IR 05000338/1993023

Insp Repts 50-338/93-23 & 50-339/93-23 on 930822-0918.No Violations Noted.Major Areas Inspected:Plant Status, Operational Verification,Maint Observation,Training & Action on Previous Insp Items

ML20059D832
Person / Time
Site:	North Anna
Issue date:	10/15/1993
From:	Belisle G, Taylor D, York J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20059D831	List: IR 05000338/1993023 ML20059D826
References
50-338-93-23, 50-339-93-23, NUDOCS 9311030016
Download: ML20059D832 (2)

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UNITED STATES oa*g#e,aro%*2 NUCLEAR REGULATORY COMMISSION 7 REGION 11 a- S 101 MARIETTA STREET. N.W.. SUITE 2900  ;

$ 9l ATl.ANTA, GEORGIA 30323-0199

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Report Nos.: 50-338/93-23 and 50-339/93-23 Licensee: Virginia Electric & Power Company 5000 Dominion Boulevard Glen Allen, VA 23060 '

Docket Nos.: 50-338 and 50-339 License Nos.: NPF-4 and NPF-7 j Facility Name: North Anna 1 and 2 Inspection Conducted: August 22 - September 18, 1993 I Inspectors: [(// h M J. W. Yo~rk, Acting Senior Resident Inspector

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D. R. Taylor, Resident Inspector

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Approved by-

. Tl c R7d [N G. ( A'. BeTiste,Mection Chief

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Division of Reactor Projects SUMMARY Scope:  ;

This routine inspection by the resident inspectors involved the following '

areas: plant status, operational safety verification, maintenance observation, training, surveillance observation,'and action on previous inspection item ,

Inspections of licensee backshift activities were conducted on the following i days: August 28, 29, and September 8, 9, 10, and 1 Results:

In the plant operations area: l

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Sampling of the spent fuel pool for boron on a seven day interval was identified as an additional conservatism to maintaining the spent fuel pool K effective less than 0.95 (paragraph 3.c.). j Operator communications training concerning repeat back usage was found to be E informative (paragraph 5).

In the maintenance area:

Surveillances observed during the inspection period appeared to be well  !

controlled with procedures being of high quality and informativ Engineering /0perations interface was excellent (paragraph 6).

9311030016 931015 PDR ADOCK 05000338 i G PDR

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Additional information was discussed concerning cracked cladding / base metal corrosion on charging pumps (paragraph 4).

In the engineering area:

One unresolved item was identified regarding compliance with IEEE 279-197 The condition, which involved routing of DC power cables to the underfrequency reactor coolant pump trip circuitry, was identified during the Design Basis Document process. The as-built condition for both units does not comply with the Updated Final Safety Analysis Report and may not meet IEEE 279-1971 requirements for physical separation (paragraph 3.a.).

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In the plant support area:

Three 10 CFR 50.72 notifications were made to the NRC due to state notifications concerning multiple contamination events. The events were caused, in part, by relaxing requirements for respirator usage (paragraph 3.d.).

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REPORT DETAILS Persons Contacted Licensee Employees L. Edmonds, Superintendent, Nuclear Training

• R. Enfin 93r,-Assistant Station Manager, Operations and Maintenance J. Hayes, Superintendent of Operations D. Heacock, Superintendent, Station Engineering

• G. Kane, Station Manager

• P. Kemp, Acting Asst. Station Manager, NS&L

• J. Leberstien, Staff Engineer, Licensing '

W. Matthews, Superintendent, Maintenance J. O'Hanlon, Vice President, Nuclear Operations D. Roberts, Supervisor, Station Nuclear Safety

• R. Saunders, Assistant Vice President, Nuclear Operations D. Schappell, Superintendent, Site Services R. Shears, Superintendent, Outage and Planning

• J. Smith, Manager, Quality Assurance '

A. Stafford, Superintendent, Radiological Protection J. Stall, Assistant Station Manager, Nuclear Safety and Licensing Other licensee employees contacted included engineers, technicians, operators, mechanics, security force members, and office personne NRC Resident Inspectors

• J. York, Acting Senior Resident inspector l

• D. Taylor, Resident Inspector

• Attended exit interview ,

Acronyms and initialisms used throughout this report are listed in the last paragrap . Plant Status Unit 1 operated at or near 100% power the entire inspection perio I Unit 2 began the inspection period at 74% power in a coastdown for i refueling. On September 6, a unit ramp down began and on September 7, l the unit was taken off line. The unit entered Mode 6 on September 1 . Operational Safety Verification (71707)

The inspectors conducted frequent visits to the control room to verify proper staffing, operator attentiveness and adherence to approved procedures. The inspectors attended plant status meetings and reviewed operator logs on a daily basis to verify operational safety and I

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compliance with TS and to maintain awareness of the overall . operation of the facilit Instrumentation and ECCS lineups were periodically reviewed from control room indications to assess operability. Frequent plant tours were conducted to observe equipment status, fire protection programs, radiological work practices, plant security programs and housekeeping. Deviation Reports were reviewed to assure that potential safety concerns were properly addressed and reported. Selected reports were followed to ensure that appropriate management attention and corrective action was applie Under Frequency Reactor Coolant Pump Trip Circuit At 2:10 p.m., on September 2, 1993, the licensee determined that'a nonconforming condition, related to IEEE 279-1971 separation requirements in the Unit 2 Reactor Coolant Pump Underfrequency portion of the Reactor Protection System, constituted an event or condition that alone could have prevented fulfillment of a system needed to shutdown the reactor. The DC power supply cables to the UF sensing relays were routed in such a manner that a single failure could render two of the three circuits inoperabl Specifically, power supply cables for two circuits were in the same cable tray along with other non-safety related cables. The third circuit's power supply cable was physically separated from the other two cables. The postulated single failure is an internally generated short circuit that could adversely affect the two UF relay power supply cables that were run together. Without DC power, the UF relay cannot generate an input signal to the RPS logic. One logic channel associated with the non-separated power cables was declared inoperable and placed in the TRIP conditio The RCP UF RPS trip function then met single failure criteria since the power cables to the remaining two channels were physically separate The inspectors reviewed and assessed the licensees actions and timeliness to resolve this issue. The issue was identified during the licensee's DBD review process. Initially the issue involved classification (safety or non-safety) of the UF and UV reactor coolant pump trips. After a number of consultations with the Architect / Engineering firm, it was determined to be safety-related. In June of 1993, the review indicated that the design may not satisfy IEEE 279-1971. On' July 6, 1993, a PPR meeting was conducted and subsequently, a DR was issued on August 3, 199 The initial DR response indicated that, " Electrical engineering had performed a preliminary evaluation (in accordance with IEEE 379-1977 6.1 (5)) of the potential for a single failure resulting in loss of the 2/3 RCP bus UF reactor trip and concluded that no ,

known failure will cause loss of the circuits. Therefore, based :

on this preliminary evaluation, the circuits remain operable."

The condition was reevaluated prior to the DR close out. The failure mode noted above was identified and a 10 CFR 50.72 report was mad No failure was postulated for Unit l l

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On September 2,1993, the inspectors walked down the DC power supply cables to the UF reactor coolant pump trip circuitry with system engineers and station management. The walkdown included both Unit 1 and 2. The UF channels for Unit 2 were in the previously discussed configuration with one channel tripped. For Unit 1, the power cables for all three channels weie routed for approximately 60 feet through a common conduit. No other cables were routed through that condui l The inspectors independently reviewed the licensee's DR. response and compared it to the UFSAR and IEEE standards. Sections of the UFSAR contradict the as-built conditions. The following are a few I of the contradictions l

- Section 3.1.17, Protection System Reliability and Testability, stated in paragraph 3.1.17.2, that the design l meets the requirements of IEEE 279-1971, " Criteria for i Nuclear Power Generating Station Protection Systems." The redundant logic trains, reactor trip breakers and safety feature actuation relays are electrically isolated and physically separated. Further separation of the channels is maintained within the separate trains. Section 3.1.1 further states that RPS was designed in accordance with IEEE 279-197 Section 3.1.19 Protection System Failure Modes, stated in paragraph 3.1.19.2 that each reactor channel is designed on the deenergize-to-trip principle, so that a loss of power or 1 disconnecting or shorting of a channel causes that channel ]

to go into its tripped mode. The paragraph further states

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that there are certain additional trips which provide input into the reactor trip channel which are designed on the energize-to-operate principle. These inputs are related to anticipatory trips and their operation or failure to operate does not adversely affect the ability of the deenergize-to-trip protection to function. These anticipatory trips are not considered to function in the bases for the safety analysi A review of Accident Analysis, UFSAR 15.3.4, and discussion with the licensee indicated that the UF RCP trip is required in the basis for the accident analysis for a decrease in bus frequency. A review of the trip circuitry also indicated that the UF trip is an energize-to-trip type. The power cables in question are normally energized and power availability was provided by local indicting lamps on the UF trip relay panels. If DC power was lost, neither the failure nor the protection channel loss would be indicated in the control roo The inspectors concluded that the as-built condition for the Unit 1 UF RCP trip channels did not meet the UFSAR description. The

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! inspectors also questioned the use of IEEE 379 as a bases for ;

evaluating this issue. The inspectors noted that both units were l t

committed to IEEE 279-1971 and not committed to IEEE 37 The inspectors held further discussion with the licensee and on September 20, SNSOC approved safety evaluation 93-SE-0T-76 that which documented the following licensee's conclusions:

"IEEE 379 provides guidance for assuring that single failure criterion is not violated where independence cannot be readily demonstrated. Several single failure modes have been evaluated per IEEE 379 and no credible events were identified. Therefore, while the physical separation requirements of IEEE 279 have not been met, no credible single failures have been identified compromising ultimate initiation of a reactor trip for an underfrequency even IEEE 379 permits this type of evaluation."

The failure modes evaluated by the DR included:

- The ESGR is not a missile producing area

- No external fire sources were identified (Appendix R assumes ,

loss of all equipment in the fire area), and an automatic fire suppression system is provide An open circuit will not result in loss of another circui A single grounded conductor has no impact on the 125 VDC ungrounded system. Ground indicating lamps are used to identify grounded conductors requiring procedural actio The 125 VDC operated circuits utilize 600 V cable; therefore, high potential is not a failure mechanis The most likely point of short circuit occurrence is within the UF Aux. Relay Panels, which are individually fuse For a short circuit on one of the circuits, the class 1E 125 VDC circuit breaker will protect the circuit preventing it '

from damaging the other circuit In addition, the licensee indicated that possible changes to the UFSAR and/or the as-built configuration were under review. The acceptability of utilizing IEEE 379 to the address the separation issue remains under review. This issue is identified as URI 50- ^

339/93-23-01, IEEE 279-1971, Review UF RCP Trip Circuitr The inspectors also looked at the timeliness for resolving this issue and reporting the Unit 2 design deficiency in accordance with 10 CFR 50.72. The inspectors considered the time from initial identification to initial resolution reasonable given the issue's complexity and safety significance. The inspectors also noted that the 50.72 notification was not initiated until the day after the Unit 2 channel was placed in TRIP. The licensee indicated that at the time the channel was tripped, the decision on operability was still uncertain and placing it in trip was conservative until operability could be determined. On September 2, the evaluation was completed and the 50.72 report was mad .- .

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b. Employee Concern Program (TI 2500/08)

The resident inspectors were tasked with determining if the licensee had an employee concern program for employees to raise safety issues. The licensee recently developed and implemented a procedure that provided guidance for processing allegations of conditions adverse to safet Procedure VPAP-0204, Allegations of Conditions Adverse to Safety, was put into effect on June 19, ,

1993. This procedure covers potential problems with plant equipment, security, and other subjects, e.g., safe work ,

practices, sexual harassment and inadequate Q The nuclear employee training program made employees aware of this program and instructs them to identify issues to supervisors and to the NRC if their issues are not resolved by the supervisor Both contractor and licensee employees were covered by this procedure. Two safety concerns were received in 1993 before the

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procedure went into effect and one concern was received on September 13. No experience has been accumulated to determine the program's effectivenes Spent Fuel Pool Design The inspectors reviewed information regarding spent fuel pool shutdown margin. The review was conducted in response to a surveillance at Palisades Nuclear Plant which found a sample Boraflex coupon to be nearly empty of Boraflex. North Anna's spent fuel pool contains double wall construction type 304 stainless steel storage cells. The double wall construction provides four vented (open to the pool) compartments in which B C4 (Boraflex) neutron absorber elements are placed for criticality control. Borated material plates were inserted between the two wall s . In addition the spent fuel pool water is maintained, although not required, at greater than 2300 ppm boro The inspector discussed this issue with the licensee to determine if North Anna could have similar problems and what affect this condition would have on K effective. The licensee was aware of the issues with Boraflex, although they did not know if Palisades fuel pool rack construction was similar to theirs. However, the licensee did indicate that no method to assure Boraflex integrity exits at North Anna. The inspectors reviewed the assumptions used in the spent fuel pool's design analysis. The assumptions included the fact that the fuel pool water has no soluble poiso The inspectors noted that procedural controls are in place to maintain the spent fuel pool boron concentration greater than 2300 pp Spent fuel pool sampling was conducted at least once per week per the licensee's nuclear plant chemistry manual. The inspectors concluded that maintaining the boron concentration in the spent fuel pool water greater than 2300 ppm added additional

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conservatism to the design requireme-t to maintain K effective less than 0.9 I Personnel Contamination Events j On three separate occasions during the reporting period, the licensee made 10 CFR 50.72 reports to the NRC due to state notifications concerning multiple contamination events. The events are described below:

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On September 12, three workers were contaminated while !

performing a modification to the "A" reactor coolant loo Readings were 6000 to 8000 DPM. The individuals were successfully decontaminated and exposure dose was negligibl On September 12, four contract employees were contaminated during RTD bypass removal modification in the "B" reactor coolant loop rro On September 15, four contractors became contaminated on their faces during pipe insulation remova On September 13, the inspectors discussed the first two notifications with the licensee and attended a meeting with the licensee in which the events were discussed with the workers involved. The licensee indicated that in order to save total man-rem exposure, resp'.rators were not being used for workers not directly involved with cutting the RTD pipe. The licensee estimated a 20 - 40% man-rem reduction due to increased worker efficiency when working without respirators. The licensee also noted that no major airborne problems existed during the Unit 1 RTD bypass elimination project. Not using respirators results in an increased risk for fascial contaminations. During the meeting, the licensee stressed good radiological practices to prevent i future personnel contaminations. The licensee's policy was not to allow a worker to go back to work following a contamination event until a meeting is held between the individual (s) and the radiological protection superintendent. The inspectors considered the licensee to have an aggressive approach towards minimizing personnel contamination event ,

No violations or deviations were identifie . Maintenance Observation (62703)

Station maintenance activities were observed / reviewed to ascertain that the activities were conducted in accordance with approved procedures, regulatory guides and industry codes or standards, and in conformance with TS requirement Cladding on Charging Pumps

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During the last inspection period, the inspectors documented the cladding repair to the discharge side of 1-CH-P-lC. Charging pump 1-CH-P-1B was then visually inspected and no base metal attack was note Further discussion about the charging pumps' past history revealed that 1-CH-P-1A showed a cracked cladding / base attack when inspected in late 1989. However, unlike 1-CH-P-lC, 1-CH-P-1A's attack area was located in the charging pump's suction. The licensee did not remove this indication, but replaced the clad pump casing with a stainless steel casing. The extent of the base metal attack was not determined. Vendor drawings indicated that pumps manufactured after 1975 had stainless steel casing The areas where t'ne cracked cladding / base metal attack have occurred were away from moving parts. In one case at North Anna the attack was in a low velocity / low pressure area and in the other incident the attack was in a high velocity /high pressure area. Thus, hydraulic effects could not be used to determine whether or not the cracked cladding / base metal attack will occu It is not certain whether the cracked cladding and exposed base metal areas were present during the original manufacturing process or that the cladding cracks formed and propagated during pump operation. The licensee will evaluate the need for future inspections and a recommended time period after the Unit 2 pump inspection later this yea During this inspection period, the licensee conducted an industry event or similar problem survey. This survey revealed that one utility had experienced cracked cladding and resulting carbon steel base metal corrosion attack on two charging pump These casings were replaced with stainless steel casings. Another potential example occurred in May 1981. This later example quoted deteriorated cladding along with other conditions such as breaking shafts, leaking seals,etc. This entire pump was replaced. The inspectors will follow this area during the Unit 2 charging pump inspection !

No violations or deviations were identifie l Training Communications Training On September 1, 1993, the inspector observed Licensed Operator Requalification Program, Cycle 93-5, Communication Review trainin This training was conducted by the training department to reinforce the importance of full communication repeat backs. The instructors discussed actual events like starting a CCP with an operator still in the breaker cabinet, DR N-93-1217, and specific problems regarding command and control in the simulator. This training briefly focused on increasing the operator's communication skills, reviewing the three

- communication models (ideal model, realistic model, and the two way model), and conducting a 30 question communication surve The survey

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was designed to provide a starting point for examining various verbal communication issues on which the operators are routinely trained and evaluated. The inspector found this training to be nost informativ No violations or deviations were identifie . Surveillance Observation (61726)

The inspectors observed / reviewed TS required testing and verified that tesSng was performed in accordance with adequate procedures, that test instrumentation was calibrated, that LCOs were met and that any deficiencies identified were properly reviewed and resolve AMSAC Functional Test On September 6, the inspectors observed parts of 2-PT-36R, AMSAC Functional Test from the AMSAC relay cabinet in the Unit 2 Instrument Rack Room. The procedure tests the programmable logic controllers in pairs to satisfy the functional test logic requirements. During the test performance, the AMSAC functions were verified to have initiated by observing the applicable components changing state. For example, breakers for the motor driven AFW pumps were observed to close and the turbine driven AFW pump steam admission valves to open. Additional AMSAC functions included tripping SG blowdown and sample valves, and opening the rod drive set motor supply breakers. No deficiencies were identified during the test and the procedure appeared to be well written, Safety Injection Functional Test ,

On September 8, the inspectors observed part of the periodic test 2-PT-57.4, Safety Injection functional Test, with the latest procedural revision dated September 3,1993. This test is performed at least once every 18 month '

There were several purposes for performing this test. These were to verify that all related SI and Phase A isolation valves and equipment responded as required by TS when an SI signal is manually initiated in the control room; to test the swap over from SI to cold leg recirculation on RWST low level; and to test the switch over from VCT to RWST on low VCT leve The inspectors reviewed part of the procedure, observed part of the pre-job briefing, and observed part of the test in progress from both the main control room and the ESG Several discrepancies were identified during the test. One discrepancy involved an erroneous position indication for valve, 2-HV-2306 The valve was verified as open by an operator but it displayed a closed position indication in the control room. Another discrepancy relating to a failed relay involved 2-CH-P-1A failing to trip as required when 2-CH-P-1C auto started. The affected

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circuit will be retested before startup from the outag Interface between operations and the system engineers was excellent. No discrepancies were identified by the inspectors for the manner in which the test was conducte No violations or deviations were identifie . Action on Previous Inspection Items (92701, 92702)

(Closed) IFI 50-338,339/92-14-02, Control Room Habitability Test ,

Criteri This item was opened due to previous questions concerning the control room pressurization system's ability to maintain a positive pressure during testing and testing criteria for HEPA Filter and charcoal filter efficiencies. For the control room pressurization system question, the inspectors reviewed 0-PT-76.4, Control Room Bottled Air Pressurization Test, completed on August 31, 1993. The inspectors verified that TS requirements for supplying 340 CFM of air to maintain the control room greater than 0.05 inches W.G. for at least 60 minutes was me Regarding charcoal and HEPA filter tests, the inspectors reviewed the UFSAR and TS and confirmed that testing is performed in accordance with bot . Exit (30703)

The inspection scope and findings were summarized on September 21, 1993, with those persons indicated in paragraph 1. The inspectors described the areas inspected and discussed in detail the inspection results listed below. The licensee did not identify as proprietary any of the material provided to or reviewed by the inspectors during this inspectio Dissenting comments were not received from the license Item Number Description and Reference (0 pen) URI 50-339/93-23-01 IEEE 279-1971, Review UF RCP Trip Circuitry (paragraph 3.a).

(Closed) IFI 50-338,339/92-14-02 Control Room Habitability Test Criteria (paragraph'7).

9. Acronyms and Initialisms AFW Auxiliary Feedwater AMSAC ATWS Mitigating System Actuation Circuitry ATWS Anticipated Transient Without Scram CCP Component Cooling Pump CFM Cubic Feet Per Minute CFR Code of Foderal Regulations DBD Design Basis Document DC Direct Current DPM Disintegrations Per Minute

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DR Deviation Report ECCS Emergency Core Cooling System ESGR Emergency Switchgear Room HEPA High Efficiency Particulate Air IEEE Institute of Electrical and Electronics Engineers IFI Inspector Follow-up Item .

LCO Limiting Condition for Operation -

NRC Nuclear Regulatory Commission PPM Parts Per Million PPR Potential Problem Report PT Periodic Test QC Quality Control RCP Reactor Coolant Pump RPS Reactor Protection System RTD Resistance Temperature Detector RWST Refueling Water Storage Tank SG Steam Generator SI Safety Injection SNSOC Station Nuclear Safety Operating Committee TI Temporary Instruction TS Technical Specification UF Underfrequency UFSAR Updated Final Safety Analysis Report URI Unresolved Item UV Undervoltage V Volts VCT Volume Control Tank VDC Volts Direct Current VPAP Virginia Power Administrative Procedure .

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