IR 05000271/1998001

Insp Rept 50-271/98-01 on 980125-0314.No Violations Noted. Major Areas Inspected:Operations,Engineering,Maint & Plant Support

ML20216H342
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	04/13/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20216H325	List: IR 05000271/1998001 ML20216H320
References
50-271-98-01, 50-271-98-1, NUDOCS 9804210115
Download: ML20216H342 (29)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket No.

50-271 Licensee No.

DPR-28 Report No.

98-01 l

Licensee:

Vermont Yankee Nuclear Power Corporation Facility:

Vermont Yankee Nuclear Power Station Location:

Vernon, Vermont Dates:

January 25 - March 14,1998 Inspectors:

William A. Cook, Senior Resident inspector Edward C. Knutson, Resident inspector

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John R. McFadden, Radiation Specialist, Region I William A. Maier, Emergency Preparedness Specialist l

Russell J. Arrighi, Resident inspector, Pilgrim NPS

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Curtis J. Cowgill, lil, Chief, Projects Branch 5 l

Division of Reactor Projects

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9804210115 980413

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PDR ADOCK 05000271 G

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EXECUTIVE SUMMARY i

Vermont Yankee Nuclear Power Station NRC Inspection Report 50-271/98-01 This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a seven week period of resident inspection; in addition, it includes the results of announced inspections by a regional radiation protection specialist and an in-office review of procedures by a regional

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emergency preparedness specialists.

Operations Failure to perform four-hour estimates of offgas flow rate following the loss of AOG instrument power supply ES-OG-C at 9:35 a.m. on March 1, until 7:50 a.m. on March 2,

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was a violation of TS Table 3.9.2. However, this condition was of minimal safety significance, in that the plant operated at steady state conditions during this period and that AOG operation is routinely monitored during operator rounds. Accordingly, this licensee identified and corrected violation was treated as a Non-Cited Violation, consistent with Section Vil.B.1 of the NRC Enforcement Policy. (NCV 98-01-01)

VY reported a potential un-analyzed condition involving the susceptibility of the reactor building closed cooling water (RBCCW) system to a high energy line break (HELB).

Specifically, the licensee postulated that the consequential failure of the RBCCW system following a HELB could compromise primary containment integrity. The VY staff's identification of the RBCCW system's design vulnerability reflects well on the depth of review of their Design Basis Documentation Program By the conclusion of the inspection period, VY concluded that the current RBCCW system

is within the original design and licensing basis of the unit, and that no compensatory or corrective actions were required to assure system operability. Final resolution of this issue will be tracked via inspector follow-up item IFl 98-01-02.

VY appropriately reported the results of a completed analysis which verified the design vulnerability of the standby gas treatment system to over-pressurization as a result of a design basis accident, during routine containment vent and purge operations. Interim corrective actions were determined to be appropriate and timely.

Maintenance The decision to continue troubleshooting the T-3A startup transformer intermittent ground while at power was appropriate and consistent with Technical Specifications. The transformer outage plan included appropriate administrative controls on the duration of the i

activity; prerequisites and operational limitations were well thought out, and contingency

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actions were adequately addressed. Workers were methodicalin the performance of ii

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tasks, and industrial safety and real-time housekeeping were noted to be strengths.

However, the inspector observed a few instances where the VY staff's performance did not meet management's expectations, in that: test data, which were consistent with an (

intermittent low ground was not initially recognized; one operations shift crew was not

fully cognizant of the extent of maintenance restoration; and post-maintenance ground monitoring, as specified in the maintenance plan, was initially overlooked.

As a result of a question posed during training, VY determined that two out of six channels of the APRM downscale trip function were not being tested in accordance with TS Table l

4.1.1. The failure to test APRM channels "B" and "E" downscale trip functions represented a violation of TS surveillance requirements. Once identified, the affected channels were declared inoperable and the surveillance procedure was promptly revised to test their function weekly. This licensee identified and corrected violation was not cited.

(NCV 98-01-04)

Enaineerina The VY staff's discovery and initial dispositioning of the high pressure coolant injection and reactor core isolation cooling systems' vacuum breaker design vulnerability was appropriate. The retraction of the January 15,199810 CFR 50.72 notification appears to have been poorly founded, based upon the subsequent review of the November 30,1971 letter to the NRC staff defining the vacuum breaker installation and updating the licensing and design basis. (The inspector notes that the retraction was subsequently withdrawn on March 23,1998.) The resolution of this "old design issue" within the 10 CFR 50.59 framework warrants continued inspector follow-up. Inspector follow-up item IFl 97-12-02 remains open.

Plant Sucoort The implementation of the solid radioactive waste program was managed effectively. The annual volume of generated radioactive waste showed a continuing decreasing trend and was relatively low compared to the industry median value.

Overall, the program for the transportation of radioactive materials and its related activities and the training program for these activities were being implemented effectively.

Additionally, the licensee planned to correct a weakness in that there was not any formal l

training and user documentation for their computer program for determining radioactive waste classification and DOT shipment classification for radioactive material.

The Quality Assurance audits and surveillance reports were thorough, programmatic, and well documented. However, self-assessment activity was minimal.

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TABLE OF CONTENTS EX EC UTIVE SUMM ARY.............................................. ii Summary of Plant Status

............................................1 1. Operation s..................................................... 1

Conduct uf Operations.................................... 1 01.1 Response to Advanced Offgas System instrument Power Supply Failure

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Miscellaneous Operations issues............................. 2 08.1 (Open) Inspector Follow-Up item 98-01 -02................. 2

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08.2 (Update) Violation 97-06-03: Failure to take effective corrective

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action for standby gas treatment (SBGT) system potential over-pressurization..................................... 4 ll. M ai nte n a nce................................................... 5 M1 Conduct of Maintenance................................... 5 M 1.1 Maintenance Observations

............................5 M 1.2 Surveillance Observations............................. 6 M1.3 Maintenance to Eliminate Intermittent Low Electrical Ground on Bus 5...........................................7 M1.4 Incomplete Surveillance Identified for the Average Power Range Monitoring System................................. 10 111. Enginee ring.......................................

......... 12 E2 Engineering Support of Facilities and Equipment.................. 12 E2.1 (Update) Inspector Follow-Up ltem 9 7-12-0 3............... 12

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E2.2 Additional Instances of inadequate Electrical Cable Separation... 12 '

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E8 Miscellaneous Engineering issues............................ 13

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E8.1 (Update) Inspector Follow-Up item 9 7-12-02............... 13 IV. Plant Support

................................................15 R1 Radiological Protection and Chemistry (RP&C) Controls............ 15 i

R1.1 Implementation of the Solid Radioactive Waste Program....... 15

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R1.2 Compliance with NRC and DOT Regulations for Shipping of Low Level Radioactive Waste (LLRW) for Disposal and Transportation of Other

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Radioactive Materials............................... 16 R5 Staff Training and Qualification in RP&C....................... 17 R7 Quality Assurance in RP&C Activities......................... 18 i

P3 EP Procedures and Documentation........................... 19 V. Management Meetings

..........................................20 X1 Exit Meeting Summary................................... 20 iv

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X2 Management Meeting Summary............................. 20 X2.1 Site Visit by NRC Region i Senior Management.............20 X2.2 Pre-Decisional Enforcement Conference

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X2.3 Meeting to Discuss BMOs............................ 20

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X3 Review of Updated Final Safety Analysis Report (UFSAR)........... 21 i

INSPECTION PROCEDURES USED.................................... 22 ITEMS OPENED, CLOSED, AND DISCUSSED.............................. 22 l

PARTIAL LIST OF PERSONS CONTACTED................................ 23

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LIST OF ACRONYMS U SED.......................................... 24 t

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Report Details

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. Summarv of Plant Status l

At the beginning of the inspection period, Vermont Yankee (VY) was operating at 97 percent power, in a gradual power reduction due to fuel depletion (coast down). On February 10, power was reduced to 60 percent for a control rod pattern adjustment. With -

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the exception of power reductions to conduct planned surveillance testing, the plant i

continued coast down operation for the remainder of the inspection period. The 1998 refueling outage is scheduled to commence on March 20.

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1. Operations

Conduct of Operations'

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01.1 Response to Advanced Offaas System instrument Power Suoolv Failure a.

Inspection Scooe (71707)

The inspector reviewed the licensee's response to the failure of an instrument i

I power supply in the Advanced Offgas (AOG) system.

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. Observations and Findinas i

At 9:35 a.m. on March 1, the AOG system "A" hydrogen / oxygen (H2/02)

recombiner automatically shut down on indication of low cooling water flow. When operators investigated, they determined that electrical power to the flow instrument had been lost, and that the cooling water system was operating normally. The "A" H2/02 recombiner was restarted in the manual control mode to bypass the low cooling water flow automatic shutdown feature and restore the AOG system to operation. Electrical and Controls (E&C) personnel were contacted to troubleshoot the problem.

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. Troubleshooting revealed that the cause of the problem was a failed instrument power supply (ES-OG-C). This unit provides power for instruments in both trains of the AOG system, including offgas flow rate and the automatic AOG shutdown feature of the hydrogen monitors. Technical Specification (TS) Table 3.9.2,

" Gaseous Effluent Monitoring Instrumentation," states that the minimum number of operable offgas flow rate instruments is one, and requires that the flow rate be estimated at least once per four hours if the minimum number of instruments is not satisfied. However, this requirement was not recognized until an operations shift

turnover on the morning of March 2. An estimate of the offgas flow rate was first

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performed at 7:50 a.m. based on a review of AOG system operating parameters.

' Topical headings such as 01, M8, etc., are used in accordance with the NRC standardized

. reactor inspection report outline. Individual reports are not expected to address all outline topic F'

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At 9:05 a.m., this estimate was backed up by chemisirv sample results which indicated that the main condenser air in-leakage rate was 16.3 standard cubic feet per minute (SCFM); the normal AOG flow rate is less than 30 SCFM. In addition, at 10:30 a.m., VY declared the hydrogen monitors inoperable, based on loss of their automatic shutdown feature, and initiated daily hydrogen sampling as required by TS Table 3.9.2. This was a conservative operability determination, given that the indicating function of these instruments was not affected by the failed power supply. ES-OG-C was replaced and the AOG flow rate instruments and hydrogen monitors were returned to service on March 3.

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Conclusions Failure to perform four-hour estimates of offgas flow rate following the loss of AOG instrument power supply ES-OG-C at 9:35 a.m. on March 1, until 7:50 a.m. on March 2, was a violation of TS Table 3.9.2. However, this condition was of minimal safety significance, in that the plant operated at steady state conditions during this period and that AOG operation is routinely monitored during operator rounds. Accordingly, this non-repetitive, licensee identified and corrected violation was treated as a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy. (NCV 98-01-01)

Miscellaneous Operations issues 08.1 (Ocen) Insoector Follow-Vo item 98-01-02: Potential dearadation of RBCCW system containment isolation capability a.

Backaround and insoection Scope (92903,71707,93702)

On February 20,1998 the VY staff reported, in accordance with 10 CFR 50.72(b)(1)(ii)(A), (Event No. 33763) a potential un-analyzed condition involving the susceptibility of the reactor building closed cooling water (RBCCW) system to a high energy line break (HELB). The licensee postulated that the consequential failure of the RBCCW system following a HELB could compromise primary containment integrity. The RBCCW system is considered a closed loop inside containment and, therefore, part of the containment isolation provision. The RBCCW containment inlet piping contains a single check valve and the outlet piping contains a remote manual-operated isolation valve. Both the inlet and outlet valves have manualisolations on the containment side which facilitate Appendix J local leak rate testing (air test). The inspector conducted a follow-up inspection to examine the licensee's Basis for Maintaining Operation (BMO) pending final resolution of this postulated design vulnerability.-

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Observations and Findinas The inspector learned that this issue was identified by the VY staff during the Design Basis Documentation (DBD) program review of the RBCCW system. VY i

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staff examination of their original response to information Notice (IN) No. 89-55

" Degradation of Containment isolation Capability by a High-Energy Line Break,"

dated June 30,1989, identified what appeared to be an incomplete resolution of the design concern. Accordingly, Event Report No. 98-0252 was initiated to further evaluate this postulated design vulnerability.

BMO No. 98-05, dated March 5,1998, summarized the VY staff's basis for concluding that the current RBCCW system is within the original design and licensing basis of the unit, and that no compensatory or corrective actions were required to assure system operability. However, BMO 98-05 states that ER No. 98-0252 will remain open to assess the need to possibly upgrade the VY design to

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satisfy the requirements of the Standard Review Plan, Section 6.2.4. Accordingly,.

BMO No. 98-05 does not include a corrective action plan, but simply a documented summary of the VY staff's basis for concluding the as-built RBCCW system satisfies its original design intent.

The inspector observed that the licensee's documented RBCCW system design and.

licensing basis (with respect to containment integrity and isolation capability) was briefly addressed in the Plant Design and Analysis Report, dated November 30, 1996, and in the Atomic Energy Commission's Safety Evaluation, dated June 1, 1971. Accordingly, the inspector noted that there was reasonable basis for the VY staff's conclusion that the RBCCW design was within the current licensing basis.

The inspector also noted that this particular BWR system design was subsequently enhanced via development of the 10 CFR 50, Appendix A, " General Design Criteria" and the NRC's Standard Review Plan (SRP). As summarized in NEDC-22253,

"BWR Owner's Group Evaluation of Containment isolation Concerns," dated October 1982, Section 4.2, " Conclusions," the ownars' group identified three possible options for plant systems which currently do not meet the definition of

" closed system inside containment" as stated in SRP, Section 6.2.4. These options were: (1) modify the system to meet the definition per SRP, Section 6.2.4; (2) add another remote operated isolation valve to comply with GDC 56 for those plants which have only one isolation valve; or (3) justify the current design (one isolation valve) by further analysis. The inspector understood that, at the close of this inspection periodi the VY staff was evaluating options to resolve their current RBCCW system design vulnerability, consistent with available design recommendations. More detailed design engineering staff review was awaiting a walkdown of the RBCCW system piping inside the drywell to assess potential design configuration vulnerabilities.

l Subsequent to the conclusion of the inspection period (March 23,1998, at 3:59 l-PM), the VY staff retracted their February 20,199810 CFR 50.72 notification, j

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based upon their determination that the RBCCW system configuration was withm

the current licensing basis.

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Conclusions The VY staff's identification of the RBCCW system's potential design vulnerability, t

with respect to consequential failure under a HELB scenario, reflects well on the depth of review of their Design Basis Documentation Program. Final resolution of this issue will be tracked via inspector follow-up item. (IFl 98-01-02)

08.2 (Uodate) Violation 97-06-03: Failure to take effective corrective action for standbv aas treatment (SBGT) system ootential over-oressurization a.

Backaround and Insoection Scoce (92903,71707,93702)

As documented in inspection report 50-271/97-06,section E.8.2, the licensee failed to take effective action for an identified SBGT system design vulnerability. The licensee's October 31,1997 response to the October 1,1997 Notice of Violation acknowledged the root cause of the violation as a lack of understanding of the SBGT system's pertinent licensing and design basis. Accordingly, one of a number of corrective actions outlined in their October 31,1997 response was the completion of an analysis, targeted for December 1997, of the potential for over-pressurization of the SBGT system, as a result of a design basis accident, during routine containment vent and purge operations. (This corrective action was likewise reflected in VY's LER No.97-014.) Based upon this completed analysis, on

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February 25,1998, the VY staff reported per 10 CFR 50.72(b)(1)(ii)(B)(reference Event No. 33789) that the SBGT system could be over pressurized under these design operating conditions. The VY staff also reported that additional operating restrictions had been imposed on the containment vent and purge valves, during power operations, to eliminate this design vulnerability. The inspector conducted a

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follow-up of this design issue to verify the appropriateness of interim corrective actions and to determine the licensee's plans for resolution.

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Observations and Findinas

The inspector verified that caution tags had been properly hung on the containment

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vent and purge valves SB-16-19-7A,7B,6A,6B and SGT-6 to prevent unauthorized opening of these valves during power operations. The inspector also learned that the VY licensing staff was expediting the submission of a Technical Specification Amendment to limit the number of hours that the vent and purge valves could be open per year. This operating condition is consistent with other boiling water reactor containment purge valve operating limitations. The inspector also verified that the VY staff was updating LER 97-024 to reflect these newly analyzed conditions.

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Conclusions

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The VY staff appropriately identified and reported the results of a completed -

analysis which verified the design vulnerability of the standby gas treatment system. -Interim corrective actions were determined appropriate and timely. This

violation remains open pending final resolution of this design issue.

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11. Maintenance M1 Conduct of Maintenance l

M1.1 Maintenance Observations

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Insoection Scone (62707)

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The inspectors observed portions of plant maintenance activities to verify that the correct parts and tools were utilized, the applicable industry code and Technical

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Specifications were satisfied, adequate measures were in place to ensure personnel l

safety and prevent damage to plant structures, systems, and components, and to ensure that equipment operability was verified upon completion of post maintenance testing.

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Observations, Findinas, and Conclusions The inspector observed all or portions of the following maintenance activities:

New fuel insoections, observed February 12

The inspector observed licensee inspection of two new reactor fuel bundles, installation of the channels, and movement into the spent fuel pool for storage.

Although these activities were satisfactorily performed, the inspector noted that cleanliness control could be improved. This observation was discussed with reactor engineering personnel and acknowledged. The inspector noted that the new

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channel seating tool, that had been designed and fabricated by VY, provided controlled and positive seating.

Electrical cable installations in the cable spreadina room and control room

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This activity _was ongoing throughout the inspection period, and was the corrective action for numerous cable separation issues that have recently been identified. A problem was encountered on two occasions when these installation activities resulted in contact with the conduits containing the main steam line (MSL) radiation detector cables. These conduits are routed through the cable spreading room to the radiation monitors, which are located in the back of the main control board. The MSL radiation monitors input to the reactor protection system to produce a scram if

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high radiation is detected. On January 23, a half scram was generated by channel

"D" MSL radiation monitor when work to remove fire barrier material in the bottom

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of the main control board resulted in the detector cable being disturbed. (These j

signal cables are known to be susceptible to spurious signals generated by vibration or physical contact.) On February 19, a second instance of vibration-induced signal j

noise occurred, but on this occasion, the signal spike was smaller and only

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produced a high radiation alarm.

VY initiated individual event reports in response to each of these occurrences to initiate corrective action. Following the second event, cable rerouting activities in the cable spreading room were stopped pending implementation of moro extensive corrective action and approval by the Plant Manager. Corrective actions included conspicuous posting of the MSL radiation detector cable conduits with warning signs, and review and approval by the operations manager of each cable rerouting activity that required work to be performed in the cable spreading room. The inspector considered that these actions were appropriate.

M1.2 Surveillance Observations i

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trisocction Scone (61726)

The inspector observed portions of surveillance tests to verify proper calibration of test instrumentation, use of approved procedures, performance of work by qualified personnel, conformance to Limiting Conditions for Operations (LCOs), and correct post-test system restoration.

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Observations Findinas, and Conclusions The inspectors observed portions of the following surveillance testing activities:

Residual heat removal system auarterly surveillance, observed Februarv 6

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No problems were observed during the conduct of this surveillance.

Hiah oressure coolant iniection (HPCI) system ouarterly surveillance,

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observed March 3, and HPCI time to rated flow surveillance, observed March

.12 The time to rated flow surveillance had been attempted during the March 3 quarterly surveillance, but was not completed because of a problem with the test equipment settings. The test requires that the pump has not been run in the l

previous three days, and so the test was rescheduled for March 12. On March 12, the test was completed satisfactorily.

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M1.3 Maintenance to Eliminate Intermittent Low Electrical Ground on Bus 5

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Insoection Scope (62707)

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The inspector reviewed VY's preparations for maintenance to identify and eliminate l

an intermittent low electrical ground associated with bus 5 and startup transformer

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T-38. The inspector also observed various portions of this maintenance activity.

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Observations and Findin.21 l

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Bus 5 is part of the 4160 volt AC on-site distribution system at VY. The bus receives power from the T-3B startup transformer while the plant is shut down, or from the main generator through unit auxiliary transformer T-2 and bus 2, during plant operations at power. Indications of a low electrical ground associated with bus 5 and T-38 were first observed in November 1937, following startup from the forced outage. The existence of a ground fault was a concern because the startup transformer is also the normal power supply to emergency bus 4 when the plant is shut down.

During this inspection period, VY determined that the most likely location of the low

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~ lectrical ground was the cables that connect the T-38 startup transformer output e

to the on-site distribution system. This presented a problem, in that additional l

troubleshooting would require T-3B to be de-energized; due to the system design, this would also cause the T-3A startup transformer to be de-energized. Plant l

operation at power without the startup transformers available is undesirable i

because, were a reactor scram to occur, it would result in a loss of all AC power l_

until the emergency diesel generators started and re-energized the vital electrical l.

system (a period on the order of 10 seconds). On the other hand, deferring further l.

troubleshooting until after shutdown would require use of the startup transformers

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during the shutdown, since the alternate means of receiving off-site power l

(backfeed through the main and auxiliary transformers) requires equipment alterations that can only be performed after the plant has been shut down. Under these conditions, the increased electrical load could cause the fault to rapidly degrade and conceivably result in the loss of both startup transformers due to a

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short circuit.

VY concluded that continued troubleshooting at power was the more desirable course of action. VY Technical Specifications provide no restrictions on continued

power operations with both startup transformers inoperable. However, review of the probabilistic risk associated with removing startup transformers from service indicated that the activity should be controlled in the same manner as a limiting condition for operation (LCO) maintenance activity. Accordingly, VY developed an LCO maintenance plan for the proposed startup transformer outage. This plan described troubleshooting and anticipated corrective actions, and established administrative time constraints on the activit \\

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LCO Maintenance Plan Review As established in the LCO maintenance plan, the maintenance activity would begin by de-energizing both startup transformers to allow T-3B to be disconnected from the 115KV off-site supply. While both startup transformers were inoperable, a 24-hour administrative 1.CO would be in effect. With T-3B disconnected for l

troubleshooting, power would then be restored to the remaining startup l-transformer. In this configuration, a seven-day administrative LCO would be in effect. The T-3B output cables (nine per phase, for a total of 27 cables) would be disconnected from the 4160 volt bus work to allow individual electrical checks to be perfomied. It was anticipated that degraded insulation on one cable was the cause of the low electrical ground. Once this cable was identified, it would be retired in place (disconnected from the bus work and grounded)..One cable from each of the other two phases would also be retired in place, leaving eight cables per i

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phase in service. The licensee completed an engineering evaluation which demonstrated that adequate' current carrying capability would be maintained with as few as seven caoles per phase. Off-site power would again be de-energized, again placing them in a 24-hour administrative LCO, while T-3B was reconnected.

The inspector reviewed VY's LCO maintenance plan for the proposed startup I

transformer outage. Prerequisites for the maintenance included verification that no maintenance was planned by Vermont Electric Power Company (VELCO) that would effect the 345KV/115KV switchyards, or by Vernon Hydro that would effect the reliability of the Vernon tie line; both emergency diesel generators were to be demonstrated operable, and the emergency core cooling and containment cooling I

systems were to be verified operable; and materials necessary to establish backfeed off-site electrical power through the main and auxiliary transformers was to be pre-staged. During the maintenance, no planned activities that would challenge the -

electrical distribution, emergency core cooling, or reactor protection systems were to be performed.

The inspector concluded that the LCO maintenance plan was adequate for the proposed maintenance activity. In addition, the NRC Nuclear Reactor Regulations (NRR) staff reviewed the proposed activity and concluded that it was consistent with Technical Specifications. The inspector did note that the contingencies developed in the written LCO plan, in the event of operational transients, were not

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well developed. VY acknowledged this observation, performed an additional review of possible transients, and identified expected operator responses for the same.

LCO Plan imolementation The two-day startup transformer maintenance outage was commence on March 5.

Prior to the start of work, a crew briefing was conducted by the VY manager-in-

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charge of the activity. Specific aspects of the activity were presented by l

maintenance, operations, and engineering personnel. This briefing was also l

presented to the two other on-shift crews. The inspector observed that the briefing was in-depth and thoroug l

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The startup transformers were de-energized on March 5 at 11:46 a.m. As anticipated, a single cable was identified as having a significantly lower resistance to ground than the other eight. Modification to disconnect this cable, along with one cable from each of the other two phases, was performed as discussed in the LCO maintenance plan and in accordance with minor modification 98-006.

Restoration was completed and both startup transformers were returned to service at 3:51 a.m., March 7.

The inspector observed various maintenance activities over the course of the startup transformer outage. The inspector noted that the workers were methodical in the use of procedures and performance of tasks. Electrical safety precautions were rigorously observed and real-time housekeeping was a strength. The inspector j

observed no indications that time pressure was causing the workers to proceed faster than they considered appropriate. On one occasion, however, the inspector noted that useful information obtained during testing was not recognized by the work crew or cognizant engineer. Specifically, when both startup transformers j

were initially de-energized, megger checks (a test of the cable insulation resistance to ground) were performed on the cabling to both transformers. For T-38, this was done as an initial step in localizing the ground; for T-3A, it was simply an opportunity to check the condition of the transformer and cabling while off-line (T-3A and its associated cabling is not equipped with independent ground detection circuitry).

During the check of T-3B, the technician initially observed intermittent erratic indications, with the meter dipping low. During this test, meter readings are taken at a set interval, and therefore, the intermittent low indications were not recorded.

Test personnel were initially concerned that the battery power supply for the megger was low; however, they knew from experience that a low battery would cause the meter to deflect high rather than low. Because a stable value in the expected range was obtained by the end of the test, there appeared to be no further concern for the initially erratic readings. The inspector observed that the initial erratic readings would be the expected indication for an intermittent ground.

This was apparently not recognized by the VY work crew performing the work, not recorded as a testing anomaly, and only realized after the inspector discussed his observations with the engineering support staff. Although the outcome of this test did not affect the course of the maintenance, it was nearly missed as an early indicator that the troubleshooting was on the right path.

During the restoration phase of the maintenance, the inspector attended an operations shift turnover briefing. The inspector was concerned that the on-shift crew was apparently not aware of the progress of the startup transformer restoration. Specifically, both startup transformers were inoperable (de-energized)

at the time of the brief; by the LCO maintenance plan, all of the cables were to be reconnected prior to de-energizing the startup transformers. However, the operators did not know if the cable connections had been completed. This was a

concern to the inspector, from the perspective of plant operations, since the period

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when both startup transformers were inoperable was the most risk significant period. Also, this was a concern from the perspective of work control, since operations personnel are responsible for ensuring that equipment is maintained in a i

safe condition during maintenance through the use of protective (danger) tags. At j

the time of this observation, operators were preparing to clear danger tags from the T-3B high side (115KV) bushings; although this action was correct for the point that the electricians were at in the restoration, the inspector considered that the operators could not effectively validate the workers' request for tags to be cleared if they were not fully aware of the status of the affected equipment.

To verify that the ground had been eliminated, the LCO maintenance plan indicated that a chart recorder would be used to monitor the ground detection circuit for intermittent grounds that were of such short duration that they would not trip the control room alarm. Such short duration grounds had been detected by this technique prior to the startup transformer maintenance. T-3B was returned to service on a Saturday morning; when the inspector inquired on Monday morning as to the results of the chart monitoring, VY determined that a chart recorder had not been installed. This was a minor oversight in the execution of the LCO plan and i

apparently missed due to the absence of a specific line item sign-off for this activity. When the chart recorder was subsequently installed, no ground indications were detected, indicating that the disconnected cable had been the likely source of i

the intermittent ground. VY station management acknowledged the above instances as not meeting their expectations for proper execution of the LCO plan.

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Conclusions i

The decision to continue troubleshooting the T-3A startup transformer intermittent I

ground while at power was appropriate and consistent with Technical Specifications. The transformer outage plan included appropriate administrative controls on the duration of the activity; prerequisites and operationallimitations were well thought out, and contingency actions were adequately addressed.

Workers were methodical in the performance of tasks, and industrial safety and real-time housekeeping were noted to be strengths. However, the inspector observed a few instances where the VY staff's performance did not meet management's expectations, in that: test data that was consistent with an intermittent low ground was not initially recognized; one operations shift crew was not fully cognizant of the extent of maintenance restoration; and post-maintenance ground monitoring, as specified in the maintenance plan, was initially overlooked.

M1.4 Incomolete Surveillance Identified for the Averaae Power Ranae Monitorina System a.

Inspection Scoce (61726)

l The inspector reviewed VY's determination that two channels of the average power range monitor (APRM) system were not being fully tested as specified by Technical Specifications.

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Observations and Findinas The APRMs provide input to the reactor protection system (RPS) to initiate a reactor l

scram when a potentially unsafe reactor power level is detected. Six channels of

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APRM are utilized in a coincidence circuitry in the RPS, to prevent a single spurious signal from causing an unnecessary scram and to allow on-line testing of individual APRM channels without compromising operability of the RPS. Since an inoperable APRM (due to electrical malfunction or operation at reactor power levels that are

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off-scale high or low) cannot perform its RPS function, the circuitry is designed to insert a scram signal from the affected channel, should these conditions exist.

During licensed operator candidate training, a question arose concerning VY's testing of the APRM downscale scram function. This function is activated when i

the reactor mode switch is placed in the RUN position, and causes a scram signal to be generated if an APRM indication is too low (less than two percent on a 125 percent scale) at the same time that its associated intermediate range monitor (IRM)

is indicating too high (greater than 120 percent on a 125 percent scale). The purpose of this function is to prevent reactor operation without reliable indication of

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l reactor power being provided by the neutron monitoring system. Since this trip requires input from both the APRMs and IRMs, it is only functional while both systems are detecting reactor power (referred to as the overlap region, about 0-20 percent). At higher power levels, the IRM detectors have been withdrawn from the core, and will never generate a high signal. Therefore, the logic required to satisfy the APRM downscale trip cannot be satisfied during normal full power operation.

Technical Specification (TS) Table 3.1.1 requires that the APRM downscale trip be operable when the reactor mode switch is in RUN, and TS Table 4.1.1 requires that function of the output relays be tested weekly. VY had interpreted that this requirement applied only during operation in the overlap region, and therefore considered that weekly testing of the APRM downscale trip was not required during normal full power operation.

Due to the circuitry configuration and pairing of APRMs and IRMs, the downscale trip for four of the six APRM channels were being tested as a result of other tests performed during the APRM weekly surveillance. However, it was determined that two APRM channels ("B" and "E") had separate downscale trip output relays that were not being tested. As a result of this investigation, VY concluded that the APRM downscale trips for these two channels should be declared inoperable. TS Table 3.1.1 specifies that a minimum of two channels of the downscale trip function per trip system are required for the function to be considered operable.

Since this requirement was being satisfied by the existing weekly surveillance, two channels of the function being inoperable did not place them in a limiting condition for operations (LCO) and had no effect on plant operations. The APRM surveillance procedure was subsequently modified to test the channel "B" and "E" downscale trip output relays. This non-repetitive, licensee identified and corrected violation of TS Table 4.1.1 is being treated as a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Poliev. (NCV 98-01-04)

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Conclusions As a resuit of a question posed during training, VY determined that two out of six channels of the APRM downscale trip function were not being tested in accordance with TS Table 4.1.1. The failure to test APRM channels "B" and "E" downscale trip functions represented a violation of TS surveillance requirements.. Once identified, the affected channels were declared inoperable and the surveillance procedure was promptly revised to test their function weekly. This licensee identified and corrected violation was not cited. (NCV 98-01-04)

lil. Engineering

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E2 Engineering Support of Facilities and Equipment E2.1 Lygdge) Inspector Follow-Un item 97-12-03: BMOs oronosed by VY to remain open followina the 1998 refuelina outaae As documented in section X2.3 of this report, the VY staff met with the NRC staff on March 5,.1998 in the NRC headquarters office in Rockville, Maryland, to discuss the BMOs that VY projects will remain open following restart from the 1998 refuel outage. Since this inspector follow-up item was initiated, the VY staff has significantly reduced their projection of open BMOs following restart to a total of 13 issues. These remaining BMOs will be examined further by the NRC staff in subsequent inspection periods. This item remains open.

E2.2 Additional Instances of inadeauate Electrical Cable Seoaration The issue of inadequate electrical cable separation was initially identified by VY in March 1997, as discussed in inspection report 50-271/97-03. This issue was tracked under unresolved item URI 97-03-02, and was later assessed to have been

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the result of inadequate original design specifications and installation practices, as I

discussed in inspection report 50-271/97-12.

During this inspection period, VY identified four additional electrical cable separation issues. These issues were reported as required by 10 CFR 50.72, as one-hour non-emergency notifications of conditions that were outside the design basis of the plant:

-- On February 12, six non-nuclear safety class cables were found to be routed through safety class electrical division Si and Sil raceways (Event No. 33705).

-- On February 24, one non-nuclear safety class cable was found to be routed through safety class electrical division Si and Sll raceways (Event No. 33779).

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- On February 27, an alternate shutdown power cable was found to be routed through control power and instrumentation non-nuclear safety class raceways (Event No. 33808).

-- On March 10,15 non-nuclear safety c' ass cables were found to be routed in a raceway adjacent to a safety class electrical division Si raceway, and then in another raceway adjacent to a safety class electrical division Sll raceway, where the Si and Sil raceways include elements of the same safety systems (Event No.

33870).

These issues were identified cs the result of VY's corrective actions that were initiated in response to the original cable separation issue. The licensee plans to update the original LER (No. 97-06) and the inspector will review the licensee's root cause and corrective actions as a matter of routine LER review and follow-up.

E8 Miscellaneous Engineering issues E8.1 imp _d_a,_te) Insoector Follow-Vo item 97-12-02: Potential condition outside desian basis involvina location of vacuum breakers for the HPCI and RCIC steam exhaust li.fui O

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Backaround and Inspection Scone (92903)

As previously discussed in inspection report 50-271/97-12, section E1.2, the licensee identified that the high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) systems' turbine steam exhaust line vacuum breakers were not optimally located. The vacuum breakers are located outside the torus and take suction from the reactor building via the standby gas treatment system inlet piping.

This location differs from their originallocation in the torus air space. During this inspection period, the licensee concluded that this configuration was not outside the plant design basis, as documented in Basis for Maintaining Operation (BMO) No.

98-01, and the VY staff retracted their January 15,1998,10 CFR 50.72 (b)(1)(ii)(B) report (Event No. 33545). The inspector conducted a review of the licensee's basis for this design determination and their 10 CFR 50.72 retraction.

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Observations and Findinas The inspector examined the Updated Final Safety Analysis Report (UFSAR) and the associated engineering design change documents, internal memorandums, and docketed correspondence. Included in the correspondences reviewed was a November 30,1971 letter from VY to the NRC informing the NRC's Division of Reactor Licensing staff of the details of the addition of a vacuum breaker in each of the turbine exhaust lines from the HPCI and RCIC systems. The stated purpose of this design change was to prevent recurrence of a water hammer effect observed l

I during pre-operational testing of the HPCI syste.

14 The inspector determined that although this design change (prior to issuance of the facility's operating license on February 28,1973) was not explicitly documented in the licensee's UFSAR (reference Section 6.4.1) or included in the simplified piping and instrumentation diagram (Figure 6.4-1), that this HPCI and RCIC design attribute became part of VY's current licensing and design basis via the November 30,1971 docketed correspondence. The inspector verified that VY controlled drawings (Nos. G-191169 and G-191174, " Flow Diagram High Pressure Coolant injection System" and Flow Diagram Reactor Core Isolation Cooling System,"

respectively) did clearly depict the HPCI and RCIC systems vacuum breaker design details.

The inspector examined the engineering design change request (EDCR No. 73-32)

which subsequently relocated, in 1973, the HPCI and RCIC vacuum breakers from inside the torus air space to the reactor building. The stated purpose of the design change was to eliminate a primary containment leakage path (identified via the 10 CFR 50, Appendix J, Type A, integrated leak rate test) from the torus, through the exhaust line vacuum breakers, and then through the turbine exhaust line stop checks and exhaust line check valves. Review of this 1973 EDCR package identified that it contained the requisite installation procedures (including material verification reports, weld data reports, liquid penetrant test reports, and pressure test report) and supporting engineering staff seismic and thermal stress analyses and analysis of impact on the standby gas treatment system's operation. However, in contrast to current engineering standards, the supporting safety evaluation consisted of a single paragraph simply stating that " systems engineering group has shown that: (1) probability of containment leakage is not increased, (2) probability of another malfunction or accident has not been introduced, and (3) the margin of safety as defined in the Technical Specifications has not been reduced."

The inspector discussed these observations with plant management who acknowledged the inspector's findings and agreed to reconsider their 10 CFR 50.72 retraction and the content / conclusions drawn in BMO No. 98-01. Subsequent to the conclusion of the inspection period (on March 23,1998) the licensee withdrew their March 4 retraction of Event No. 33545 after determining that the j

configuration of the HPCI and RCIC systems' vacuum breaker lines was outside the original design basis of the facility.

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Conclusions The VY staff's discovery and initial dispositioning of the high pressure coolant injection and reactor core isolation cooling systems' vacuum breaker design vulnerability was appropriate. The retraction of the January 15,199810 CFR 50.72 notification appears to have been poorly founded, based upon the subsequent review of the November 30,1971 letter to the NRC staff defining the l

vacuum breaker installation and updating t!c !! censing and design basis. The resolution of this "old design issue" within the 'JO CFR 50.59 framework warrants continued inspector follow-up. Inspector follow up item IFl 97-12-02 remains open.

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IV. Plant Support R1 ~

Radiological Protection and Chemistry (RP&C) Controls

"R1.1 Imolementation of the Solid Radioactive Waste Proaram a.

Inspection Scooe (86750-01)

The inspector reviewed the licensee's solid radioactive waste program. Information was gathered through observation of activities, tours of the radiologically controlled areas including the radioactive waste building, reactor building, turbine building, North warehouse, low level radioactive waste storage pad, and outside portion of the protected area, discussions with cognizant personnel, and review and evaluation of procedures and documents.

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Observations and Findinas l

The inspector noted that the solid radioactive waste processing and shipping

- program involved three plant organizations. Operations performed the de watering of resins. Maintenance performed handling and rigging operations and also-decontamination activities. Technical Services, in the form of the radioactive waste organization, coordinated and provided oversight for the de-watering, handling, and storage of radioactive waste and performed the shipping activities for radioactive

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waste. The radioactive waste organization consisted of a supervisor and two

.I technicians, with trained and qualified backups within Technical Services.

In reviewing the implementation of the solid radioactive waste program, the liquid waste processing methods were inspected; the dry active waste (DAW) operation was evaluacd; and the storage locations were inspected. The licensee's Process Control Program (PCP) document (Revision 6) for solid radioactive waste was reviewed and found to be adequate in scope and detail. The licensee's Sampling

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- and Analysis for Radwaste Classification procedure (Revision 5) also was detailed and provided appropriate instructions for waste classification. Liquid radioactive waste was processed through permanently installed process equipment (filters and domineralizers). The high purity, low purity, chemical, and detergent liquid process streams were cross-connected to provide flexibility in processing paths. Currently, all processed water was recycled,'and there was zero release of radioactive liquid effluents. Spent resins and filters were de-watered in high integrity containers (HICs). There was a " green is clean" program to minimize the generation of radioactive waste, and de collected materials were shipped to an offsite vendor to be surveyed prior to free release as non-radioactive waste. Before shipment to a vendor, potentially contaminated materials were sorted based on their ability to be incinerated or compacted in order to maximize volume reduction and to minimize the final volume of radioactive waste. Offsite contracted services were available for equipment / parts decontamination and for super-compaction or incineration of dry active waste. The low level radioactive waste storage pad was a fenced area

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adjacent to the protected area. This fenced pad, along with its different types of modular concrete storage containers, provided an area for storage of low level waste and higher level waste such as HICs filled with spent powder or resin. A large portion of this storage capacity was available at the time of this inspection since the licensee conducted a large number of spent resin waste shipments in 1997. The licensee had additional radioactive material holding areas inside the protected area including the North warehouse and seven sealand containers, all of which were properly secured and posted.

The licensee had established goals for radioactive waste generated, and progress on these goals was tracked. The available trend data indicated that the volume generated over the last ten years had steadily decreased. Data for a recent three year period showed that the licensee's generated volume was below the industry media, value. The licensee conducted an initiative over the last year to eliminate the remaining usage of plastic bags.

Housekeeping in the toured areas was good. Aisle ways were clear and free of clutter / debris, and storage areas were clean and orderly; contaminated areas were minimized; radioactive material was clearly and properly labeled and stored in an orderly fashion.

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Conclusions The implementation of the solid radioactive waste program was managed effectively. The annual volume of generated radioactive waste showed a continuing decreasing trend and was relatively low compared to the industry median value.

R1.2 ' Comoliance with NRC and DOT Reaulations for Shinoina of Low Level Radioactive Waste (LLRW) for Disoosal and Transoortation of Other Radioactive Materials I

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Insoection Scope (86750-01)

The inspector reviewed the licensee's transportation of radioactive meterials.

Information was gathered through observation of activities, discussions with cognizant personnel and review and evaluation of procedures and documents, b.

Observations and Findinas The shipping records for several past shipments of radioactive material and of radioactive waste (including spent resin, DAW, laundry, contaminated equipment,

irradiated hardware, and 10 CFR 61 samples) were reviewed. These records were found to be appropriate and complete. The waste classifications and Department of l

Transportation (DOT) shipment type determinations for these shipments were evaluated and met regulatory requirements. The waste manifests and shipping papers were properly completed. Of the approximately eighty shipments made in j

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1997, approximately thirty were casks of de-watered spent resin / filter material which were sent to a low level waste burial site. These numerous waste shipments were accomplished in a relatively short period of time without a significant incident.

The inspector noted that the licensee used a shipping package, for which the NRC had issued a certificate of compliance (USA /9249/A), to transport de-watered resin (Manifest No.97-033; Low Specific Activity (LSA); Type B quantity; 2.4 R/hr at the unshielded surface) to a burial site without being registered as an authorized user of the cask. Prior registration as a user of this cask was required by 10 CFR 70.12(c)(3) unless specific exemptions in 10 CFR 71.10 (effective on April 1, 1996) applied. In this case, an exemption (10 CFR 71.10(b)(2)) did apply (i.e., for

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LSA shipments which, at 3 meters from the unshielded material, did not exceed 1 rem per hour). This particular exemption became effective on April 1,1996.

However, it was not clear whether or not this unregistered cask had been used by the licensee prior to April 1,1996 for nonexempt shipments. The licensee stated that they would review their records to determine whether the cask had been used and document the results.

Additionally, the licensee routinely used an in-house developed computer program to apply scaling factors, to determine waste classification (10 CFR 61.55), to determine DOT shipping paper parameters such as type of quantity and type of shipment (49 CFR 172.203,172.403, and 172.433(f)), and to determine radioactive waste manifest parameters such as total activity (10 CFR 20.2006).

The licensee stated that a vendor computer program would replace the in-house program in the near future. The licensee further stated that formal initial user training and user manuals would be provided by the vendor. These matters will be reviewed during a subsequent inspection. (IFl 98-01-03)

c.

Conclusigna

. Effective performance was demonstrated in the area of packaging and

~ transportation of solid radioactive waste. Additionally, the licensee planned to correct a weakness in that there was not any formal training and user documentation for their computer program for determining radioactive waste j

- classification and DOT shipment classification for radioactive material.

j R5 Staff Training and Qualification in RP&C a.

insoection Scope (86750-01)

The inspector reviewed the qualifications and training of selected radioactive waste I

personnel. Information was gathered through discussions with cognizant personnel, and review and evaluation of documents.

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Observations and Findinas Training department personnel stated that the training for radioactive waste personnel was provided by a combination of in-house training and training by contractors and that the contractor-provided training courses were reviewed had approved by licensee personnel before implementation. The inspector reviewed the course materials used for the training of the radioactive waste personnel. The scope and depth of the course materials were fully adequate and documented. The inspector verified that these individuals had been recently trained in the aforementioned topics. Additionally, it was confirmed that all five individuals authorized to sign shipping paperwork had received recent and appropriate training on the DOT shipping regulations.

Training personnel were knowledgeable of the required training for radioactive waste personnel and shippers, and the current training status for these individuals was readily available. Therefore, the trCning program for radioactive waste personnel was organized and documented. The inspector also verified that the operations personnel, who operated the rapid de-watering system for resin, were trained and qualified for the operation and in the applicable operating procedure, c.

Conclusions The training program for personnel involved with solid radioactive waste activities and with the transportation of radioactive materials was organized, fully implemented, and well documented.

R7 Quality Assurance in RP&C Activities a.

insoection Scope (86750-01)

The inspector reviewed the licensee's quality assurance (QA) activities for solid radioactive waste managernent and transportation of radioactive materials.

Information was gathered through discussions with cognizant personnel and review and evaluation of documents.

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Observations and Findinas Audit 96-09, conducted in the Fall of 1996, covered radioactive waste j

management and DOT and NRC radioactive material shipping requirements. This

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audit resulted in the generation of four Event Reports (ERs) and eight Recommendations. Audit 97-09, conducted in the Fall of 1997, covered j

radioactive waste activities and the Process Control Program (PCP). This audit i

resulted in the issuance of three ERs and five Recommendations. The inspector's review of the audit reports showed that the audits were thorough and programmatic.

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The inspector reviewed four QA Surveillance Reports, performed in 1996, which covered radioactive waste evolutions and resulted in one ER and two Recommendations. These reports showed that the surveillance activities were detailed and covered the full scope of the evolutions.

The inspector reviewed seven QA Surveillance Reports, performed in 1997, which covered twenty three shipments of radioactive waste or radioactive material. Most of these surveillance activities included an auditor and a technical specialist. The Reports resulted in the generation of two ERs and five Recommendations. The licensee increased the number of QA surveillances in order to effectively assess the radioactive shipping program when the program was stressed by a large number of radioactive waste shipments being made in a relatively short period of time. The licensee conducted only one, limited scope self assessment (August 1997) from July 1996 to January 1998. This self-assessment addressed the green-is-clean program and generated two recommendations. This level of effort represented a minimal self-assessment activity.

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Conclusions The Quality Assurance audits and surveillance reports were thorough, programmatic, and well documented. However, self-assessment activity was minimal.

P3 EP Procedures and Documentation a.

Insoection Scope (82701)

Regional inspectors reviewed several changes the licensee made to the emergency plan and implementing procedures. The inspectors reviewed these changes in the NRC Region I office. They conducted this review to verify that the changes made to the emergency plan and implementing procedures were made in accordane with Part 50.54(q) of NRC regulations, i.e., that they did nat decrease the effectiveness of the emergency plan. Listed below are the changes reviewed:

OP 3506 Revision 35 Emergency Equipment Readiness Check OP 3531 Revision 11 Emergency Call-In Method OP 3504 Revision 30 Emergency Communications b.

Conclusion _g Based on the licensee's determinations that the changes did not decrease the overall effectiveness of the emergency plan, and that the plan, as changed, l

continues to meet the standards of 10 CFR 50.47(b) and the requirements of l

Appendix E to Part 50, NRC approval of these changes is not required. The in-office review of these changes indicated they were made in accordance with 10

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CFR 50.54(q). The implementation of these changes will be inspected on-site l

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during a future emergency preparedness program inspection to confirm that they have not decreased the overall effectiveness of the licensee's emergency plan.

V. Management Meetings X1 Exit Meeting Summary The resident inspsetors met with licensee representatives periodically throughout the inspection and following the conclusion of the inspection on March 31,1998.

At that time, the purpose and scope of the inspection were reviewed, and the preliminary findings were presented. The licensee acknowledged the preliminary inspection findings.

X2 Management Meeting Summary X2.1 Site Visit by NRC Reoion I Senior Manaoement On February 11 and 12, the Deputy Regional Administrator, Region I, and members of the Region I staff conducted a visit of the VY facility. The visit consisted of a tour of the unit and numerous interviews and discussions with members of the VY staff and management.

X2.2 Pre-Decisional Enforcement Conference On March 2,1998, representatives of VY met with the NRC staff in the Region I office for a pre-decisional enforcement conference to review the apparent violations of the NRC requirements documented in NRC inspection report 50-271/97-10, dated February 5,1998. The conference was open to the public. A brief summary of the conference, list of attendees, and slides used by the NRC and VY staffs was docketed via report 50-271/98-03, dated March 3,1998. The NRC staff's

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enforcement action determinations will be docketed via a separate correspondence, j

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X2.3 Meetino to Discuss BMOs On March 5,1998, the VY staff met with members of the NRC staff in headquarters to discuss the current plans and activities regarding the Basis for Maintaining Operations (BMOs) that VY plans to have in effect coming out of their 1998 refuel outage. VY stated that none of the BMOs involve an unreviewed safety question and that all but 13 of the approximate 60 currently open BMOs

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were scheduled to be closed prior to restart from the refuel outage. The inspector notes that the 13 projected open BMOs was a reduction from the approximate 24 BMOs planned to be open as determined by the licensee in the internal l

memorandum of January 2,1998 (reference inspection report 50-271/97-12, section E2.1 and IFl 97-12-03). At the close of the March 5 meeting, VY agreed to additional meetings with the NRC staff to discuss their proposed ECCS suction l

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strainer modification and ECCS pump net positive suction head concerns related to the torus temperature BMO.

X3 Review of Updated Final Safety Analysis Report (UFSAR)

l A recent discovery of a licensee operating their facility in a manner contrary to the Updated Final Safety Analysis Report (UFSAR) description highlighted the need for a l

special focused review that compares plant practices, procedures and/or parameters to the UFSAR description. While performing the ir.spections discussed in this report, the inspectors reviewed the applicable portions of the UFSAR that related to the area inspected. The inspectors verified that the UFSAR wording was consistent with the observed practices and proceduros and/or parameters.

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INSPECTION PROCEDURES USED

37551 Onsite Engineering

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61726 Surveillance Observations 62707 Maintenance Observations 71707 Plant Operations 82701 Operational Status of the Emergency Pieparedness Program 86750-01 Solid Radioactive Waste Management and Transportation of Radioactive Materials 92903 Follow-up - Engineering 93702 Prompt Onsite Response to Events at Operating Power Reactors I

ITEMS OPENED, CLOSED, AND DISCUSSED OPEN NCV 98-01-01 Failure to perform 4-hour offgas system flow rate estimates, non-cited violation IFl 98-01-02 Potential degradation of RBCCW system containment isolation capability i

IFl 98-01-03 Use of unregistered 9249 cask prior to April 1,1996 and training for user's manual to be provided for radwaste computer program NCV 98-01-04 Failure to perform surveillance testing of APRM downscale trip functions for APRMs B and E, was not cited in accordance with Section Vll.B.1 of the NRC Enforcement Policy.

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CLOSED NCV 98-01-01 NCV 98-01-04 i

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DISCUSSED VIO 97-06-03 Failure to take effective corrective action for SBGT system potential over-pressurization IFl 97-12-03 BMOs proposed by VY to remain open following the 1998 refueling outage IFl 97-12-02 Potential condition outside design basis involving location of vacuum breakers for the HPCI and RCIC steam exhaust lines l

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PARTIAL LIST OF PERSONS CONTACTED i

i G. Maret, Plant Manager

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F. Helin, Tech. Services Superintendent l

E. Lindamood, Director of Engineering l

M. Balduzzi, Operations Superintendent K. Bronson, Operations Manager M. Watson, Maintenance Superintendent M. Desilets, Radiation Protection Manager R. Gerdus, Chemistry Manager G. Morgan, Security Manager S. Jefferson, Scheduling Manager-Operations J. Laughney, Quality Assurance Supervisor-Duke Engineering J. McCarthy, Radwaste Supervisor M. Pletcher, Technical Instructor-Training G. Weyman, Technical Assistant-Chemistry i

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24 LIST OF ACRONYMS USED BMO Basis for Maintaining Operation CFR Code of Federal Regulation CR control room CS cora spray.

-DAW Dry Radioactive Waste DOT Department of Transportation EDG emergency diesel generator ER Event Repon GL Generic Letter HIC High Integrity Container l

HPCI high pressure coolant injection IFl Inspector follow-up item IN information Notice

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l LCO Limiting Condition for Operation LER Licensee Event Report LPCI low pressure coolant injection LLRW Low Level Radioactive Waste LSA Low Specific Activity MCC motor control center j

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.NNS Non-nuclear safety l

PCP Process Control Program PORC Plant Operations Review Committee i

QA Quality Assurance RHR residual heat removal RCA Radiological Controlled Area RP&C Radiation Protection and Chemistry RW Radioactive Waste TS Technical Specifications i

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UFSAR-Updated Final Safety Analysis Report l

URI unresolved item

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VY Vermont Yankee

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