relief Valves and Manual Depressurization ,

4.1 System Description

4.2 Manual Depressurization j 4.3 Availability of Safety / Relief Valves 5.0 Feedwater Coolant Injection (FWCI) System

5.1 System Description

j 5.2 Assessment of FWCI Operation 5.3 Assessment of FWCI Initiation and Availability of i

Relays and Initiating Components 5.4 Availability of FWCI System Pumps

] 5.5 Availability of FWCI Makeup Water l 6.0 Availability of Emergency AC Power System i

! -6.I System Description

6.2 Gas turbine Operations j- 6.3 Gas Turbine Availability j 6.4 Availability of Emergency AC Bus Breakers / Relays 7.0 Availability of Isolation Condenser

7.1 System Description

7.2 Equipment Availability 7.3 Visual Inspection 7.4 Operations Simulation 8.0 Administrative Controls l

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Table of Contents 2 9.0 Human Factors Engineering 9.1 Equipment / Facility Identification 9.2 Protective Covers for Relays 9.3 Other Emergency Measures 10.0 Facility Tours 10.1 Areas Toured 10.2 Findings 11.0 Unresolved Items 12.0 Exit Meetings Attachment

EXECUTIVE

SUMMARY

This report documents the results of an announced team inspection performed at the Millstone Point Unit 1 Nuclear Power Plant from June 17, 1985 through June 28, 1985. The inspection examined plant hardware and the operational activities relating to a dominant accident sequence identified in NUREG/CR-3085,

" Interim Reliability Evaluation Program (IREP): Analysis of the Millstone Point Unit 1 Nuclear Power Plant". This probabilistic risk assessment (PRA) study identified the key components and activities which could contribute signift-cantly to the core melt accident sequences or mitigate the consequences of such events. The insights and details of one dominant accident sequence in the IREP study were utilized to develop and conduct the inspection.

f Objectives The inspection objectives were to assess: 1) the availability of selected equipment identified as important contributors to the cause of the accident or important to prevent / mitigate the selected accident sequence in the IREP, and

2) the ability of the plant staff to respond to and recover from the accident sequence.

To assess the integral aspects of the plant status and operations, the com-

plementary station programs, not addressed in the IREP study, were also 1

evaluated. They included administrative controls, implementation of training and QA/QC programs, station initiatives on safety measures, and human factor engineering.

Application of IREP Study Results To increase the effectiveness of the inspection effort and concentrate on risk, the most dominant accident sequence was selected from the Millstone Unit 1 IREP study and the inspection items were identified accordingly.

The selected accident sequence in the IREP study identified the specific plant features and activities important to the accident initiating event, preventive and success features, contributing factors to the consequences, and necessary recovery actions required to mitigate the accident sequence. The items inspected and the conduct of the inspection were derived from details of the selected dominant accident sequence, "A Loss of Normal AC Power".

The initiation of the selected accident sequence was triggered by a loss of normal station power, followed by a failure of a safety / relief valve to reseat after opening. The contributing factors to the core melt consequence were the failures of the Feedwater Coolant Injection (FWCI) into the core and of the operator to manually depressurize the reactor coolant system, which would permit the low pressure coolant system to operate. These failures of the FWCI

Executive Summary 2 system and of manual depressurization will prevent coolant coverage of the core, resulting in core melt down in hour. To mitigate the consequence of the initiating and contributing events, it would be necessary to recover the offsite power within h hour. The sequence frequency was estimated to be 7E-5/ reactor year, according to the IREP study, and it accounted for 23% of the total core melt frequency, thus dominating all other accident sequences. It should be noted that the plant feature and activities selected for inspection have a greater impact on safety than this sequence freqency suggests, as they are common to the majority of all dominant sequences identified in the IREP.

Availability of Equipment The inspection rationale to evaluate the availability of plant equipment was to assure that they would be available to operate in accordance with their intended safety functions should their services be demanded. Accordingly, the plant maintenance program relative to the selected equipment, was inspected to assure that the preventive measures, corrective maintenance, routine work controls (including jumpers, tagging and work orders), and periodic surveillance were 2

addressed and performed effectively. The conduct of inspection, thus, included evaluation of the station maintenance activities to ascertain that it is per-formed adequately and effectively in accordance with the prescribed written procedures, and that generic problems and recurring failures of the equipment were adequately addressed in the station maintenance and surveillance programs.

To assess the implementation of the programs, "AS FOUND" states of the equipment were evaluated by performing "walkthroughs" visual inspection, witnessing of test-in progress and surveillance simulations. The effectiveness of the pre-ventive and corrective maintenance measures were evaluated by reviewing appro-priate work records and the performance trend of the equipment, as well as con-trols and measures for fire prevention and environmental qualification.

Plant Operations and Recovery Actions The Millstone-1 IREP study provided insights related to plant operations, and identified that operator recovery actions or errors made during the course of an accident were very important to the successful recovery from an accident.

Thus, to assure plant safety, a high degree of equipment availability must be complemented by the ability of the plant staff to respond and recover from accidents.

Plant operations were evaluated to ascertain that operators were familiar with the plant equipment and the associated plant procedures during normal, abnormal and emergency situations. The operation of plant systems and equipment identi-fied in the selected accident sequence was demonstrated by the plant staff during "walkthrough" simulations of recovery actions. The operators were evaluated for their ability to utilize the control room indications, to under-i

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Executive Summary 3 l

l l stand automatic features under design-base operations, their knowledge of

operating procedures and to operate the plant equipment manually. Local or i i alternate train operation was also simulated when normal operation or recovery l action failed due to equiement failures. Control room operations were assessed i s to assure that proper symptom-oriented emergency operating procedures were j available and capable of being effectively used during the accident situation f i and under stress. Also, station operating procedures were verified to be '

j technically correct and clearly identified important operating instructions.  ;

1 The foregoing inspection logic is depicted in Figure 1.

Findings 1

j The inspection findings demonstrated that the plant programs designed to assure i i hardware availability were adequate and the plant staff exhibited an excellent j knowledge of plant operations, equipment and procedures. The plant staff, l during responses to simulated events / activities, readily demonstrated their knowledge of procedures, physical locations of equipment and familiarity with r i overall plant operations. The operator responses were indicat;ve of the j effectiveness of training on procedures and equipment.

I l The high degree of equipment availability and operational readiness is

! indicative of effective management and administrative controls.

l 1 The equipment identified by the IREP accident sequence was found to be included '

) in the Plant " quality" list and controlled under " safety-related" standards, i

The physical plant was well maintained and clean. This indicates that there is I a positive management and staff attitude.

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l The inspection team identified two violations and made observations in j the following areas. l

Human Factors I

• Equipment Identification i

The plant equipment and their status were identified by color-coded status tags and alpha-numeric identifications, and the areas were i

posted with information, precautions, and signs, related to fire j doors, radiological warnings and security zones. The plant in j general exhibited excellent posting and designations. ,

1 l The following exceptions were observed:

1 The emergency condensate transfer pump did not have any iden-tification.

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Three different equipment identification systems are used, which  !'

result in discrepancies in the equipment identification between '

procedure OP337 and electric plugs on CRP 932-F. The three i systems used are station identification numbers, Ebasco's (the -

j architect / engineer), and General Electric drawings. '

i Control Room Indication

! The control room indication for the isolation condenser level is l given in " feet" and the procedure OP 307 is in " inches".

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• Inadvertent Closure of Vital Breaker i

i During battery charger alarm circuit modification in the control room j on April 17, 1985, a production test electrician unknowingly bumped a .

4160 volt tie-breaker in rear panel 908, relay 110-RX, with his leg, closing the breaker. The eleven GE HGA relays, including the 110-RX, are not provided with protective covers to prevent such an event, j The licensee initiated a work order to install the protective covers.

Potential Isolation of Feedwater Coolant Injection Pumps l The Millstone Unit 1 IREP study identified that failure of FWCI pump i pressure permissive switches would prevent feedwater coolant injec-tion and contribute to the core melt accident sequence during the loss of normal AC power accident.

Suction pressure switches PS-2-16 and PS-2-54 provide low suction i pressure, start prohibit interlocks for FWCI pump A; pressure j switches PS-2-20 and PS-2-56 provide the same interlocks for FWCI e I pump B. A single failure of any one of two pressure switches will

! prevent starting of the associated FWCI pump, and therefore, result

! in a loss of FWCI.

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4 These pressure switches are required to be calibrated every refueling ,

! outage, and the calibration data is recorded using " Calibration r j Report" data sheets. This " Calibration Report" is merely data '

3 logging sheet and no written procedure is provided controlling the l calibration activities during which the pressure switches must be j isolated by closing the switch isolation valves. Upon completion of 1 the calibration, the pressure switches must be valved in. If not,  !

! the associated FWCI pump would be isolated and would not start even  !

if the FWCI pump actuation was initiated. The lack of procedural

! control over these devices increases the probability of failure to return the systems configurations to operability status and the j potential for error is further enhanced because there is no control

room nor local indications provided for the status of the pressure

! switch isolation.

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Executive Summary 5 i

This failure to provide a written procedure for safety-related activities is one example of a violation discussed in the report.

Facility and Equipment Surveillance Isolation Condenser Snubber During a visual examination of snubbers on June 18, 1985, the NRC i staff noted that the self-aligning rod end bushing on pipe support ICHR-8 for the isolation condenser, had partially slipped out of the rod end. This bushing connects the snubber to the rigid pipe support via a ball and clevis pin. With the bushing partially disengaged from the snubber, the load carrying capability of the snubber during a seismic event or water hammer is reduced. Subsequent inspection identified two other snubbers, ICSN-4A and 48, that had an excessive-ly large clearance between the bushing and the clevis.

Review of surveillance procedure, MP739.6, " Mechanical Visual Inspec-i tion", indicated that the inspection procedure was incapable of l identifying the problems with the rod end bearings. This inadequacy in the surveillance procedure is another example of a violation discussed in the report.

• Motor Operated Valve (MOV) Surveillance

On May 15, 1985, a MOV housing on the Low Pressure Coolant Injection i

(LPCI) system fell off. Based on a review of Plant Incident Report

' (PIR) No. 19-75 and discussions with licensee representatives, the staff noted that the station procedure for LPCI operability surveil-lance did not specify sufficient instructions to identify the cracked or loose parts during " walk-down" visual inspections. The licensee has committed to revise the surveillance procedure to clarify the vistal inspection procedures.

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• Radiological Controls During a routine plant tour on June 27, 1985 at approximately 1:00 3

p.m. the NRC staff observed that the door to the Scram Discharge Volume (SDV) high radiation area was unlocked and unattended. In accordance with the station procedure SHP 4906 and Technical Specifi-cation 6.12.2, this area was posted with signs, "High Radiation Area, RWP Required for Entry" and " Door Must Be Locked At All Times".

The failure to follow written procedure and thus to maintain the high radiation area access door with a lock or guard constitutes a violation.

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Executive Summary 6 Conclusion The NRC staff's conclusion was that the plant equipment was in a high state of operational readiness and could be relied upon to respond to accidents.

• The staff concluded that the plant staff was well saalified and, with a high probability, could recover from an accident.

The licensee could improve certain procedures to assure specific equipment availability by taking prompt corrective actions to the staff observations in the findings paragraphs above.

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