IR 05000482/1997023
| ML20199H950 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 01/29/1998 |
| From: | Johnson W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20199H882 | List: |
| References | |
| 50-482-97-23, NUDOCS 9802050159 | |
| Download: ML20199H950 (19) | |
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ENCLOSURE 2 U.S. NUCLEAR REGULATORY COMMISSION REGION IV .
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Docket No.: 50-482 License No.: NPF-42 Report No.: 50482/97-23 Licensee: Wolf Craek Nuclear Operating Corporation Facility: Wolf Creek Generating Station Location: 1550 Oxen Lane, NE Burlington, Kansas Dates: December 14,1907, through January 24,1998 Inspectors: J. F. Ringwald, Senior Resident inspector B. A. Smalldridge, Resident inspector K. M. Thomas, Project Manager, NRR -
Approved By: W. D. Johnson, Chief, Project Branch B i
ATTACHMENT: ' Supplemental Information M N 9902050159 900129 PDR ADOCK 05000482 G PDR~
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EXECUTIVE SUMMARY Wolf Creek Generating Station NRC Inspection Report 50-482/97-23 Ooerations
.- Safe and effectively controlled plant evolutions were supported by the consistent use of three-way communications. Effective turnovers were consistently noted (Section 01.1).
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In general, operators responded well to a containment atmosphere process radiation monitor failure that caused a containment purge and control room ventilation isolatio Operators in the control room did not make the expected plant announcement to inform the nuclear station operators and other plant personnel of the major ver:tilation cystem realignment (Section 01.2).
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The inspectors noted a weakness in the plant component labeling program in that the labels on the local breakers for the emergency diesel starting air compressors did not contain the name of the operated component (Section O2.1).
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When changing the charging pump lineup, several operating practicas did not meet management expectations, including an operator failure to make plant announcements, an operator failure to monitor pressurizer level long enough to assure adequate controller ,
performance after placing it in automatic, and the discovery of a general practice of -
reactor operators failing to wait for nuclear station operators to complete assigned procedure steps or get supervisory approval before continuing with subsequent procedure steps (Section 04.1)
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A plant safety review committee considered an unreviewed safety question detennination without the benefit cf pertinent reference information omitted by the issue presenter (Section 06.1).
Maintenance
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Instrumentation and control technicians' willingness to proceed with a surveillance test despite their uncertainty in lifting leads led to failures to comply with surveillance -
procedure instructions and caused the test equipment power source ground fault interrupter to trip (Section M1.4).
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The material condition of those plant systems and components evaluated during this
- inspection period were good, with few equipment deficiencies (Section M2.1).
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The licensee ended a longstanding practice of using a calibtation parameter specified by health physics supervision, rather than that specified by the calibration procedure, after the issue was questioned by the inspector (Section M3.1).
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2-Engineefing
Overall, the licensee's implementation of its 10 CFR 50.59 program was in accordance :
- with the program requirements. Safety evaluations were performed when required and contained sufficient information to support the conciusion that unreviewed safety i-questions did not exist.- However, procedural guidance existed regarding the conduct of
- tests and experiments that is not consistent with 10 CFR 50.59, and one minor weakness in updating the Updated Safety Analysis Report (USAR) was identified (Section E1.1)._
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Plant Sunoort
The licensee identified indications of a fuel leak. Chemistry personnel properly evaluated the available data and coordinated an appropriate response with operations
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(Section R1.2).
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Report Details i
Summary of Plant Status
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The plant operated at essentially 100 percent power throughout the inspection perio l. Operations 01 Conduct of Operations
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.0 Control Room Observations
. Insoection Scone (71707)
- The inspectors observed control room observations on a daily basis throughout the inspection perio b.' Observations and Findinas Throughout the inspection period, the inspectors observed consistent use of three-way
communications between operators. This enhanced operator performance of routine evolutions, including system operating procedures and surveillance test On December 17,1997, during a shift turnover, the offgoing operator provided a briefing
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of the plant status to the oncoming operator that conveyed an unusually detailed sense
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of recognition that every indication and control discussed was assochied with physical systems, structures, and components in the field.
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On January 15,1998, an operator reported that a controller had been placed in automatic. The supervising operator responded by repeating back the information while also correctly identifying the controller. This use of three-way communication also -
communicated supervisory expectations in an effective manner by giving the operator the supervisor's personal example of what was expecte On January 15,1998, operators also demonstrated the use of a practice the crew had adopted to " protect" an operator engaged in a critical task from distractions. All control room operators except the operator involved in the critica! task were directed to tespond to inquiries, telephone calls and Gaitronics communications that were unrelated to the critical task. At the same time, the supervising operator directed the operator performing the critical task not to engage in any activity that was not directly involved in the critical task. During subsequent conversations, the shift supervisor ttated that this was their response to NRC comments made associated with NRC Inspection Report 50-482/97-09, Section 0 . . . - . - . - . . - - . - - _ . - - _ - . _ - - - . . .
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s E Conclusions
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Safe and effectively controlled plant evolutions were effectively supported by the
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consistent use of three-way communications. . Effective turnovers were consistently *
, noted and one operator's turnover briefing was significantly better than mos .
. 01.2 ContainmerQuroe and Control Room Ventilation Isolation 7-
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n- a,' Insoection Scone (71707)
The inspector observed operators' response to a containment purge isolation and control'
, room ventilation isolation signal.
[ Observations and Findinas
On January 15,1998, operators received engineered safety feature signals for a
. containment purge isolation an4ontrol room ventilation isolation. Operators noted that -
h Containment Atmosphere Proom Radiation Monitor GT RE32 spiked above the Hi-Hi
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setpoint. The supervising operm h Nopriately directed operators to stop the -
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evolutions in progress when they couio do so safely and entered the appropriate alarm :
i- : response and off-normal procedures. The supervising operator demonstrated positive - *
control, and operators performed thorough walkdowns of the control boards to assess = .
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During subsequent troubleshooting, technicians found that the power supply had failed :
s and that the failure mode created spiking of the monitor output.- After replacing the power supply and completing all postmaintenance tests, the monitor returned to normal'
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[ operation and operators declared it operable.
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Neither the shift supervisor nor the supervising operator made plant announcements to 3 inform the nuclear station operators and other plant personnel that this major ventilation
- realignment had occurred. While this did not create any observable problems,~it had the - ,
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potential to creato confusion. . During subsequent discussions, the operations manager
- ~s tated that this did not meet management expectations and that the operations manager
l would address this with the shift supervisor, Conclusions :
' In general, operators responded well to a containment atmosphere process radiatio'n monitor failure that caused a containment purge and control room ventilation isolatio <
Operators in the control room did not make the expected plant announcement to inform nuclear station operators and other plant personnel of the major ventilation system realignment.
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O2.1. Plant Comoonent I aha!!na Proaram ' l
-. a.' Inapaction Scope (71707)
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t The inspectors observed plant component labels during routine plant walkdown.
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- Observahons and Findings
, On January 7,1998, the inspectors noted that the component labels for the emergency 7 r _
- diesel starting air compressor loca! breakers did not contain the noun name of the '
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- operated component or the main power supply. The component labels contained only - 7 7 the unique alpha-numeric identifier. The central work authority initiated an engraving
- . request to make and install the labels.
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The inspectors noted a weakness in the plant component labeling program in that the .
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labels on the local breakers for the emergency diesel starting air compressors did not -
contain the name of the operated componen Operator Knowledge and Performance
104.1. Operator Controller and Procedure Use
L Insoection Scone (61726)
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- The inspector observed a portion of a surveillance test on a centrifugal charging pum . ,
j : Observations and Findings
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- On January 15,1998,' operators performed a portion of Surveillance Procedure
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STS BG-_100A, " Centrifugal Charging System "A" Train Inservice Pump Test,"
Revision 19.- During the test the operator used Procedure SYS BG-201, " Shifting r
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Charging Pumps," Revision 31, to start the charging pump. The inspector noted that the 1 . operator performing the test made a plant announcement for starting the charg!ng pump, but failed to make the expected announcement for stopping the normal charging pum The inspector also noted that, after shifting the controller for Valve BG PCV-121, .
charging header flow control valve; to automatic, the operator only waited 20 seconds
- ! prior to signing off the procedure step to monitor pressurizer pressure level to ensure that E - the controller functioned properly to control level. -During subsequent discussions with the operations manager, the manager stated, that for this controller, if the v_alve position
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failed as is, it may take 4 or 5 minutes for pressurizer level to trend sufficiently for an operator to notice it. During discussions with the system engineer, the engineer said that
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- it would take minutes, rather than seconds, for one to detect a problem if the valve failed v
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as is, given the large volume of the reactor coolant system compared to the charging flow rates involved. However, every controller application can be different, ains operators tend to leam how quickly a given controller responds to change After starting the charging pump, the reactor operator directed the nuclear station operator to perform Steps 6.1.7 through 6.1.10 of Procedure SYS BG-201, The reactor operator then performed subsequent steps, and completed Steps 6.1.11 through 6.1.17, before receiving the notification that the nuclear station operator completed Steps 6, through 6.1.10. Administrative Procedure AP 15C-002, Section 6.2.3, requires
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procedure performers to complete or properly mark a step as not applicable before proceeding to the next step. When the inspector asked the shift supeivisor what the procedure use expectations were, the shift supervisor stated that the standard operator practice had been for operators to continue with procedure steps in the control room without waiting for nuclear station operators to complete the assigned steps, in response to the inspector's question regarding management expectations in this area, the operations manager determined that this understanding of procedure use was not isolated to one crew or a few operators and was not consistent with management expectations. The manager initiated management briefings for all operators to correct this understanding of procedure usage. The failure of operators to use Procedure SYS BG-201 in accordance with the requirements of AP 15C-002 is a violation of Technical Specification 6.8.1.a. (50-482/9723-01).
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During an operator surveillance test, several operating practices did not meet management expectations, including an operator failure to make plant announcements, an operator failure to monitor pressurizer level long enough to assure adequate controller performance after placing it in automatic, and the discovery of a general practice of reactor operators failing to wait for nuclear station operators to complete assigned procedure steps before continuing with subsequent procedure step Operations Organization and Administration 06.1 Plant Safety Review Committee Meetino Insoection Scooe (71707)
The inspector attended one plant safety review committee meet,ag, Observations and Findinas On December 23,1997, the inspector observed a plant safety review committee meeting that considered an unreviewed safety question determination which, if approved, would permit plant personnel to wear contact lenses with respiratory protection equipment. The committee did not approve the unreviewed safety question determination for several
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- p r reasons, incl 0 ding lack of preparation by the presenters, concems about licensee commitments, and lack of clarity in the wording of the unreviewed safety question
- determinatio After the meeting, the inspector asked the chairman and a presenter about a reference -
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document which suggested that hard contact lenses should not be used with respiratory . l
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protection equipment. The presenter did not provide this information to the committee,
and suggested that contact lens technology had improved to the point that prior concems
about wearing contact lenses while using respirators were no longer applicable, in -
- subsequent discussions with the chairman, the inspector expressed concem that the .;
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committee was vulnerable to presenters not providing pertinent information such as in this case. The committee discussed this concem during a subsequent meeting, and the
! chairman expressed the expectations that committee members review available documents and question presenters in a manner that minimized the potential for
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recurrence, i Conclusions
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A plant safety review committee considered an unreviewed safety question detemiination L without the' benefit of pertinent reference information omitted by the issue presente ,;
11. Maintenance M1 Conduct of Maintenance
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M1.1 General Comments on Maintenance Activdies
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. The inspectors observed all or portions of the following work activities:
103636, Task Baseline as-left motor-operated valve test on .
EM HV8293B p 126932, Task 1 EF LE27, Conduit repair
= RPP06-305, Task i Eberline PM-7 Calibration, restricted area exit portal monitor E'~
STN FP-211, Revision 6 Diesel fire pump monthly test WP 126671, Task 1 and 2 SGK04B, Removal and rigging
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' The inspectors found no concems with the maintenance observe ]
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The inspectors concluded that the maintenance activities we,re bsing performed as require j
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g M1.2 General Comments on Surveillance Activities
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- Inapaction Scope (82707)
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- The inspectors observed all or portions of the following surveillance activitie .;
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> STS BG-100A, Revision 19 Centrifugal Charging System A train inservice pump test STS 1C-4508, Revision.12 Channel calibration containment atmosphere and reactor ,
. coolant system leak detection radiation Monitor GT RE-31 ;
b i Observations and Findings
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- Except as'noted in Section M1.3, the inspectors found no concems with the surveillance
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- c, Conclusions b- Except as noted in Section M1.3, the inspectors concluded that the surveillance activities -
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6-M1.3 Process Radiation Monitor Surveillance
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The inspector observed a portion of the performance of Surveillance Test STS 1C-4508.
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. On January 7,1998, technicians performed a calibration check on Containment Atmosphere Radiation Monitor GT RE0031c As the technicians began setting up for the test, the inspector noted that a technician failed to provide the information specified in the
- caution block just prior to Step 8.4.1.2 to the control room operator. The technician i provided this information after being prompted by the inspector. A few minutes later,
[ Step 8.4.2.4 required the technician to remove the field wires from Terminal Block TB1- When the inspector asked the technician how the field side of the terminal block was i identified, the technician replied that it appeared to be the field side based on how the i, .
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wiring was routed in the cabinet, but if the test did not complete as expected, the technician would initiate a change to the procedure to clarify the proper lead to lift. Then the technician connected the leads from an autotransformer to the two terminals specified in Step 8.4.1.5. The technician reading the procedure properly identified the two terminals, but failed to identify which was to receive the hot lead and which was to receive the neutrallead. After the technician connected the leads, the technician reading the procedure re-read the step, including the details regarding where the hot and neutral lesds were to be connected. The inspector noted that the hot and neutral leads were on opposite terminals from what the procedure required and that the technician did not notice this error. The inspector asked what the color code was for hot and nautral wires and this prompted the technicians to detect and cor ect the miswirin The caution block just prior to Steps 8.4.1.3 through 8.4.1.10 warned the technicians to
"Use EXTREME caution while working inside the Motor Controller Assy. [ assembly), Use insulated tools,120 VAC may be present." The technician used a screwdriver with a taped shaft, but an uninsulated bit. While the technician was correcting the miswiring in Step 8.4.1.5, the inspector noted that the technician's finger contacted the uninsulated portion of the flat screwdriver bit with each revolution. The inspector questioned this practice while the technician performed the step. The response confirmed that the technician had not no' iced the touchin When the technicians energized the autotransformer, the ground fault interrupter on the extension cord tripped. This prompted the technicians to call a timeout to review the circuitry in the shop. The technicians determined that they had lifted the vendor lead from Terminal TB1-1, rather than the field wires as directed by the procedure. This led the technicians to initiate Performance improvement Request 98-0057 and prepare On-The-Spot Change 98-0030 to Surveillance Procedure STS 1C-4508. The failure of the technicians to comply with the surveillance procedu a is a violation of Technical Specification 6.8.1.a (50-482/9723-02).
c. Conclusions Instrumentation and control technicians' willingness to proceed with a surveillance test despite their uncertainty in lifting leads led to failures to comply with surveillance procedure instructions and caused the test equipment power source ground fault interrupter to tri .. . . -. - -- -. . -- . --
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-8-M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Review of Material Condition Durina Plant Tours Insoection Scone (61726) -
During this inspection period, routine plant tours were conducted to evaluate plant :
material conditio Observations and Findinas In general, where equipment deficiencies existed, the deficiencies had been identified for corrective action. The fo' lowing exceptions were identified;
. While observing the conduct of STN FP 221, ' Diesel Driven Fire Pump Monthly Test," Revision 6, the inspectors noted that oil and grease leakage on and around the diesel-driven fire pump engine had not been cleaned. in addition, much of the lighting in the diesel-driven fire pump room was inoperabl The inspectors noted that the licensee had initiated material condition improvements to the diesel-driven fire pump and pump room soon after the deficiencies were identifie ~. After a walkdown of the control room ventilation system, the inspectors noted in NRC Inspection Report 50-482/97-14 (Section E2.1) that the control room
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ventitation system had not received as much at:ention as other safety system The inspectors noted corrosion on the valves and tubing leading to Valve GK V0775, the access control fan coil unit chilled water supply line drain valve, and Valve GK V0776, the access control fan coil unit chilled water supply line vent valve. The inspectors also noted that the copper tubing supporting
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these valves vibrated noticeably while the unit was operatin During this inspection period, the inspectors noted that the material condition of Valves GK V0775 and -V0776 appeared unchanged from the condition identified 4 months previously in NRC Inspection Report 50-482/97-14. The licensee identified that the tubing and valves had been inspected and cleaned in September 1997, as a result of the original observation per Action Request 23883. The licensee generated Action Request 27177 in response to this observatio Conclusion The material condition of those plant systems and components evaluated during this inspection period were good with few equipment deficiencies.
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M3 Maintenance Procedures and Documentation l M3.1 Exit Tumstile Portal Radiation Monitor l Insoection Scooe (62707)
The inspector observed a portion of the calibration of the restricted area exit portal radiation monito Observations and Findinos On December 22,1997, a technician performed a calibration of the restricted area portal radiation monitor using Radiation Protection Procedure RPP 06-305, *Eberline PM 7 Calibration Restricted hea Exit Portal Monitor," Revision 1. During the procedure, the inspector end techniciar, discussed the technician's use of a reliably detectable activity value of 500, rather than 200 as specified in the procedure. The technician explained that the previous superintendent of radiation protection directed the tecnnicians to use the higher value in order to limit the detector count time. The technician also explained that the note prior to Step 9.2.14 permitted ihe technician to use values different from the procedure with the approval of health physics supervision. The technician also explained that this practice of setting reliably detectable activity to 500 had been in effect for the past 3 year The inspector discussed this issue with the manager of chemistry and radiation protection. After evaluating the issue, the manager said that this practice had been changed to require the technicians to perform the calibration using the reliably detectable activity value of 200 as specific by the procedure, Conclusions The licensee ended a long-standing practice of using a calibration parameter specified by health physics supervision, rather than that specified by the calibration procedure, after the issue was questioned by the inspecto M8 Miscellaneous Maintenance issues (92902)
M8.1 (Closed) Violation 50 482/9719-04: Failure to replace a shield plug per the maintenance procedure. The licensee's response letter referred to their violation response letter for NRC Inspection Report 50-482/97-20. The inspector verified the corrective actions describeo in the licensee's response letters for NRC Inspection Report 50-482/97-19, dated December 12 and 23,1997, to be reasonable and complete. No similar problems were identifie .- - - . . . . - . - - . - _ _ - - . - . - - - - - - - ~ . - - - . .
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.f E 1.1 ' 10 CFR 50 59 Imolementation (37001)
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a.- Inspection Scope
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! The inspector reviewed the licensee's program guidance and assessed the licensee's._
, - performance implementing its 10 CFR 50.59 safety valuation program during 1996 and L 1997." Specifically, the inspector reviewed a sample J 10 CFR 50.59 screenings (i.e. -
- applicability determinations) and associated unreviene safety question determinations-t and a sample of 10 CFR 50.59 screenings that did not iequire an unreviewed safety .
.o . question determination. In addition, a sample of USAR changes was reviewe .
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, Procedure AP 26A-003,? Screening and Evaluating Changes, Tests, and Experiments,"
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. Revision 3. ' The procedure delineated the licensee's methods, training requirements. , . ,
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- and responsibilities to determine and document whether facility changes can be made = j'
- without prior NRC approval. The following observations were made regarding the licensee's 10 CFR 50.59 program:-
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.- The "10 CFR 50.59 USQD Worksheet" is a valuable tool for performing safety =
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- evaluations. The addition of the requirement in Procedure AP 26A-003 in s September 1997 for individuals performing safety evaluations to use the worksheet should ensure that rigorous safety evaluations are performe :
4 The inspector identified that the definition for " Conduct of Tests and Experiments ,
not Described in the Updated Safety Analysis Report" contained in-6 Procedure AP 26A-003 is not consistent with 10 CFR 50.59. Specifically, the
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- definition states,-"The _ tests and experiments which are relevant to this phrase are -
those which are not described in the scope of the USAR and may: (1) degrade _ ,
. the margin (s) of safety during normal plant operations or during anticipated
. transients, or (2) adversely affect the adequacy of structures, systems, or .
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components designed to prevent accidents or to mitigate the consequences of an
. accident." The licensee informed the inspector that the definition was developed
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using guidance contained in NSAC-125, " Guidelines for 10 CFR 50.59 Safety _
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Evaluations." However, the NRC hs not endorsed this gbidance and
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10 CFR 50.59 does not provide fcr this screening of tests and experiments to determine if a safety evaluation is required. The inspector dH not identify any instances where a safety evaluation was not performed for a test or experiment.
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The inspector evaluated the implementation of the 10 CFR 50.59 program by reviewing a
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sample of the completed 10 CFR 50.59 screenings, a sample of safety evaluations
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-11-described in the Wolf Creek Generating Station Annual Safety Evaluation Report for 1996, and a sample of 10 CFR 50.59 safety evaluations performed in 1997. In addition, the inspector performed a review to verify that changes being made in the plant were being reflected in the USAR. The following observations were made:
10 CFR 50.59 Screenings The inspector reviewed the following 10 CFR 50.59 screenings to verify that safety evaluations were not required for the changes:
LS-05694 R/O; Change hioh and low level alarms for the secondary spent resin storage
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- Steam generator lower level tap root valve encapsulation LS-05999 R/O: Approval of alternate disc valve assembly and valve connector LS-06179 R/O: Change fuse size for heat tracing LS-06521 R/O: Elimination of lubrication for O-rings in diesel generator system CKL 2L-008: Change high/ low level limits for potable water tank level and basin
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WP 108639-15: Installation of temporary thermocouple probes into the inboard and outboard bearing housings for normal charging pump WP 121651: Temporary modification to disable alarms due to failure of an electric hydraulic controller power supply cooling fan flow switch WP 114490-64: Temporary relocation of excore nuclear instrumentation system audio count volume control WP 126420: Temporary modification to revise setpoint for Annunciator 388 (Loop 3 Delta T low deviation)
AP 28-006: Revision to Procedure AP 28-006, "Nonconformance Control" RNM-C-0574: Revision to Procedure RNM-C-0574, " Multi-Contact Aux Relay HFA" AP 27-006: Revision to Procedure AP 27-006, "WCGS Security Organization" LS-06604 R/O:. Installation of packing in a motor-driven auxiliary feedwater pump WP 113971-03: Temporary modification on startup transformer oil level switch WP 118801: Temporary modification to install recorder on battery buses
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-12-WP 107034 51: Installation of metal guard on inverter MPM C151Q-01: Revision to Procedure MPM C1510-01, " Containment Equipment Hatch Maintenance and Operation'
GEN 00-004: Revision to Procedure GEN 00 004, " Power Operation" SYS EJ-121, SYS EJ 120: Revision to Procedures SYS EJ-121,"Startup of A RHR Train in Cooldown Mode," and SYS EJ-120, "Startup of B RHR Train in Cooldown Mode'
STS AE-001: Revision to Procedure STS AE-001, " Main Feedwater isolation Valve Acc Discharge Test'
LS-05922 R/O: Replacement of radioactive waste pump impeller LS-07589 R/O: Removal of handwheel for essential service water traveling water screens WP 119878: Temporarf mcdification to install recorder for troubleshooting Based on the inspector's review of the changes and the USAR, the inspector concluded that the screenings were performed in accordance with the licensee's procedure (AP 26A-003) and 10 CFR 50.59 and that safety evaluations were not required for the change CFR 50.59 Safetv Evaluations The inspector reviewed the following safety evaluations for permanent modifiestions, temporary modifications, USAR changes, and defacto design changes to verify that the safety evaluations were adequate and prior NRC approval was not required for the changes:
96-0009: Temporary modification to install electric heaters to remove ice from essential service water travelir.g screens 96-0037: Temporary modification to provide cooling water to induction heater 96-0038: Procedure revision to allow the use of the safety injection system as an operable Mration flowpath in Mode 6 with the reactor vessel head removed 96-0041: Temporary modification to install pressure gauge in residual heat removal system 96-0060: Change in refueling machine overload and load reduction requirements 96-0072: Change to USAR regarding containment recirculation sump mesh size
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i 96-0073: Installation of tubing for safety injection pump mechanical seal drainage !
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96-00 4:J Changes to Reactor Coolant Pump No.1 seal leakoff local indication
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essential service water pump house piping and lake
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1 96-0139: Replacement of motor operators with handwheels from moisture se,parator/ reheater main steam supply bypass valves f
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96-0159: - Tomoorary modification to repair contact block for containment isolation signal u Phase B control switch ,
,96-0180: Correction to USAR regarding description of main steam isolation valves
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96-0202: Revision to USAR regarding pressure drop for essential service water strainers
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, 97-0001: Modification to change speed of refueling machine '
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I 97-0026: Revision to USAR to clarify description of emergency makeup requirements '
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97-0068: USAR revision to enhance the description for several valves;
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-97-0084! USAR revision to clarify auxiliary feedwater system description l
197-0108: Temporary test procedure for postmodification testing of transfer switches for ;
Jattery Charger NK25 -
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-97-0132:'- Revision of numbering _ scheme for several emergency plan procedures
~, 97-0141: Temporary test procedure to verify main _ steam isolation valve performance at-
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K The inspector concluded that, in general, the safety evaluations contained sufficient information to support the conclusion that an unreviewed safety question did not exist.-
- The inspector did not identify any changes that were made that required prior NRC
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- approval; Arnual 10 CFR 50.59 Reoort and USAR Unddag p
j' The inspector reviewed the licensee's Annua! Safety Evaluation Report, dated March 11,
- 1997, and determined that the safety evaluation summaries were of high quality and
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14-contained sufficient detail for NRC review. In addition, the inspector selected the following plant modifications that were summarized in the report to verify that the USAR was appropriately updated:
96-0038: Procedure revision to ailow the use of the safety injection system as an operable boration flowpath in Mode 6 wh the reactor vessel head removed 96-0060: Change in refueling machine overload and load reduction requirements 96 0086: Chenges to Reactor Coolant Pump 1 seal leakoff localindication 96-0117: Installation of resistance temperature detectors to monitor temperature in the essential service water pump house piping and lake 96 0139: Replacement of motor operators with handwheels on moisture separator / reheater main steam supply bypass valves The inspector concluded that, for Changes 96-0038,-0086, and 0117, the USAR had been properly updated to reflect the associated changes and for 90-0139 the USAR update was being processed. However, the inspector identified that for 96-0060 the USAR was not consistently updated. Specifically, the change was to allow one of the two automatic overload cutoffs to be automatically bypassed when a fuel assembly bottom nozzle is less than or equal to 2 inches above the full-down position in the core, upender, and rod cluster control assembly change fixture during raising of the fuel assembly. However, when pa0e 9.134 of the USAR was updated to reflect this change (Revision 10), the information adnd to page 9.134 of the USAR indicated that a specific overload will be bypassed insteau of one of the two overlo d cutoffs. The change was made correctly to pags 16.9-3 of the USAR. The licensee initiated a change notice to revise the USAR. The inspector concluded that the minor weakness in updating the USAR for 96 0060 was an isolated occurrenc c, Conclusion Overall, the licensee's imolernentation of its 10 CFR 50.59 program was in accordance with the program requirements. Safety evaluations were performed when required and contained sufficient inf',rmation to support 'he conclusion that unreviewed safety
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questions did not exist. However, procedural guidance existed regarding the conduct of testa and experiments that is not consistent with 10 CFR 50.59, and one minor weekness in updating the USAR was identifie .
15-IV. Plant Supped R1 Radiological Protection and Chemistry Controls R Potential Unlocked Hiah Radiation Area BouDdAIY Insoection Scone (71750)
The inspector reviewed the results of the licensee's investigation of a reported unlocked high radiation area boundary, Observations and Findinas On December 14,1997, a health phys;cs technician reported that the radwaste filter bay crane, a locked high radiation area boundary, was unlocked, when the technician did not recall unlocking it. The health physics technician, two operators, and two maintenance technicians went to the filter bay to replace the filter element in one of the filter housing During the setup, the health physics technician heard the crane movement and questioned whether the crane had been unlocked. The technician subsequently initiated Performance Irnprovement Request 97-4069, which prompted a Severity Level ll investigatio The inspector reviewed the results of the licensee's investigation and determined that it was thorough. The inspector agreed with the conclusion that the only plausible explanation for th. Issue was that the health physics technician who questioned the unlocked crane was preoccupied and did not recall unlocking the crane. The sound of the crane movement captured the technician's attention and this prompted the questio The licensee's investigation determined that other matters contributed to the technician's likety preoccupatio Conclusions The licensee's investigation of a reported unlocked high radiation area boundary was appropriate and thorough. The licensee's conclusion that the boundary was unlocked per their administrative procedures appeared to be appropriat R1.2 Fuel Leak Insoection Scoce (71750)
The inspector reviewed the data the licensee used to determine that they had a small fuelleak, and reviewed their respons _ _ _ _ _ - _ _ _ _ _ _ _
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, Observations and Findings On December 15,1997, chemistry personnel determined that the fuel reliability index exceeded 5E 4 micro-curies per grain. Seven days later the licensee followed their.
[ administrative procedure by entering Action Level 1 for failed fuel and increased the
! _ charging and letdown flow rate to 120 gallons per minute. Since detecting the leak the _ ,
reactor coolant gross activity remained essentially constant the fuel rollability index has j remained essentially constant just above the Action Level 1 point, and the Xenon-133 to _ {
Xenon +135 activity ratio has slowly risen from 0 to 3.4. These indications were
.l consistent with what would be expected from a single, tight fuel lea l Chemistry personnel consuNed with their fuel vendor, and the fuel vendor believes that the leak is from a previously bumed assembly. However, this conclusion is not i
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supported by available data since the gross activity and cesium activity levels are too low
. to permit an accurate analysis. The licensee will continue to monitor this lea Conclusions l The licensee identified indicatione of a fuelleak, Chemistry personnel properly evaluated 4he available data and coordinated an appropriate response with operation V. blanagement bieetings
- X1 Exit R4eeting Summary The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on January 23,1998. The licensee' acknowledged the findings
. presente ' The inspectors asked the licensee whether any materials examined during the inspection should D
be considered poprietary. No proprietary information was identifie ; ;= ~-
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ATTACHMENT PARTIAL LIST OF PERSONS CONTACTED Licensee M. J. Angus, Manager, Licensing and Corrective Action G. D. Boyer, Chief Administrative Officer J. W. Johnson, Manager, Resource Protection O. L. Maynard, President and Chief Executive Officer B. T. McKinney, Plant Manager R. Muench, Vice President Engineering W. B. Norton, Manager, Performance Improvement and Assessment C. C. Warren, Chief Operating Officer INSPECTION PROCEDURES USED IP 37001 10 CFR 50.59 Implementation l IP 37551 Onsite Engineering l
l lP 61726 Surveillance Observations IP 62707 Msintenance Observations IP 71707 Plant Operations IP 71750 Plant Support Activities IP 92902 Followup Maintenance ITEMS CPENED. CLOSED. AND DISCUSSED Opened j 50-482/9/23 01 VIO Procedure Steps Performed Out of Sequence (Section 04.1).
504 82/9723-02 VIO Process Radiation Monitor Surveillance (Section M1.3).
GQ1ed 50-482/9719-04 VIO Unposted locked HI rad area due to incomplete maintenance (Section M8.1).
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