ML20059D084
| ML20059D084 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 08/24/1993 |
| From: | Richards S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| References | |
| OLA-2-I-MFP-139, NUDOCS 9401060409 | |
| Download: ML20059D084 (18) | |
Text
. . _
@~4tJ 3 f 3 - u t. n -
up * . . - -
t
/k/F#-/3 188518 NEtASRtc RY
- /pa asc,** UNITED STATES AFFAIRS
, !' k h/N 93 NUCLEAR REGULATORY
'N. f i REGION v COMMISSION T,P. APR 201992 7 l $- Idb 14s0 MARIA LANE i
l
\ , s, *.#Y[ WALNUT CREEK. CAUFORNIA 94s9q. HD DISTRIBUTION T 28 P ; .AHRON l RMS ONLY.
Apg y 7 l
l Pacific Gas and Electric Company ;
l 77 Beale Street ,
i San Francisco, California 94106 -
Attention: Mr. G. M. Rueger, Senior Vice President and General Manager Nuclear Power Generation Business Unit Gentlemen:
l
Subject:
NRC Inspection of Diablo Canyon Units 1 and 2 l This refers to the routine inspection conduct'ed by H. Wong and M. Miller curing the period of February 4 through March 16, 1992. This inspection ;
examined your activities as authorized by NRC License Nos. DPR-80 and DPR-82.
At the conclusion of the inspection, di:,ct.ssions of their findings were hCd '
with you and other members of your staff.
Areas examined during this inspection are described in the enclosed inspection repcrt. Within these areas, the inspection consisted of selective l
examinations of procedures and representative records, interviews viith ,
perstonel, and observations by the inspectors. ;
s U
NC violations or deviations were found within the scope.of this inspection. i However, aoparent weaknesses were identified in the maintenance performeo on sore of the backdraf t dampers of the containment f an coolers unit in Unit I and in the actions taken to correct the deficiencies. These weaknesses could have led to the inoperability of three containment fan cooler units during certain design bases accidents. This matter, which was discussed.at the management meet M held on April 2, 1992, is still being reviewed and will be adoressed in a future inspection report.
i In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of
! this letter and the enclosure will be placed in the.NRC Public Document Room.
i
) Should you have any questions concerning this inspection, we will be pleased to discuss them with you.
Sincerely, S. . Richards, Chief ;
Reac or Projects Branch
Enclosure:
Inspection Report Nos. 50-275/92-05 and 50-323/92-05 RECEIVED ,
i APR 211992 l com, E."Sm c.am 9401060409 930824 8
r DR ADOCK 05000275 C PDR
j -- A 2~
-.*,LJ - 1 4 - .6- 5 6
l s
! 0 9
e il i &,
1
)
l 1
I l 1
i i
1 1
l e I 1
i
)
4 4
8 1
l i
s l
l l
I bQ e,D t
I w.
m N 10 1
\% i b h.l0 RW 8
v' h ,9
? M o l
,[ UU k
~
~
_e. ,
"i % g b I N l s
F1 =
[J .
p 2
i 1
a y 's g .. 1 C'r e
u Q v _
s x.'$ i cc W ?
s 1
l
$ V'. Q &< & a d b -j ,% ',i I 3O i4T ; -Q-.. )
~
N e-x .f
'e _o N 3 %%
! l !
$1B % =
1 3'
^f > t
~
-I y 6 tc 3
. e L T e 9
' N 4 M E O b d i
1
,r7.-- e , . , - - , *, .m,e , , , .- - w, 4 ,--, , r-,< m- - - w , # ,- 2
. .- - - . . - .- . . - - . .- =_ ~- . - . - . _ . .
188518 U. S. NUCLEAR REGULATORY COMMISSION REGION V Report Nos: 50-275/92-05 and 50-323/92-05 Docket Nos: 50-275 and 50-323 License Nos: DPR-80 and DPR-82 Pacific Gas and Electric Compary-
~
Licencaa:
77 Beale Street, Room 1451 San Francisco,~ California 94106- ,
Facility Name: Diablo Canyon Units I and 2 laspected at: Diablo Canyon Site, San Luis Obispo County, California Inspection Conducted: February 4 through March 16, 1992)
Inspectors: H. Wong, Senior Resident inspector M. Miller, Resident inspector
-i 2-Acoroved ty:
P. H. Jo son, Chief, Reactor Projects Section 1 Date Signed Errary:
Insoectio from Feb uary 4 through Parch 16, 1992 (Recort'Nos. 50-275/92-05 ,
arc H -323 eE-M ) s Areas inspected: The inspection included routine inspections of-plant l
cperations, maintenance and surveillance activities, followup of_onsite ;
- I j
events, open items, and licensee event reports (LERs), as well-as selected l independent inspection activities. Inspection. Procedures 61726, 62703, 71707,-
71710, 92701, and 92703 were used'as guidance during this inspection.- q Safety Issues Management System (SIMS)' Items: None- l l
Results:
General Conclusions on Strengths and Weaknesses:
Strengths - ,
The licensee organizations quickly responde1 to the unexpected. outage of 1
Urit i after the reactor trip of March 6, 1992. .The meetings conducted appeac eo effective in determining the actions necessary.to; determine th'- e l of the f eedwater puttp loss and reactor trip and in prioritizing .
catte work activities. 1 e
188518 '
The licensee response to a quadrant power tilt alarm during a power reductior, in Unit 2 on February 16, 1992 was timely and occurate.
Licensee personnel discussed the matter with Westinghouse representatives and carefully monitored core flux parameters to assure that the reason for the alarm was understood and that core flux behavior was as expected.
Weaknesses -
During investigation of the problems with containment fan cooler units (CFCOs), it was identified that installation errors had occurred which had gone undetected and had caused the CFCUs to be inoperable for a .
significant period of time. In addition, a problem with reverse rotation of a CFCU had been identified, but had not been considered abnormal and, therefore, no actions were taken to correct the problem. This appears to indicate a need for greater attention to those work activities considered relatively simple and the need for more thorough and timely corrective actions f or deficient conditions.
This matter is still being reviewed and will be discussed in a future inspection r rort.
Sionificant Safety Matters: None Summar. of Violations: None Opsn Itcre Sum ary:
1 nea iten opened, 7 items closed, and 2 items remain open.
l I
t
188518 DETAILS
- 1. Persons Contacted Pacific Gas and Electric Company
- G. M. Rueger, Senior Vice President and General Manager, Nuclear Power Generation Business Unit J. D. Townsend, Vice President and Plant Manager, Diablo Canyon Operations
- W. H. Fujimoto, Vice President, Nuclear Technical Services
- D. B. Miklush, Manager, Operations Services
- M. J. Angus, Manager, Technical Services
- B. W. Giffin, Manager, Maintenance Services
- W. G. Crockett, Manager, Support Services J. E. Molden, Instrumentation and Controls Director
- W. D. Barkhuff, Quality Control Director R. P. Powers, Mechanical Maintenance Director
- D. A. Taggart, Quality Performance and Assessment Director
- T. L. reebel, Regulatory Compliance Supervisor H. J. Phillips, Electrical Maintenance Director
- R. C. Anderson, Manager, Nuclear Engineering and Construction Services
- M. R. Tresler. Project Engineer, Nuclear Engineering and Construction Services J. A. Shoulders, Onsite Project Engineering Group Manager S. R. Fridley, Operations Director R. Gray, Radiation Protection Director
+2. J. Griffin, Senior Engineer, Regulatory Compliance J. V. Boots, Chemistry Director
- J. B. Hoch, Manager, Nuclear Safety and Regulatory Affairs
- T. A. Moulia, Assistant to Vice President Diablo Canyon Operations
- C. A. Dougherty, Quality Assurance Senior Supervisor
'J. E. Tompkins, Nuclear Safety and Regulatory Affairs Director
- R. C. Russell, Nuclear Safety and Regulatory Affairs
- c. A. Dettman, Director Nuclear Opt..; ions and Suppurt, and Assistant to Senior Vice President Nuclear Regulatory Commission
- S. A. Richards, Chief, Reactor Projects Branch, Region V
- Denotes those attending the exit interview.
The inspectors interviewed several other licensee employees including shift supervisors, shift foremen (SFM), reactor and auxiliary operators, maintenance personnel, plant technicians and engineers, and quality assurance personnel.
- 2. Operational Status of Diablo Canyon Units 1 and 2 Durirg the inspection period, Unit 1 operated at 100% power, except between March 6 and 10, 1992 when a plant trip occurred due to the loss of a aain feedwater pump. This event is discussed in paragraph 4.a below.
Unit ? operated essentially at 100% power, except for February 16-17
2 188518 and March 14-15, 1992. On February 16-17, Unit 2 reduced power to approximately 50% for condenser cleaning. On March 14-15, Unit 2 reduced power to 55% to install a wiring change to the power supply circuits to speed sensors of both main feedwater pumps.
On February 16, 1992, at 10:30 p.m. when Unit 2 Thereduced power to 50%, a power tilt ratio quadrant power tilt ratio alarm was received.
increased to 1.048 at about 1:30 a.m. on February 17 and then steadily decreased. The licensee, in consultation with Westinghouse, determined that the power tilt was a result of slightly different efficiencies of the secondary loorts, resulting in uneven power distribution in the core.
This causes about a 0.5% power tilt during normal 100% power operation, whicn is within the requirements of Technical Specifications. The decrease in power resulted in an uneven xenon transient in the core, which exaggerated the power tilt. The licensee decreased power below 50%
as required by Technical Specifications and the power tilt decreased during power ascension and stayed within Technical Specification limits.
On March 14, 1992, Unit 2 again reduced power and a similar power tilt occurred and exceeded the Technical Specifications limit of 1.02 at 11:30 p.m. The tilt decreased to less than 1.02 at 1:05 p.m. on March 15.
During this time, the licensee reduced the high flux trip points to 94't as required by Technical Specifications.
- 3. Operational Safety Verification (71707)
- a. General During the inspection period, the inspectors observed and examined act1vities to verify the operational safety of the licensee's facility. The observations and examinations of those activities we re conducted on a daily, weekly or monthly basis.
On a daily basis, the inspectors observed control room activities to verify compliance with selected Limiting Conditions for Operations '
(LCOs) as prescribe- .n the facility Technical Specifications (TS).
Logs, instrumentation, recorder traces, and other operational records were examined to obtain information on plant conditions and to evaluate trends. This operational information was then evaluated i
to determine if regulatory requirements were satisfied. Shift j turnovers were observed on a sample basis to verify that all i pertinent information on plant status was relayed to the oncoming crew. During each week, the inspectors toured the accessible areas of the facility to observe the following:
(a) General plant and equipment conditions (b) Fire hazards and fire fighting equipment (c) Conduct of selected activities for compliase with the licensee's administrative controls and approeed procedures (d) Interiors of electrical and control panels h) Plant housekeeping and cleanliness (f) Engineered saf ety f eature eouipment alignment and conditions (g) Storage of pressurized gas bottles i
3 188518 The inspectors talked with operators in the control room and other plant personnel. The discussions centered on pertinent topics of' general plant conditions, procedures, security, training, and other aspects of the work activities.
- b. Radiological Protection The inspectors periodically observed radiological protection practices to determine whether the licensee's program was being implemented in conformance with facility. policies and procedures and in compliance with regulatory requirements. The inspectors verified that health physics supervisors and professionals conducted frequent planttourstoobserveactivitiesinprogressan[wereawareof:
significant plant activities, particularly those related to radiological conditions and/or challenges. ALARA~ considerations were found to be an integral .part of each RWP (Radiation Work Permit).
- c. Physical Security Security activities were observed for conformance with regulatory" reauirements, implementation of the site security plan, and administrative procedures including vehicle and personnel access screening, personnel badging, site security force manning, compensatory measures, and protective and vital area integrity.
Exterior lighting was checked during backshift inspections.
NC violations or deviations were identified.
4 Onsite Evert followup (93702)
- a. Unit 1 Loss of Main feedwater Pump and Reactor Trip Dr. M.zrch 6. 199? at approximately 10:30 am, with Unit 1 operating at 1001 power, main feedwater pump 1-1 tripped on overspeed due to the failure of the power supply to the speed sensors of the pump.- At the time of the event, licensee personnel were working on a lube oil cooler for the pump which was adjacent to the speed, sensor power supply cabinet. On the loss of the power supply to the speed sensors of feedwater pump 1-1 and the failure of the transfer to the power supply of the other feedwater pump, the sensed speed of pump 1-1 went to zero and the pump control system attempted to increase ,
pump speed. This led to feedwater pump 1-1 reaching the'overspeed !
trip setpoint of approximately 6400 rpm.~ The other feedwater pump l decreased its speed due to the increased speed of feedwater pump l l-1. Control room operators initially reduced turbine load in During the load
~
response to an expected main feedwater pump loss.
reduction, feedwater pump 1-1 tripped.
4 188518 )
l On the loss of feedwater pump 1-1, control room operators initiated ]
a rapid reduction of turbine load and had lowered loid to approximately 50%, with the 40% steam dump valves (to condenser) J modulating open as designed. However, this was not enough to prevent a reactor trip on low-low' steam generator water level approximately one and one-half minutes after the trip of feedwater pump 1-1. Contributing to the difficulty in recovering steam generator water level was the fact that the operating feedwater pump (1-2) had reduced its speed due to the' increased speed and flow from the overspeeding feedwater pump 1-1.
Plant response to the reactor trip was for the most part )
uncomplicated and there was no initiation of safety injection. The j issues requiring resolution were: (1) the identification of the ;
cause of the speed sensor power supply. failure to feedwater pump .
! 1-1; (2) determination of the cause of three 10% steam dump valves (tc atmosphere) opening after the reactor trip; and (3)
I determination of the cause of the start of diesel generator 1-1 approximately 32 seconds after the reactor trip. These issues are discussed below.
l
- 1) The cause of the power supply failure of the-speed sensors to feedwater pump 1-1 was identified to be the opening.of e-fusible link in an inverter. The cause of the link opening was not determined, but was possibly due to vibration when a portion of platform grating adjacent to the power supply cabinet was dropped into position, or due to overheating due tc the high ambient temperature. The opening of the fusible link caused excessive current to be drawn and the input' fuses were blown resulting in the failure of the power supply. A single-
. rcotr supply provides power to both speed sensors of a-feedaater pump. A transfer sch: e from the other feedwater- !
tuip power supply was intended to provide redundancy, !
, The failure to transfer to the power supply of feedwa' ~ pump 1-2 was found by the licensee to be due to a strall plastic ,
l shaving discovered between relay contacts. This shaving .
i appeared to be from the terminal insulation on the relay and appears to have fallen into the contact area during installation of the wiring.
A new inverter was installed for feedwater pump 1-1 and relays I and fusible links were inspected in feedwater pumps.1-l'and
! 1-2. The primary corrective action to prevent another loss of power supply, possibly resulting in a similar event, was a wiring change to provide independent. power supplies for the'two speed sensors on each feedwater. pump. This. change provided for-one speed sensor on each pump to be powered from an inverter, with a transfer scheme to the other feedwater pump power supply
- f necessary (same as in the past configuration).
In addition, the other speed sensor on each pump would be powered by other AC sources. This change in power supply configuration was installed on the Unit 1 feedwater pumps prior to the restart of l
l l
t i
l .
l 5 188518
! Unit I and was also completed on the Unit 2 feedwater pumps during a power reduction on Marcb 14-15, 1992.
- 2) The opening of three of the four 10% steam dump valves was found to be due to the sensitivity of the steam dump valve control system. Volume boosters had been added to the pneumatic portion of the control system during the last outage, l
but the boosters had not been tested to assure that adverse l
effects were not present during transient conditions.
l During the March 6, 1992 event, the steam dump control system switched from the load reject mode of control (during the manual load reduction) to the steam pressure mode of control, (after the reactor trip). In the load reject mode of control the no-demand pressure in the actuator lines is 0.5 psig'and in the steam pressure control mode the no-demand pressure in the-actuator lines is 3.0 psig. In the switch from load reject to steam pressure mode of control, a pressure perturbation was started (on the change from 0.5 tn 3.0 psig) in the actuator' lines which was amplified by the recently added volume
- boosters. Because the steam dump valves will begin to open at l
approximately 3.5 psig, this caused the slight opening of the l 10% steam dump valves duririg the event. This was demonstrated in subsequent testing of the steam dump valve control system.
The opening of three 10% steam dump valves rather than four was because one valve (PCV-22, the one which had not opened during the event) was found to be set at a lower no-demand pressure (1.0 psig) rather than at the intended 3.0 psio.-
If necessary, control room operators could have opened or closed the 10% steam dump valves using manual switches in the ccntrol room. By March 19, 1992, adjustments had been made to all of the steam dump valves in Units 1 and 2 to tune out the pressure perturbation.
- 3) The start of diesel generator 1-l'was due to the relatively light loaoing of its associated bus (Bus H) and the fact that its voltage drops slower than the other two safety-related 4160 V buses. This causes the other buses to load first onto the startup bus which momentarily brings down the voltage of the startup bus. The drop and recovery of startup bus. Voltage was sufficiently long to initiate the start of diesel generator j 1-1 without loading it onto Bus H. Bus H successfully l
transferred from the auxiliary transformer to the startup bus l in this event. This condition had been seer before at Diablo i Canyon and was evaluated in NCR DC2-88-EM-N095. The licensee concluded in this NCR that an unnecessary diesel generator start was considered an acceptable occurrence when considering the ineffectiveness and cost of a possible design change.
Because a design change could possibly prevent a necessary i diesel generator start, licensee management considers the start i of a oiesel generator when it may not be needed to be less of a concern than the failure of a diesel generator to start when recuired.
l l
l 188518 6
l Other activities accomplished during the unit outage included inspection and repair of the backdraft dampers associated with the l Unit I containment fan cooler units (CFCUs).
1
! The CFCU work included post-maintenance testing after completion of i all work to assure there was no reverse rotation of the fans when shut down. The licensee also identified a cracked packing gland flange on a manual drain valve (513) on the pressurizer spray line.
The manual drain valve had wedges installed as a temporary measure to retain pressure on the packing by 1ransmitting force through the packing f ollower ring. The che is also normally closed with system pressure under the valve disc.
l The plant resumed operation on March 9, 1992, after repairs and work i activities were completed, and 100% power was reached on March 10, l 1992. The reactor trip event is described in NCR DCl-92-EM-N010, and a Licensee Event Report will be submitted to the NRC.
I Inspector Findings -
l The inspector observed the response to this event several 1.
minutes after the indication of feedwater pump problems.
Control room operators responded appropriately and utilized proper procedures. Control room instructions and guidance were clear and concise. Sufficient management and licensed personnel were present to provide assistance during the evert.
- 2. The inspector observed several management meetings which were ccnducted following the event. These meetings were to evaluate the causc of the feedwater pump trip and reactor trip and deterrr.ine corrective actions. Meetings were also conducted ic deternine the scope and priority of work to be accomplished during this unexpected outage. These meetings were successful in making licensee management aware of the root cause determinations, obtaining agreement on necessary corrective actic>ns, and in m edinating and prioritizing work activities.
- 3. The inspector reviewed the past history of feedwater pump inverter failures. This review indicated that there have been several inverter failures in the past which appear to be indicative of a long standing problem. The inspector will continue to review the inverter history during the next inspection period.
- b. Diesel Generator 1-3 Test with Cardox System Initiation On March 11, 1992, during routine surveillance testing of diesel generator 1-3, the carbon dioxide fire suppression (Cardox) system actuated shortly after the diesel generator started. Because the actuation of the carbon dioxide system causes roll-down doors to shut and thus deprive the generator of cooling, operators immediately shut down the diesel, opened the roll-down doors, and startec troubleshooting the failure. The licensee also tested the othcr diesel generators in both units within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to confirm operability as required :.j Technical Specifications. The licensee
4-1~ 7 1G8518 1*:
was unable to confirm the cause.of the failure and has sent a-l circuit card from the carbon dioxide system to the vendor for root
! cause analysis. The licensee's determination of the cause of this f ailure and whether it is considered a valid f ailure of the diesel generator will be reviewed in a subsequent inspection (Followup Item 50-275/92-05-01).
No violations or deviations were identified.
- 5. Maintenance (62703) i The inspectors observed portions _of, and reviewed records.on, selected ma....enance activities to assure compliance with approved procedures, Technical Specifications, and appropriate industry codes and standards.
Furthermore, the inspectors verified maintenance activities were performed by qualified personnel, in accordance with fire protection and
~
housekeeping controls, and replacement parts were appropriately-certified. These activities included:
- Work Order R0091799, Unit 2 RHR Pump 2-2 Motor Oil Sampling
- Work Order C0096975, Unit 1 CFCU 1-2 Backdraft Damper. Repair ,
- Work Order C0096844, Unit 2 Inspection and Repair of FCV-372 (AFW l Check Valve)
- Work Order C0096503, Unit 1 investigation and Repair of'DG.-1-2 Engine Temperature Indicator
- Work Order C0097368, Unit 2 Inspection and Cleaning of Main
, Condenser On Fetruary 25, 1992, the inspector observed. oil sampling of the Unit ? residual heat removal (RHR) pump motor 2-2. The' associated work oroer (WO R0091799) specified that'the sampling be performed in accordance with maintenance procedure MP M-56.7, Steps 7.1.a through-7.1.g. Included in these steps was a note that the " oil samples should be taken while the equipment is operating or shortly after l
! the equipment is shutdown." The person taking the oil sample was unsure of how the note was to be-interpreted because there were no specific steps in the work' order calling.for the operation of the-RHR pump. The RHR pump had not been recently run prior.to drawing the oil samples.
The inspector discussed the intent of the work order.and mainten c e procedure with an electrical maintenance manager and the sponsor
- the procedure. The conclusion was that the intent was to obtain 7.n oil sample after a pump run, but it was-recognized that the procedure and work order were inconsistent and that appropriate scheduling guidance had not been communicated to properly coordinate.
the pump run and drawing of the oil samples. Licensee management-l stated that the procedure would be clarified, particularly Appenoix
- 8.7, ar.c that scheduling guidance would be provided to properly l
i
8 188518 .
coordinate the oil sampling, in addition, guidance would be provided to maintenance personnel to assure that work orders and procedures are understood prior to continuing work activities.
No violations or deviations were identified.
- 6. Surveillance (61726)
- a. Observations By direct obs'ervation and record review of selected surveillance testing, the inspectors checked compliance with TS' requirements and plant procedures. The inspectors verified that test equipment was calibrated, and that test results met acceptance criteria or were appropriately dispositioned. These tests included:
- STP M-9A Unit 1 Monthly DG Test (DG 1-2)
- STP M-8A2 Unit 1 Emergency Air Lock Leak Test STP v-3R5 Unit 1 Stroke Test of FCV-95 (AFW Steam Admission Valve)
- STP M-16C Unit 2. Train A Slave Relay Test.
STP M-8Al Unit 1 Personnel Air Lock Leak Test
- b. During the testing of the steam admission valve (FCV-95) to the Unit I turbine-driven auxiliary feedwater pump on March 10, 1992, licensee personnel identified that on the closure of valve FCV-95 there was a larger than expected amount of stem bending. Stem bencing in the past had been viewed as one indicator that a higher force would be reauired to open the valve the next time. Licensee engineering personnel concluded that FCV-95 should be:restroked and monitored with the expectation that the bending stress-~would %
reduced. The valve opened successfully (11.89 amps for pullout):
however, on closing, licensee personnel noted even higher bending ~
l stresses than previously seen and also a deeper seating. position.
than previously. This valve position was also considered undesirable in that both conditions indicated a higher force would be required to open the' valve.
The valve was cycled open again, with 15.65 amps measured to open the valve. On closure of the valve, while an even deeper seat position was noted, only slight stem bending occurred. This was considered acceptable by licensee engineering personnel. -The-next test of FCV-95 occurred on March'17, 1992 and resulted in successful opening of the valve and approximately 15.5 amps on pullout.
The licensee's engineering organization (NECS) has been evaluating' the data obtained during the testing of FCV-95. In a memorandum dated March 5, 1992, NECS has proposed that the weekly testing'of-FCV-95 in Unit I be extended to every two weeks for March and April I -
l I-
-i e - y -n--, A- - v e -
m
9 188518 and then to monthly beginning in May 1992, due to the idantification of a reliable indicator of the potential for high pullout force i (stem position), and to a lesser degree the reliability of stem !
bending as an indicator. These indicators have been correlated, j with stem position being more consistent. An additional reason for 1 J
reducing the testing frequency is the gearing changeout done after the December 24, 1991, failure of FCV-95 to open. This change has increased the opening capability of the operator such that there is added assurance that the valve will open when intended. NECS has recomended that thermography and certain diagnostic equipment no longer be required during valve testing as the information gathered by these instruments is no longer of use. ,
Further, NECS indicates that the valve has begun to show evidence of wear during the weekly testing. There has been a gradual increase in the pullout current to open the valve and a decrease in the seating thrust in closing the valve. This is
- attributed to greater stem friction'as the '!alve is cycled and is postulated by NECS-to be due to the gradual reduction of lubrication on tha stem threads. -
The March 5,1992 memorandum also recomends that FCV-95 in Unit 2 be instrumented to monitor only stem position rather than the full set of instrumentation now provided on the Unit 1 valve.
Based on a review of the licensee's data correlating stem position and pullout current, the gearing changeout made, and the fact that Unit 2 opening current data are relatively consistent (6.2 - 12.5 amps), the licensee's proposal appears appev;riate.
Nc violations or deviations were identified.
- 7. Eroineered Safety Feature Verification (71710)
Dur mg the inspection period, selected 7,rtions of the auxiliory feedwater system for Units 1 and 2 were inspected to verify that system configuration, equipment condition, valve and electrical lineups, and local breaker positions were in accordance with plant drawings and Technical Specifications.
No violations or deviations were identified.
l
- 8. Open Iten Followup (92703) l
- a. Unresolved item 50-275/91-01-01: Licensee Appendix R Audit (0 pen)
In December 1990, the Nuclear Operations Support (N05) group performed an audit of safe shutdown systems.and procedures. This audit was described in a memorandum to the plant manager dated January 29, 1991. Followup by the NRC inspector is described below. ,
Licensee corrective actions to date and planned corrective actions l are included with each issue.
l :
i l
10 188518 After completion of design basis documentation in the areas of electrical circuitry and safe shutdown methodology, which are estin,ated to be completed around June 1992, the licensee plans to start the second phase of the Appendix R validation budgeted under BLI No. 453. This will include validation of the safe shutdown time ~M
~
line; emergency lighting locations, including availability of emergency lighting during outages (formerly NRC Open Item 91-03-03, l closed in this report); equipment repair procedures; and use of l auxiliary equipment.
l As a result of the 1990 NOS audit, the licensee identified inproper implementation of configuration. control in the following areast o Emergency Procedures - Procedures are currently being' reviewed l for agreement with design basis documentation. -
o Combustible Loading - The licensee is evaluating the need for control over small (less than one percent) changes in the i combustible loading.
o Safe Shutdown Circuit Analysis - The licensee is completing validation of the original safe shutdown circuit analysis.
Several issues regarding circuit separation have been identified, including the inadequate separation in the Unit 2 reactor trip switchgear room (Inspection Report 50-275/92-01, paragraph 5.b.), inadequate protection of cables in the Unit 2 diesel generator room corridor, and inadequate separation of circuits in containment, which the licenset plans to report in a future LER.
Additior:a1 specific concerns being followed by the NRC are listed t+1a:
o Remote Shutdown Panel - The licensec stated that surveillance testing of hot shutdowr, panel mechanical components would be in place by December '" ,1991. Most of this testing is in place; however, testing for some valves has not been implemented. The licensee stated that testing would be implemented by July 15, 1992. For the testing which has been implemented, no failures have been noted to date.
- , Operator Training - The licensee documented that the training for procedure M-10, " Fire Protection of Safe Shutdown Equipment," has occurred and has now been included in biennial review for operator requalification training. Therefore, this specific concern is closed.
o Heat Trace - Heat tracing circuity for the Appendix R emergency boration flow path was not reviewed to validate appropriate circuit separation. The circuity runs through various-areat, of thC auxiliary building, and may not be properly separated. The licensee agreed to address this issue. The licensee stated -
that boration can be accomplished using the RWST for all fire areas. Therefore, redundant capability is available.
1 1
- ~
- 11 l,c 188518; t
o Failure Trending of Emergency Lighting Units - Although the licensee now has the capability to trend failures of individual 4
I l
lighting units, there has been no assigned responsibility for this function.
o Equipment Operability Evaluations'- Revision 6 of Justification l for Continued Operation (JCO) 90-17 addressed inoperability of- i the positive displacement charging pump, which would be used l for a safe shutdown during a fire'in the centrifugal charging -l 2; pump room. The JC0 did not specifically'. identify the safe . i shutdown method nor its associated circuit analysis to be used l as compensatory measures. The licensee agreed that the JC0 4
could h:ve.been written more clearly. The pgj g i
rewrite the JC0 by mid-April 1992 and will disc!
uss ne plans.to the issues 3 at that time. Since the licensee stated that fire protection compensatory measures are'in place in the centrifugal charging.
pump room, including suppression, detection, and hourly fire 2 watches, the safety significance of this issue.is low. '
N ,
i o Root Cause - The preliminary 1991 audit. conclusion for these' discrepancies involved a disconnect between NECS and the actual plant configuration and procedures. The licensee stated during this report period that that root cause still appeared to be 4 valid.
) L. folicwup Item 50-275/92-01-01: Removal of Conduit Bracket During Unit 1 Refuelinc Outage (Closed)
This issue involved the discovery by the NRC inspector of a pair of conduit brackets which had been removed during the Unit I refueling
- outage in 1991 and had not been reinstalled. The lica"ee initiated
- OE 00009474 te document the corrective actions taken. These actions included reinstallation of the brackets, discussions with the personnel involved in the incident, and the initiation of a training improvement plan specific to this event which will be included in
! the Quarterly maintenance training.- Based on these actions, this 2 item is closed.
j c. Followup ltem 50-275/92-01-02: Timing of Independent Verification During Surveillance Testing (Closed)
This issue involved the inspector's. observation that personnel did not clearly understand the appropriate timing of the performance of independent verification _during surveillance testing. The licensee i
initiated QE Q0009487 to documtat.the issue and the required corrective actions. These actions include the revision of selected surveillance procedures; revision of procedure NPAP C-104, independent Verification of Operating Activities, to clarify the intent of the independent verification during surveillance testing; development of a checklist to be used during interdepartmental i reviews of surveillance testing procedures; and revision of the Operations Policy documents to place guidance regarding independent verification in the proper location. Based on these actions, this item is closed.
12 188518 i <
(s .
- d. Unresolved item 50-275/91-24-01: Flooding of Component Cooling Water (CCW) Heat Exchanger Room (Closed) ,
N On August 2, 1991, during preparations for a routine cleaning of the CCW heat exchanger, maintenance workers f ailed to properly follow the work order clearance instructions. This resulted in flooding of the Unit 1 CCW heat exchanger room. The work order. required that the auxiliary ~. saltwater system inlet valve to the heat exchanger be gagged before the heat exchanger manway was opened.
Instead, with the inlet valve shut but not gagged, workers opened -
the manway, entered and exited the heat exchanger, and then attempted to gag the valve shut. The. workers thought the work order steps were in the wrong sequence. The nianway was left open. . During the attempt to gag the valve, the valve opened, flooding'the heat.'
exchanger area. LIhis could have caused. a significant personnel safety hazard had Ap individual been in the heat exchanger when the inlet valve opened.,
The licensee determined that there was only minor significance to.
plant safe'", since the redundant safety train was available.
The licensee determined that the root cause of the problem was personal error, in that the instructions in the work order were not followed in sequence. . Corrective actions taken include issuance of two maintenance bulletins; mandatory discussion with all foremen and appropriate journeymen on the importance of reviewing clearances to ensure isolation of energy sources from the work area, and on.the im;mtance of constructive and timely feedback to Work' Planning when a work order is not correct; and counseling of the specific ,
individuals involved. The licensee has also identified other.
prudent, peripheral actions.
Based on the corrective actions, this item is closed.
- e. Followup Item 50-275/91-09-03: Normal Lighting and Public Adluss (PA) Systen. Power Supply (Closed)
During the March 7, 1991, loss of offsite power event, the Unit I normal lighting and PA systams became inoperable. In addition, emergency lighting in containment was. unavailable due to maintenance. The NRC Augmented Inspection Team (AIT) identified a
! need for the licensee to investigate the adequacy of lighting and communications.
l The licensee had earlier identified (in December.1989) that emergency
! lighting for safe shutdown may not'be adequate in all areas of the j plant. Additional walkdowns were performed after the March 7 event..
! Investigation and corrective action is being implemented under'a 1 budget line item for fire protection safe shutdown, discussed in
~
l this report. As interim compensatory measures, portable-lighting i has beEn proVided to operators. Further followup of the lighting i issue will be followed under Followup Item 50-275/91-01-01. l l
- 188518 The AIT concern for the failure of the PA system was reviewed, j including the containment evacuation alarm function. The licensee i determined that the PA system, by design, is not safety-related, and I that placing the PA system on vital power or on an uninterruptable l power supply was not a warranted expenditure in comparison with other projects providing more safety benefit. Based on the above discussion and inclusion of the lighting issue in Followup Item 50-275/91-01-01, this item is closed, l
- f. Unresolved Item 50-275/91-27-01: Containment Air Lock Leak' Tester -
Unit 1 (Closed)
The issues remaining open regarding this followup item were: (1) whether the non-quality classification of the automatic leak tester was appropripte even though it is used to satisfy Technical Specifications requirements, and (2) whether a revision to the current training program was needed.
The non-quality classification of the tester appears appropriate in that *.he licensee's Q-list document specifies that testing devices are not' classified, but are covered in the QA program to maintain l
accuracy within necessary limits. This is consistent with 10 CFR Part 50, Appendix B, Criterion XII, Control of Measuring and Test Equipment.
A revision to the training. program regarding the air lock had been issued just prior to the event. It had been considered that the recently revised training was sufficient to cover the event. Based j
on further questions by the inspector, the ifcensee issued Training Irrprovement Proposal #2300 to emphasize the need to perform a manual test if the automatic tester fails. This training will be l
included in upcoming licensed and non-licensed operator training and also will be included in the next revision of the training lesson auides. Based on this, this unresolved item is closed.
- g. Unresolved item 50-275/91-20-02: Missed Surveillance Test (Closed) i This issue was described in Licensee Event Report 50-275/91-012-00 and was reviewed and closed in Inspection Report 50- 275/91-40.
Therefore, this unresolved item is closed.
l
- h. Followup Item 50-275/92-17-02: Site Strategy for Personnel Errors (0 pen)
This issue involved the inspector's observation that between August and October 1991 there appeared to be a high number of noteworthy l personnel error events. The licensee initiated a human error i reduction plan which includes letters and meetings to express management's expectations for human performance, review training
- regarding independent verification and provide recommendations, develop a training video, provide a policy on pre-job briefings, and develop a tracking mechanism for human errors contained in NCRs and I QEs for review with department directors. This plan is described in Diablo Canyon Objective 3000.01/017.1.
! 14 188518 '
l Most items are currently on schedule, with the only item behind l
schedule being the development of an independent verification I training video tape. This tape is scheduled to be completed by the l end of March 1992.
j., One March 18, 1992, several personnel errors occurred in Unit 2 which resulted in the isolation of the number 1 seal leakoff line of reactor coolant pump 2-4 for approximately 63 minutes. This event was initiated during troubleshooting of the leakoff line flow
' transmitter. While the licensee is reviewing this event, preliminary information indicates the following errors occurred:
- Inadequate tailboard discussion for the involved parties to understand the scope of the work being performed.
- Assumption of shared responsibilities for the troubleshooting activities; however, there was no communication of who was responsible for what aspects of the job.
- Comunications with control room personnel during the work did not develop an understanding of the actions actually being l performed.
- An alarm for high number 2 seal leakoff received in the control room should have alerted control room operators of the incorrect valve alignment.
It appears that pump performance parameters (seal leakoff temperatures, seal differential pressure, and pump vibration measurements) returned to normal after the event and Westinghouse is being consulted to determine whether any significant pump _
decradation occurred. ,
_These personnel errors are significant in that a number of barriers failed to detect the incorrect valve alignment., Based on errors evident during this event and pending the licensee's evaluation of the root causes and whether additional corrective actions are appropriate, this followup item will remain open.
Exit 9.
On March 19, 1992, an exit meeting was conducted with the licensee's representatives identified in Paragraph 1. The inspectors summarized the i scope and findings of the inspection as described in this report, i
i l