IR 05000397/1996008
| ML17292A375 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 07/23/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML17292A374 | List: |
| References | |
| 50-397-96-08, 50-397-96-8, NUDOCS 9607250203 | |
| Download: ML17292A375 (34) | |
Text
ENCLOSURE U.S.
NUCLEAR REGULATORY COMMISSION
REGION IV
Docket No:
50-397 License No:
NPF-21 Report No:
50-397/96-08 Licensee:
Washington Public Power Supply System Facility:
Washington Nuclear Project-2 Location:
3000 George Washington Way P.O.
Box 968, MD 1023 Richland, Washington Dates:
May 12-June 22, 1996 Inspectors:
R.
C. Barr, Senior Resident Inspector G.
D. Replogle, Resident Inspector J.
W. Clifford, Project Manager J.
L. Dixon-Herrity, Resident Inspector (Wolf Creek)
T.
R.
Meadows, License Examiner V. L. Beaston, Electrical Engineer, NRR Approved by:
H. J.
Wong, Chief, Branch E
Division of Reactor Projects 9607250203 960723 PDR ADOCK 05000397
EXECUTIVE SUMMARY Washington Nuclear Project-2 NRC Inspection Report 50-397/96-08 This routine announced inspection included aspects of licensee operations, engineering, maintenance, and plant support.
The report covers a 6-week period of resident inspection.
~Oerations
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The control room staff generally demonstrated good three-way communications; however, on occasions the inspectors observed lapses in the formality of their communications (Section 01.2).
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The licensee refueled the reactor safely.
For'he second consecutive year no fuel mispositioning errors occurred, which was particularly noteworthy since this year's refueling involved a full core offload (Section 01.3).
Shift supervision effectively controlled the reactor startup from Refueling Outage Rll (Section 01.4).
Because intermittent high electronic noise (spiking)
on three of the four source range monitors (SRM) was not corrected prior to reactor startup, false high period alarms complicated the operators'pproach to criticality (Section 01.4).
Maintenance Generally, maintenance and surveillance activities were performed well (Section Ml).
The performance of the reactor pressure vessel cold hydrostatic leakage test was improved when compared to past performance (Section M1.3).
The licensee found that in-service testing of some excess flow check valves was not performed during the last refueling outage (R10)
and two-reactor core isolation cooling valves were not properly tested prior to the startup from this refueling outage (Rll) (Section M1.5).
Licensee control of overtime during outages warrants strengthening.
During the last weeks of the outage, deviations from overtime guidelines were authorized for all instrumentation and control craftsmen.
The use of overtime during this period was caused by weaknesses in outage planning practices (Section M1.6).
En ineerin The licensee had moved fuel assemblies to the spent fuel pool (SFP)
within the bounds of the design bases; however, the NRC reviewer identified procedure weaknesses and inconsistencies associated with the movement of irradiated fuel (Section El.l).
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Electrical calculations and surveillance procedures in support of the, licensee's conversion to improved standard technical specifications were thorough and well documented.
The licensee electrical engineers were very knowledgeable about the calculations and the electrical system (Section El.2).
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The licensee failed to take adequate corrective actions in response to an NRC Information Notice related to motor pinion key failures.
This contributed to the failure of the motor pinion key of a low pressure core spray injection motor operator, resulting'in the valve being in a degraded condition.
However, the subsequent evaluation and corrective actions were considered to be sound (Section E8. I).
Plant Su ort The inspectors found that the licensee propped opened fire doors to facilitate area ventilation and did not consistently restore these doors in a timely manner when having the door open was no longer required (Section Fl.l).
Cy Re ort Details Summar of Plant Status On March 3, 1996, the plant was shut down for economic dispatch due to excess electrical generation in the Northwest.
On April 15, the plant began Refueling Outage R11.
The plant was in Mode 4 or Mode 5 for Refueling Outage Rll until June 12, 1996.
Operators then began plant startup (Mode 2).
On June 19, the plant entered Mode 1 and operators began power ascension.
On June 21, operators synchronized the main generator to the electrical distribution system.
The inspection period ended with the plant at approximately 23 percent power.
I.
0 erations Ol Conduct of Operations 01. 1 General Comments 71707 Using Inspection Procedure 71707, the inspectors conducted frequent reviews of ongoing plant operations.
The inspectors observed the conduct of shutdown, prestartup, and startup activities.
The inspectors assessed communications and coordination of plant activities.
In general, the conduct of operations was professional and safety-conscious; specific events and noteworthy observations are detailed in the sections below.
01.2 Reactor Shutdown and Prestartu Observations a.
Ins ection Sco e
71707 71711 The inspectors observed communications and coordination of plant activities, verified selected safety system lineups, and assessed the licensee's use of overtime during the outage.
b. Observations and Findin s
Communications:
Generally, the control room staff demonstrated good three-way communications during shutdown activities.
However, on occasion, the inspectors observed lapses in formality where an operator or supervisor would casually nod or hand-wave to acknowledge a
communication.
At times, the control room staff appeared unaware of ongoing control room outage work and unaware of contractor personnel working in the back panel area.
Equipment Lineups:
Yalve lineups for the Residual Heat Removal System (RHR)
B, Containment Atmosphere Control System A and Diesel Generator 2 support systems were independently verified using licensee procedures and drawings.
The inspectors found all components correctly positioned,
C.
Overtime Usage:
The inspector reviewed selected work time records of operators.
Records were assessed with respect to Plant Procedure Hanual (PPH) 1.3.27,
"Excessive Hours Worked Control," Revision 17, and Technical Specification (TS) 6.2.2.
The inspector found that, in general, the licensee documented deviations from exceeding overtime limits and the plant manager authorized, these deviations.
Of the 10 operations staff records reviewed, the inspector identified one instance where a senior reactor operator (SRO)
had exceeded the licensee's over time limits without obtaining the appropriate authorization.
From Hay 28 to June 7,
1996, the SRO worked in excess of the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (actual maximum, 75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br />)
in any 7-day period.
Although the SRO was not working on safety-related work during the subject period (the TS requirements do not apply),
PPH 2.3.27 specified that the SRO obtain verbal authorization from his,immediate superior before exceeding the limits.
This was not accomplished.
No other problems with operations personnel'were observed.
Conclusions The control room staff generally demonstrated good three-way communications; however, on occasions the inspectors observed lapses in the formality of their communications.
The licensee performed safety system valve and switch lineups effectively.
The inspectors identified one instance where a licensed SRO while performing nonsafety-related activities exceeded licensee overtime limits without appropriate authorization.
01.3 Refuelin 0 erations a.
Ins ection Sco e
60710 b.
The inspector observed portions of reactor refueling on Hay 20-22, 1996.
Observations and Findin s
The inspectors observed that the operators and contractors followed PPH 6.3.2,
"Fuel Shuffling and/or Offloading and Reloading,"
Revision 9.
This included verification by the refueling shift supervisor that the correct fuel assemblies were placed in their assigned locations.
The inspector independently checked a sample of the fuel movements and verified them to be correct.
For the second consecutive year no fuel mispositioning errors occurred, which was particularly noteworthy since the full core was offloaded this refueling outage.
The refueling supervisors and contractors exhibited a good questioning attitude and took proper action when problems were encountered.
At one point, the supervisor corrected poor communication practices of refueling
personnel on the bridge.
On another occasion, refueling operations were suspended to correct a problem with the camera used for the verification of fuel assembly serial numbers.
The inspectors noted that refueling personnel used good radiation protection and foreign material exclusion practices.
c. Conclusions The licensee refueled the reactor safely.
For the second consecutive year no fuel mispositioning errors occurred, which was particularly noteworthy since the full core was offloaded this refueling outage.
4.4
~R>>
a.
Ins ection Sco e
71711 From June 12-22, 1996, the inspectors observed portions of the reactor startup and power ascension.
b, Observations and Findin s
Regarding the startup to criticality, operations demonstrated good command and control, appropriate procedure adherence practices, and effective communications.
The control room supervisor actively participated in the startup and provided good oversight to the crew.
During reactor startup, the inspectors observed that intermittent high electronic noise signals (spiking)
on three of the four SRHs challenged operators by frequently causing false high period alarms.
The problem had the greatest impact on the operators when the reactor approached criticality.
The SRH spiking began approximately 4 weeks prior to reactor startup.
The licensee conducted troubleshooting, but could not identify the source of the electrical noise and, therefore, could not correct the problem.
The licensee planned to perform additional troubleshooting during the upcoming economic dispatch period.
c. Conclusions Shift supervision effecti,vely controlled the reactor startup from Refueling Outage Rll.
Because spiking of the SRHs was not corrected prior to reactor startup, false high period alarms complicated the approach to criticalit Operati onal Status of Faci 1 ities and Equi pment 02.1 En ineered Safet Feature S stem Walkdowns 71707
The inspectors used Inspection Procedure 71707 to walk down accessible portions of the following engineered safety feature systems:
reactor core isolation cooling (RCIC)
RHR Trains A, B and C
high pressure core spray low pressure core spray (LPCS)
Equipment operability, material condition, and housekeeping were acceptable in all cases.
Several minor discrepancies were brought to the licensee's attention and were corrected.
The inspectors identified no substantive concerns as a result of these walkdowns.
II. Maintenance Hl Conduct of Maintenance Hl. 1 General Comments a.
Ins ection Sco e
61720 61726 62703 The inspectors observed all or portions of the following surveillance activities and work order tasks:
PPH 7.0.0:
PPH 7.0.2:
PPH 7.4.4.6.1.1A:
PPH 7.4.0.5.25:
Shift & Daily Checks in Hodes 1, 2,
& 3 Shift & Daily Instrument Checks, Hode
Reactor Pressure Vessel Heatup Surveillance Reactor Pressure Vessel Cold Hydro Leakage Test PPH 7.4.6. 1.2.7:
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WOT ZD12-04:
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WOT YHCO-01:
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WOT WT5001:
Hain Steam Isolation Valve (HSIV) Leak Rate Testing (documentation review only)
RCIC Governor Valve Linkage Realignment Freeze Seal Installation CRD-V-101/5027 Repair
Observations and Findin s
The work performed under these activities was professional and thorough.
The inspectors observed that procedures were followed and activities were stopped when problems were encountered.
The following describes the inspector's observations.
H draulic Control Unit Valve Re air a
~ Ins ection Sco e
62703 b.
The inspector assessed the work instructions, quality of work and procedure adherence associated with the repair of control rod drive hydraulic control Valve CRD-V-101/5027.
Craftsmen performed the work using Work Orders YHCOOI and WT5001.
Observations and Findin s
On May 22, 1996, the inspector observed work associated with establishing a freeze seal on the insertion line of Valve CRD-V-101/5027.
Because the task had the potential to drain the reactor cavity, the licensee performed a
CFR 50.59 evaluation for the work.
The inspector reviewed the safety evaluation and found that the analysis appropriately addressed the concern.
The inspector considered procedure quality acceptable, observed that the mechanics followed the work instructions and used good radiation protection practices.
The craftsmen had a good questioning attitude during the installation of the freeze seal and stopped when they could not achieve the procedure-required temperature to verify the adequacy of the freeze seal.
Engineers determined that. the temperature sensor had been located too far from the area of the freeze seal.
The mechanics moved the sensor and the required temperature was achieved.
The inspector found that PPM 10.2.72,
"Freeze Seals Using Liquid Nitrogen as the Freezing Agent," Revision 3, did not precisely identify the location for taking the temperature measurement.
Improved guidance appeared warranted.
The inspector also noted that the craftsmen used appropriate foreign material exclusion controls.
However, the work package referenced the wrong foreign material exclusion control procedure (PPM 1.3.38),
instead of the procedure for foreign material control around open systems.
This indicated a need for additional attention to detail by the work planne H1.3 PPH 7.4.0.5.25 Reactor Pressure Vessel Cold H drostatic Leaka e Test a
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Ins ection Sco e
61726 b.
The inspector assessed the quality of the surveillance procedure, observed operator controls during reactor coolant system pressurization, and assessed the reactor coolant system for leakage.
Observations and Findin s
The inspector found that the licensee had improved the quality of the procedure since it was last performed.
The procedure had been revised to improve the operators'bility to control pressure by installing video cameras that provided direct observation of the high accuracy pressure gauges.
This eliminated a communication link so that the operator controlling reactor pressure and level could immediately assess the impact on level and pressure of valve manipulations.
The inspector observed that operators adhered to the procedure.
The inspector observed excellent command and control during the test.
The inspector observed that the licensee's walkdown of the reactor coolant system was thorough.
The licensee identified a small crack in the socket weld of a 3/4-inch leak off line of a reactor recirculation isolation Valve RRC-V-67A and a number of flange, body-to-bonnet, and fitting leaks.
The leaks were corrected prior to reactor startup.
H1.4 HSIV Local Leak Rate Testin LLRT a
~ Ins ection Sco e
61720 b.
The inspectors reviewed pertinent testing procedures to evaluate the effectiveness of HSIV LLRT and held discussions with test personnel to understand the performance of the LLRT.
Observations and Findin s
The inspectors found that the licensee was wetting the HSIV internals before stroking the valves for LLRT as required by PPM 7.4,6. 1.2.7,
"HSIV Leak Rate Testing," dated April 26, 1994.
Wetting of the HSIV internals concerned the inspectors"'ecause the practice could honconservatively affect the test results.
The licensee asserted that the practice was recommended by the valve manufacturer and General Electric and was intended to preclude damaging the valves.
The licensee also believed that the test results were not affected by wetting the internals.
The licensee agreed to provide technical justification and other documentatio'n to support their position.
This issue is an unresolved item pending further NRC review of the licensee's justification (URI 50-397/9608-01).
-10-Hl. 5 Failure to Perform Selected In-Service Testin IST a.
Ins ection Sco e
61726 The inspectors reviewed licensee-identified problems associated with failing to perform IST on some safety-related valves.
b. Observations and Findin s
On June 13, the licensee identified that stroke time testing and valve position indication verification were not performed for RCIC Valves RCIC-V-63 (inboard steam supply isolation valve)
and RCIC-V-76 (warmup line isolation valve) prior to declaring the valves operable and prior to startup, Both valves are containment isolation valves.
TS 4.0.5 requires that the valves be tested in accordance with Section XI of the ASME Code, which in turn requires the valve testing.
Per PPH 3. 1. 1,
"Master Startup Checklist," operations personnel should have verified that all required testing had been completed prior to startup, In response to the finding, the licensee identified that motor-operated valve diagnostic testing was performed on the valves during the outage and concluded that the testing could be credited as meeting the requirements of the missed IST surveillances.
Additionally, per problem evaluation request (PER)
296-0426, dated Hay 24, 1996, the licensee also identified that remote valve position indication and full stroke exercises for approximately 85 excess flow check valves were not correctly scheduled and, therefore, not performed during Refueling Outage R10.
The tests were required to be accomplished each refueling outage per the'IST Program and TS 4.0.5.
Subsequently, the licensee reported that the required testing was performed during the current outage (Rll) and no operational problems were identified.
The failure to perform the required testing on the noted RCIC valves and excess flow check valves is an unresolved item pending NRC review of the licensee's evaluation and corrective actions (URI 50-397/9608-02).
H1.6 Overtime Usa e
a.
Ins ection Sco e
62703 The inspector reviewed selected work time records.
Records were assessed with respect to PPH 1.3.27,
"Excessive Hours Worked Control,"
Revision 17, and TS 6.2.2.
b. Observations and Findin s
The inspector reviewed approximately 40 maintenance personnel work time records and found that the plant manager had authorized the entire instrumentation and controls staff to exceed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any 7-day period for the period from May 26 to June 14, 1996.
During this time, many of
-11-M1,7 the instrumentation and controls craftsmen worked each day of this period between 10 to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, i.e., greater than 14 continuous days.
This appeared to'be excessive usage of overtime.
The plant manager acknowledged that overtime had been used more than he would have
'preferred during the later portion of the outage.
Through further investigation, the inspectors found that the excess overtime usage was caused by poor outage planning practices.
Planners had not appropriately estimated some jobs and insufficient manning was provided to account for emergent work.
Tem orar Power Cables a
0 Ins ection Sco e
71707 b.
The inspector noted a number of temporary nonsafety-related power cables in contact with safety-related conduit in the reactor building and reviewed the possible effects of the lack of separation between the cables and discussed concerns with engineering personnel.
Observations'and Findin s
The inspector identified four instances where temporary nonsafety power cables contacted safety equipment or crossed divisional boundaries.
The inspector was concerned because these temporary power cables could result in connecting two divisions of safety-related power in the event of a significant electrical fault.
The licensee is not committed to NRC Regulatory Guide 1.75,
"Physical Independence of Electrical Systems,"
which prohibits this type of temporary cable routing.
PPH 1.3.9,
"Temporary Modifications," Attachment 7.4 listed six acceptable routing combinations for structure and raceway that could be used to support temporary cable.
These combinations did not list contact between Divisions
and 2 as an acceptable configuration.
Additionally, the attachment stated that temporary cables should not be laid on safety-related equipment.
The inspector noted that the attachment provided guidance for the routing of temporary electrical cables and not requirements.
Further, the attachment was located in the temporary modification procedure which was not commonly referred to when installing temporary cables.
The inspectors discussed the concerns with electrical engineering personnel.
The engineers>> assessed each example, explained that the routing of these temporary cables did not meet management expectation or separation criteria, initiated a Gold Card to document the discrepancies, and had the cables rerouted or removed.
Additionally, the licensee walked down the reactor building to identify and correct other cable separation issue NRC inspectors in Inspection Report 50-397/94-29 expressed the same concern over the routing of temporary power cables.
In response to this concern the licensee wrote a memorandum,
"Electrical Separation Requirements,"
dated November 3, 1994.
The memorandum stated that PPH 1.3.9,
"Temporary Modifications," Attachment 7.4,
"Separation Criteria for Temporary Cables,"
should be used for the positioning of temporary cables.
However, as noted previously, the guidelines of PPH 1.3.7 were not specified as requirements.
The licensee had also identified an electrical separation issue associated with the adjustable speed drive and digital feedwater (ASD/DF)
~
modifications (PER 296-0411).
Based on these examples, the inspector considered the licensee's attention to electrical separation required strengthening.
Because of these inspector-identified issues, the licensee plans to review and revise procedures to prevent further electrical separation issues created by temporary power cables.
Conclusions on Conduct of Maintenance M8.1 Haintenance and surveillances were generally performed well.
Two unresolved items were opened related to the LLRT of HSIVs and the failure to perform IST on safety-related valves.
The performance of the reactor pressure vessel cold hydrostatic leakage test was improved when compared to past performance.
Excessive usage of overtime was required because of poor outage planning practices.
Actions to improve the controls for the routing of temporary power cables have not been entirely effective in implementing management expectations.
Miscellaneous Maintenance Issues (92902)
Closed Violation 50-397 9419-01:
lights dropped into the reactor vessel.
In June 1994, while repositioning the service platform, a,light dropped into the reactor vessel when workers severed a light cord while rotating the service platform.
Due to an inadequate retrieval plan, nonexistent contingency actions, and poor shift briefings and turnovers, a second light was lost when the service platform was moved a second time.
Prior to retrieving the lights, one light went into RHR piping.
In response to the event and subsequent NRC violation, the licensee:
(1) established requirements to immediately inform appropriate levels of management when items are dropped into the vessel; (2) implemented requirements for contingency and retrieval plans for dropped items;
-13-(3) implemented additional controls, when moving the service platform, to ensure that lights and other items are not inadvertently severed or dropped into the reactor vessel; and (4) provided periodic training to pertinent operations and maintenance personnel regarding the event.
The inspectors observed activities during the most recent refueling and noted that controls and procedures were improved when compared to previous years.
The corrective measures appeared to be effective.
III. En ineerin Conduct of Engineering SFP Desi n and 0 eration Review Audit Sco e
92703 The inspector reviewed the FSAR, TSs, and plant procedures related to operating practices, design basis, and licensing basis for the SFP.
The inspector also reviewed information on current practices, logs of past refueling outages, and a draft analysis for Refueling Outage R12.
Observations and Findin s
Based on a limited review of past refuelings, the reviewer determined that the licensee has apparently moved fuel to the SFP within the bounds of the design bases.
The reviewer confirmed the licensee's verification that for the most recent full core offload (1992) fuel was not moved until 8 days after plant shutdown which met the 5-day design assumption.
The reviewer identified two procedure weaknesses associated with the operation and control of the SFP.
Licensee procedures do not include administrative controls which would preclude movement of irradiated fuel to the SFP prior to less than 5 days after shutdown (the time assumed i'
the design analysis).
Licensee procedures also do not include actions to take if RHR Pump A fails while in shutdown cooling and the SFP is in RHR/fuel pool cooling (FPC) assist mode using RHR Pump B,
The design analysis assumes that, for "nominal" (refueling) core offloads, fuel is not moved prior to 20 days after plant shutdown.
This portion of the analysis assumes SFP cooling is provided only by the SFP cooling system, assuming a single failure.
The design analysis also assumes that for (RHR)/(FPC) assist mode of operation, fuel is not moved prior to 5 days after plant shutdown.
This portion of the analysis assumes a design heat load based on 12 cycles of nominal core offloads with a full core offload at the 12th refueling outage.
These analyses are documented in the original 1976 plant design calculation and the FSAR
C.
contains tabular data based on these analyses.
The licensee considered the design basis allowed movement of irradiated fuel to the SFP 5 days after shutdown, if the RHR/FPC assist mode is available.
However, these design assumptions are not in the FSAR, the TS, or plant procedures.
Lastly, the inspector identified that the design basis assumptions allows SFP temperature to increase to 150'F when circulating water is interrupted for draining the reactor cavity and dryer-separator pit.
However, current operating procedures restrict SFP temperature to 125'F.
Licensee procedures and TS do not contain specific consideration for allowing SFP temperature to increase to 150'F.
The licensee acknowledged these procedure weakne'sses.
Followup review by the resident inspector has verified that appropriate procedural changes have been made to assure that procedures reflect the SFP design bases.
Conclusions The licensee has moved irradiated fuel to the SFP within the bounds of the design bases.
The inspector identified procedure weaknesses and inconsistencies associated with the operation and control of the SFP which the licensee has collected.
E1.2 Review of Electrical Calculations
'a ~ Ins ection Sco e
37551 b.
Electrical reviewers from the Office of Nuclear Reactor Regulation conducted onsite inspection of electrical calculations and surveillance procedures in support of the licensee's conversion to improved standard technical specifications.
The reviews included calculations for new allowable values for the loss-of-power instrumentation and the reactor protection system electrical protection assemblies.
Observations and Findin s
The electrical reviewers inspected the following calculations and surveillance procedures along with other related design documents such as drawings, TS, and the FSAR:
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E/I-02-87-07 WNP-2 Plant Hain Bus Voltage Calculations for Normal and Loss-of-Coolant Accident Operations
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E/I-02-90-01 Low Voltage System Loading and Voltage Calculations
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E/I-02-93-08 Input Voltage Variation Determination for Reactor Protection System (RPS) Electrical Protection Assembly Equipment
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2.12.18
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2.12.24
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2.12.58 Loss-of-Voltage Relay Settings for Class lE, 4. 16KV, Division
5 Division 2 Buses SM7 E
SH8 Loss-of-Voltage Relay Settings for the Class 1E, 4. 16KV, Division 3 Bus SM4 Degraded Voltage Relay Settings for the Class 1E, 4. 16KV, Division 1, Division 2 5 Division, 3 Buses SM7, SM8 L SM4
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E/I-02-93-1282 Undervoltage Relay Setting Calculations for RPS Electrical Protection Assembly
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E/I-02-93-1283 Underfrequency Relay Setting Calculations for RPS
'lectrical Protection Assembly
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E/I-02-93-1284 Overvoltage Relay Setting Calculations for RPS Electrical Protection Assembly
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E/I-02-95-01 Overcurrent Protective Device Settings and Coordination Calculations for 480 Volt Distribution Systems
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EES-4 Setpoint Methodology
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PPH 7.4.3.3. 1.67 Channel Functional Test for 4. 16KV Emergency Bus Degraded Undervoltage (SH7)
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PPH 7.4.3.3. 1.70 Channel Calibration Procedures for 4. 16KV Emergency Bus Primary Undervoltage
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PPH 7.4.3.3. 1.74 Channel Calibration Procedures for 4. 16KV Emergency Bus Degraded Undervoltage The reviewers noted that the bases for the calculations and assumptions were well documented and thorough and found the licensee engineers knowledgeable about the calculations and the electrical system.
The l'icensee's setpoint methodology was very detailed and thorough.
The licensee had a well coordinated degraded voltage protection scheme that considered the effects of degraded voltages on all buses.
No deficiencies with the calculations were identified.
c. Conclusions Electrical calculations and surveillance procedures in support of the licensee's conversion to improved standard technical specifications were thorough and well documented.
The licensee electrical engineers were very knowledgeable about the calculations and the electrical syste E2.1-16-Engineering Support of Facilities and Equipment Fuel Debris Filters a.
Ins ection Sco e
37551 n
The inspector reviewed and discussed PER 296-0437, which documented the potential for fretting in ASEA Brown Bovari (ABB) SVEA-96 fuel, with the licensee.
b. Observations and Findin s, C.
The licensee will use SVEA-96 fuel for the first time following Refueling Outage Rll.
The fuel assemblies have debris filters at the bottom of each fuel channel which consist of inconel springs which are secured in place with stainless steel pins around the spring.
During testing, the vendor observed fretting where the pins and springs contact.
In a worst-case scenario, the fretting could continue until spring failure occurred.
Small spring parts could then be transported into the fuel channels and initiate fuel cladding failures.
The vendor notified the licensee of this issue and the licensee initiated a
PER to assess its safety impact.
The licensee consulted ABB and determined that the potential for debris filter fretting was not an immediate safety concern.
The licensee noted that the flow in the test loop was approximately 130 percent of the flow experienced at WNP-2.
The licensee learned that several European plants have used SVEA-96 fuel, operating at flow rates similar to WNP-2, and without any spring failures.
The licensee plans to monitor the results of inspections being planned through the summer in the European facilities and to examine the debris filters at the conclusion of the current operating cycle.
Conclusions E2.2 The licensee evaluation and planned actions appeared acceptable.
Review of Facilit and E ui ment Conformance to FSAR Descri tion A recent discovery of a licensee operating their facility in a manner contrary to the FSAR description highlighted the need for a special focused review that compares plant practices, procedures, and/or parameters to the FSAR description.
While performing the inspection discussed in this report, the inspectors reviewed the applicable portions of the FSAR that related to the areas inspected.
No problems were identifie H-17-E8.1 Miscellaneous Engineering Issues (92903)
Closed Unresolved Item 50-397 9606-03:
LPCS-MO-5 motor pinion key failure.'n April 21, during a required inspection of the LPCS injection valve motor operator (LPCS-M0-5), the licensee identified that the motor pinion key was sheared axially (described in PER 296-0287).
The motor pinion key transfers torque between the motor shaft and the pinion gear set, which in turns drives the actuator to move the valve.
The licensee determined that the val.ve remained operable since it had stroked under static conditions (no differential pressure)
on two occasions just'prior to discovery of the sheared key.
However, since operational requirements under static and dynamic conditions are not identical, the inspectors concluded that valve operability could not be guaranteed.
Through subsequent investigation, the licensee identified that the corrective actions in response to NRC Information Notice (IN) 94-10,
"Failure of Motor-Operated Valve Electric Power Train Due to Sheared or Dislodged Motor Pinion Gear Key," dated February 4, 1994, were inadequate.
The IN, in part, addressed a motor pinion key failure at another facility, where the motor-operator was similar to LPCS-MO-5.
The corrective action for this type of problem was to replace the original key made from 1018 steel with a key made from 4140 steel.
In this case, the engineer mistakenly assumed that all SB-3 motor pinion keys had been replaced in response to IN 81-08, "Repetitive Failures of Limitorque Operator SMB-4 Motor-To-Shaft Key," dated March 20, 1981.
Contrary to what the engineer believed, keys were only replaced in motor-operators with motors of 150 ft-lbs and greater.
The motor on LPCS-MO-5 was 100 ft-lbs.
As corrective actions, the licensee:
(1) reviewed all pertinent NRC INs and Part 21 Notices to ensure that previous corrective actions were appropriately taken; and (2) performed additional engineering analysis to identify other motor-operators (not specified in the INs) that could be susceptible to similar failures.
The licensee identified seven other motor-operators that met the criteria for key replacement and installed 4140 keys in those operators during the current outage.
The inspectors reviewed the final PER and corrective actions and considered the licensee's evaluation to be thorough and sound.
The failure to take appropriate actions in response to IN 94-10 (to preclude the LPCS-MO-5 motor pinion key failure) was a violation of
CFR Part 50, Appendix B, Criterion'VI (Corrective Actions) which requires, in part, that measures be taken to assure that conditions adverse to quality are promptly identified and corrected.
Because the licensee identified and corrected the specific problem and identified the failure to appropriately resolve a previous IN, and the valve remained degraded but operable, this violation is being treated as a noncited violation, consistent with Section VII.B.1 of the NRC Enforcement Policy (NCV 50-397/9608-03).
f
-18-E8.2 E8.3 Closed Violation 50-397 9434-03:
holes cut in the main control room floor precluding control room pressurization.
This violation described the causes and corrective actions for an event that rendered the control room ventilation system inoperable, not capable of pressurizing to TS limits, as a result of creating two holes in the control room floor.
The licensee determined that the root cause of this problem was an inadequate procedure.
The procedure did not provide the work planner sufficient guidance to recognize the effect that the work would have on safety-related equipment.
The licensee's immediate corrective action was to plug the holes and verify the integrity of the control room ventilation boundary.
The more significant long-term corrective actions included; revising procedures, training plant and engineering staff on barrier impairments, and including'he related licensee event report (LER) in industry experience training.
The inspector verified that the licensee completed the proposed corrective actions and considered these actions adequate.
Closed LER 50-397 94-21; holes cut in the main control room floor precluding control room pressurization.
This LER described the causes and corrective actions for an event that rendered the control room ventilation system inoperable, not capable of pressurizing to TS limits, as a result of creating two holes in the control room floor.
This item was the subject of the above described violation (Section E8.2)
and is closed based on that review.
IV. Plant Su ort R2 R2.1 Status of Radiological Protection and Chemistry Facilities and Equipment RHR Tem orar Shieldin a.
Ins ection Sco e
71750 I
On May 22, 1996, the inspectors toured the reactor building to assess the room material condition and radiological posting practices.
b. Observations and Findin s
The inspectors assessed RHR Heat Exchanger Room B for material condition and housekeeping and observed a sheet of temporary lead shielding, which was suspended from scaffolding, contacting Snubber RHR-998N.
The inspectors were concerned this could affect the operability of the snubber.
The licensee initiated PER 296-0422 to address this problem and a work package to trim the lead shielding.
The inspectors reviewed PPM 1.3.44,
"Control of Temporary Shielding,"
Revision 7, Attachment 8.2,
"Temporary Shielding Installation Guidelines,"
and found that the procedure provided adequate guidance.
In discussions with the technician who installed the shielding, the
-19-,
inspector learned that the technician had noted the contact during installation of the shielding and bent the lead sheet away from the snubber.
The inspectors concluded the contact between the temporary shielding and the snubber resulted from inadequate attention and workmanship during installation to assure lead shielding remained in its intended position.
The snubber program engineer evaluated the as-found condition and determined that the snubber's operability was not affected.
In addition, RHR Train B was inoperable for other reasons at the time the concern was identified.
C. Conclusion Fl Fl.l To ensure that operability of safety-related equipment is not challenged, attention to detail during the installation of temporary shielding could be improved.
Control of Fire Protection Activities Control of Fire Doors While touring the radiological controlled area, the inspector observed a
weakness in the control of fire doors.
The licensee propped open the 441 foot access doors, fire doors, to the radiologically controlled areas for extended periods for convenience to facilitate entry and exits.
The licensee did establish compensatory fire watches for these impairments.
The, licensee propped open the east door to improve ventilation and reduce the area temperature.
The inspector noted that even on days when the temperature was comfortable the door remained propped open.
The licensee propped open the west door for two reasons:
to improve ventilation to reduce area temperature and to accommodate and abnormal turbine building ventilation lineup.
The inspector found that the licensee did not close the fire door after the turbine building ventilation lineup was restored to normal.
The licensee was responsive to the inspectors'oncerns and closed the fire doors.
The licensee plans to review their control and tracking of impaired, fire doors.
The inspectors concluded that the licensee propped open fire doors for convenience and did not restore these doors in a timely manner when opening was no longer require V. Kana ement Neetin s
Xl Exit meeting Summary The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on June 28, 1996.
The licensee acknowledged the findings presented.
The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary..
No proprietary information was unidentifie PARTIAL LIST OF PERSONS CONTACTED Licensee P.
Bemis, Vice President for Nuclear Operations L. Fernandez, Licensing Manager G. Smith, Plant General Manager C. Schwarz, Operations Manager J. Swailes, Engineering Director D. Swank, Regulatory and Industry Affairs Manager R. Webring, Vice President Operations Support NRC J. Clifford, Senior Project Manager, NRR
-22-INSPECTION PROCEDURES USED IP 37551: Onsite Engineering IP 60710: Refueling Activities IP 61720:
Containment Local Leak-Rate Testing IP 61726: Surveillance Observations IP 62703:
Maintenance Observations IP 71707:
Plant Operations IP 71711:
Plant Startup from Refueling IF 71750: Plant Support Activities IP 92902:
Followup - Engineering IP 92903:
Followup - Maintenance IP 92904:
Followup - Plant Support
~0eeed ITEMS OPENED, CLOSED, AND DISCUSSED 50-397/9608-01 URI wetting MSIVs prior to leak rate testing 50-397/9608-02 'RI failure to perform IST of safety-related valves 50-397/9608-03
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NCV LPCS-MO-5 motor pinion key failure Closed 50-397/9419-01 50-397/9434-03 50-397/9606-03 50-397/94-21 VIO dropped lights into reactor vessel VIO holes in control room floor preclude pressurization URI LPCS-MO-5 motor pinion key failure LER holes in control room floor preclude pressurization
-23-LIST OF ACRONYMS USED FPC FSAR IN IST LER LLRT LPCS MSIV NCV NRC PER PPM RCIC RHR RPS SFP SRM SRO TS URI VIO WNP-2 fuel pool cooling Final Safety Analysis Report Information Notice in-service testing licensee event report
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local leak rate testing low pressure core spray main steam isolation valve noncited violation U.S. Nuclear Regulatory Commission problem evaluation request plant procedure manual reactor core isolation cooling residual heat removal reactor protection system'pent fuel pool source range monitor senior reactor operator Technical Specifications unresolved item violation Washington Nuclear Project-2