IR 05000206/1989018
| ML20247B144 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 08/25/1989 |
| From: | Johnson P, Kirsch D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20247B124 | List: |
| References | |
| 50-206-89-18, 50-361-89-18, 50-362-89-18, NUDOCS 8909120359 | |
| Download: ML20247B144 (15) | |
Text
g.gmy
..s ; j ' m .
-
T- m
, .
-
-
,
,
u , ,
l ,'h '- (~4 - -
-
',
% {LV 4
,
'
F
" ;) 3 ,
D ,
'* U.S. NUCLEAR REGULATORY COMMISSION
'
<
E: REGION V ym . ,
'
'
[ReportNb ~ 50-206/89-18. 50-361/89-18,.50-36'2/89-18 Docket Nos.- .
._50-206,;50-361, 50-362 iLicense_No * DPR-13, NPF-10, NPF-15 Licensee: - Southern California Edison Company -
a ,
.P. O. Box-800, 2244, Walnut Grove Avenuej
,
,
.Rosemead, California 92770
'
,
Facility Name: ' San Onofre Units 1,'2cand 3 a .
Insp'ection;. a t: San.Onofre, San'Clemente, California;g
-
Inspection conducted: June 18.1989throughJuly29,jl989-x Inspectors: F..R. Huey, Senior Resident Inspector
"
C. W. Caldwell, Senior Resident Inspector c
A. L. Hon, Resident Inspector
'
. Accompanying LInspector:
- )[Kirsch,
"U.' M Chief 7-2 & T9 Date Signeti
.
Reactor Saf ty Branch
, Approved By:
.
-
-
./ Johnson, Chief
"/87 Date Signed
"
React' d r Projects Section 3
' '
~
o . Inspection Summary Ins ection on June 18 through July 29, 1989 (Report Nos. 50-?06/89-18,
- 61/89-15, and 50-362/89-18)
~ %
i . Areas Inspected: Routine resident inspection of . Units l', 2 and 3 Operations ,2
. g' Program including the following areas: operational safety verification', z
.T radiological protection, . security, evaluation of plant trips and events,
^
monthly surveillance activities, monthly. maintenance activities,trefueling
- - activities, independent inspection, licensee event. reports, evaluation.of QA'
- program implementation. onsite review committee, nuclear safety group review,
. management safety evaluation involvement, nonlicensed staff training, and
~ '
,
, followup of previously identified items. . Inspection procedures 30703, 35502, 40500,-.41400, 60710, 61726, 62703, 71707, 71710, 90712, 92700, 92701, 93702-were covere '
Safety Issues Management System (SIMS) Items: None i
9 '
i
'
.-
o n
_____--a__-._____umm-_--_m.a___ ___.__.____m
n- _ - - -
.
. . .
,
. =
4 4
-2- Results:
General Conclusions and Specific Findings:
An evaluation of the licensee's quality oversight groups continued during this inspection period. The program appeared to include clearly defined responsibilities and enhanced preparation and documentation of oversight activities, a
.Significant' Safety Matters: Non !
Summary ' of. Vin 1ations:
Two violations were identified during this report period. One '
violation, applicable to Unit 2, concerned inattention to detail by
. operations personnel which resulted.in an atrs.pheric dump valve (ADV)'
being inoperable. The other. violation, app 7fct'le te maintenance o tivities in Unit 1, concerned an inadeque te. procedure which resulted in temporary cables being routed across redundant Class 1E 4KV cable trays. Thus, appropriate train separation was not maintaine Open Items Summary:
During this report period, one new followup item was opened and three were closed; one was examined and left open. In addition, one item was examined and left unresolved.
L______ _ _ _ _ _ _ _ _ -
e ,
b> . e,
-
'
,*s +
. <
,. .
d,
DETAIL . Persons Contacted Southern California Edison Company
- C. McCarthy, Vice President and Site Manager H. Morgan, Station Manager D. Shull, Jr., Nuclear Oversight Manager
- D. Herbst, Quality Assurance Manager D. Stonecipher, Quality control Manager
- R. Krieger, Operations. Manager
- L. Cash, Maintenance Manager
- K. Johnson, Acting Technical. Manager
- K. Slagle, Deputy Station Manager
- D. Nunn, Manager of Engineering & Construction
- M. Merlo, Nuclear Design Engineering Manager
- F. Nandy, Manager of Nuclear Licensin P. Knapp, Health Physics Manager
'
P. Eller, Security Manager C. Chiu, Assistant Technical Manager J. Schramm, Operations h oerintendent, Unit i V. Fisher, Operations ,a atendent, Units 2/3 J. Patterson, Assistant Maintenance Manager, Unit 1 R. Santosuosso, Assistant Maintenance Manager, Units 2/3
- R. Plappert, Compliance Manager
- 0. Brevig Licensing Supervisor
- R. Baker, Licensing Engineer San Diego Gas and Electric Company
- J. Winter, Site Representative
- Denotes those attending the exit meeting on July 28, 1989.
. The . inspectors also contacted other licensee employees during the course of the inspection, including operations shift superintenderits, control room supervisors, control room operators, QA and OC engineers, compli-ance engineers, maintenance craftsmen, and health physics engineers and technician . Plant Status Unit 1 The Unit was resterted on June 28, 1989, after completion of a 32-day outage to rept., auxilie y feedwater automatic actuation circuitry and
. perform modifications to' steam generator (S/G) wide range level instru-n ~ mentation. The Unit operated for five days until it was shut'dowr on b July 3, following a rapid power reduction due to high temperatures on
! '
the "A" reactor coolant pump motor upper thrust bearing. The problem L
.was determined to be oil lift system hose insulation and shim material from the motor thrust bearing housing which became dislodged and blocked
.
.
r
'
1 ,
. ,. 2 a - .
bs - -
..
the oil flow. .The problem was corrected and the Unit was restarted on
. July 20. The reactor operated for four days until-it was manually l -tripped on July 24 due to low water level in the."A" S/G. The cause of the low S/G level was a failure of plant personnel.to properly incorpo-
.. . rate' design change information from the.S/G level instrumentation
modification (performed during the outage previously mentioned) into station surveillance procedures. After completion of corrective actions..the reactor was returned to_ service on July 26 and operated for the balance of'the inspection perio '
Unit 2 The Unit operated at full > power throughout this inspection perio Unit 3
'On June 30, 1989, following 74 days of operation, the Unit was shut down due to a shaft seal leak on low pressure safety injection (LPSI) pump 3P-015. The seal was replaced and the unit was restarted on July 9, 1989, following satisfactory testing of the LPSI pump. The unit operated at full power for the remainder of this inspection perio . Operational Safety Verification (71707)
The. inspectors performed several plant tours and verified the opera-bility of selected emergency systems, reviewed the tag out log and verified proper return to service of affected components'. Particular attention was given to housekeeping, examination for potential fire hazards, fluid leaks, excessive vibration, and verification that mainte-nance requests had been initiated for equipment in need of maintenanc The inspectors also observed selected activities by licensee radio-logical protection and security personnel to confirm proper implemen-tation of and conformance with facility policies and procedures in these area Misalignment of an Atmospheric Dump Valve (ADV) (Unit 2)
At approximately 10:00 a.m. on July 20, 1989 the inspector found that the controller equalizing valve, S21301MU1264, for atmospheric dump valve (ADV) HV8419 had been left ope This rendered the ADV inoperable. Prior to this, the ADV had been removed from service for preventive ~ maintenance and placed in a 72-hour administrative action statement at 4:10 a.m. on July 18, 1989. It was returned to service at 12:53 p.m. on July 19, 1989. (The licensee was in the process of submitting a Technical Specification amendment request to specify a 72-hour action statement and appropriate surveillance requirements for the ADVs).
In response to the inspector's concern over the position of the equalizing valve, the licensee declared HV8419 inoperable and re-entered the 72-hour administrative action statement. After closing the equalizing valve, the ADV was stroked satisfactorily and returned to servic .g _ _ - - -- - ___- --,__- -
p1 # . .
N N
'
3' '
p, 4 &
o n
.g .
'
'As'a result-of this event, the licensee: initiated an Operations
!& '
Division investigation. This investigation found that duringL y ,
preventive maintenance to replace the bonnet drain; valve, the L, ' '
operator was asked to manually open the ADV to vent steam in order,
,
to' reduce.the. condensation in the drain pipe. 'To.do this. the y t, ," ', '
. operator opened the equalizing valve'in order to manually open.the
'
, ,
- ADV as specified in procedure, S023-3-2.18.1 PADV. Operation . The
'
.
~ operator then returned to the control room and documented on' Work
'
u
- "
v c Authorizatiori' Request (WAR) 2-8903610 that 2HV-8419 was. opene However, the' steps performed were.not documented,in detai '
- Procedure.S023-3-2.18 specified that local manual ADV. opening'
<
,
required. closing the; air. supply valve, opening,the actuato equalizing: valve, disengaging.the locking pin a'nd turning the-handwhee To close~ the 'ADV manually, these steps are reversed as:
specified in,the procedure. When'the. maintenance activity was-
'
completed on the follcwing day, the operator closed the ADV
. manually (by relying on memory) and returned' the 'ADV to service per theirestoration steps of the WAR.. However, the' equalizing valve was left open; The WAR did not specify' stroke- testing of the ADV '
fromlthe control room, since the licensee believed that the drain
~ valve repair would not affect the ADV operation. Thus, the action:
state. tent was_ exited at 12:53 p.m. on July 19, 1989, but theLADV was. inoperable.
"
-The inspector noted that this ADV was stroke tested on' July 14 L1989 (prior to the outage) and the next scheduled test was planned for August 14, 1989. The~ weekly alignment check specified in Attachment 5 of S023-3-2.18.1 did not specifically re W r checking of the equalizing valve. Thus, the potential existed fc. tLe ADV be inoperable from July 18, 1989 to August 14, 198 For corrective' action, the licensee briefed all operators on this event to reemphasize the importance of attention to detail when
, operating' safety related equipment. Furthermore, the licensee planned to revise the ADV operation procedure ~ to require stroke testing the ADV from the control room after each manual operation and to include:the equalizing valve position in the weekly alignment check.
.
The failure to follow the procedure for operation of ADV HV-8419 is an apparent violation (50-361/89-18-01).
One violation was noted in this area during the. inspectio e Evaluation of Plant Trips and Events (93702) Unit 3ShutdownDuetoJnoperableLPSIPump
. 0n June 30,1%, while performing a routine In-Service Test (IST)
on low press.fre safety injection (LPSI) pump 3P015, excessive shaft
. seal leakagt was observed. The leakage was measured at about 1400 cc/ min as co9 pared to the FSAR assumed cross leakage of 500 cc/mir:.
!
i p
i i l
- Y '
_ _ _ _ . ._______-__-___.l_A
. . _ - _ _ _ _ _
,
>
-
s .
-
,
,
, "
~
The licensee investigated the condition and found that the
"
mechanical seal was cracked from expansion of a swollen rubber seal i that was soaked by motor lube oil leakage. For corrective actio the licensee replaced the damaged seals and installed a deflecto '
to keep . lube oil away from the rubber seal. However, when post-maintenance testing of the pump was conducted on July 4,1989, the
- seal still leaked. Upon disassembly, debris was found on.the seal surface dnd it was believed that this may have caused excessive clearance. The licensee attributed this problem to improper seal installation and another seal was installed prior to returning the Unit-to service on July 8, 1989. The licensee continued to inves-tigate the root cause of the seal failure and planned to install oil deflectors on all of the LPSI pumps and the containment spray pumps for both Units 2 & The licensee discussed this event in Licensee Event Report (LER)
89-08, submitted on July 31,1989 (shortly after the end of the inspection period). The licensee's evaluation and corrective actions.will be reviewed further during followup on the LE Unit 1 Shutdown due to High RCP Motor Bearing Temperature On July.3, 1989, Unit I was shut down from 85% power due to a high temperature condition on the "A" reactor coolant pump (RCP) motor upper thrust bearing. The licensee's investigation determined that oil flow to the bearing was blocked by foreign material from oil lift system hose insulation and by shim material from the motor thrust bearing housing which had become dislodged. The licensee found that the insulation material, installed during the last outage without proper review, was incompatible with the lube oi Followup to this event was conducted as part of an NRC Maintenance Team Inspection (during this report period), since the cause was considered to be maintenance related. The results of that inspec-tion effort are documented in inspection report 50-206/89-16. This item is closed (50-206/89-18-01), Unit 1 Manual Trip Due to Loss of Feedwater On July 24, 1989, Unit I was manually tripped from 76% power when the level in "A" steam generator (S/G) dropped rapidly as a result of the closure of the "A" main feedwater regulating valve (MFRV).
At the time, technicians were performing a surveillance test on "A" S/G level instruments. The operators attempted to restore main feedwater to the S/G by coordinating efforts with the technicians performing the surveillance. However, when it was obvious that the S/G level could not be restored, the operators manually tripped the reactor as required by procedure. The licensee performed a root cause analysis of the event and determined that the cause was a failure to properly reflect design change package (DCP) information (for changes made to the S/G level instruments) into surveillance procedures. The review indicated that S/G high level contacts had l
'
_ . . _ _ _ _ _ ._._ _ ___
_
. _ -
,
.
'-
. .
, . 5
A been reconfigure-(as part of the DCP) to shut the MFRV solely on a high level in the S/G. Previously, another contact was also
- required to be closed in the circuitry to result in closure of the MFRV. A consequence of this event was that the S/G was essentially boiled dry before the reactor was trippe The licensee analyzed the condition and was able to take credit for a Westinghouse analysis performed in May 1982 on S/G integrity con-siderations following postulated steam and feedline breaks. That
, . analysis showed that the S/G could boil dry without any adverse effects to the S/G from thermal stresses. The inspector reviewed this analysis and'found that it bounded the July 24,1989 event _ at
. Unit 1. However, the, inspector considered.that additional licensee evaluations were necessary. In particular, the inspector had the following concerns:
-
The licensee's root cause nonconformace report (NCR) indi-cated that this event would not be counted as a thermal. cycle on the S/G since it can handle more than 20,000 initiations of auxiliary feedwater (AFW) (with less than 150 gpm flow) with the S/G in hot standby. The inspector pointed out to the licensee that this event was different from the scenario in the NCR since the S/G was dry. Thus, it may be necessary to count this as a thermal cycle.and consider the overall impact of this event and the potential impact of future similar events on the S/Gs (e.g., the effect of chemical plateout on S/G tubes as a result of boiling dry).
.
Operating instruction 501-2.3-5, TCN 1-3, " Abnormal Steam Generator Water Level,." states that if the S/G level in any one S/G is less than 10% narrow range with no sign of recovery, then trip the reactor and the turbine. This procedure may not provide the detailed guidance necessary for the operators to respond in this situation since it allows an unnecessary thermal cycle on the S/G caused by almost boiling
' i t d r If the operators had been able to restore main feedwater to the dry S/G, a large amount of relatively cold water (approxi-mately 400 degrees) would have been injected to the S/G. This
.could have adversely affected the S/G or caused a power excur-sion in the reacto The licensee acknowledged the inspector's concerns and indicated that they would perform additional evaluations of the effects of boiling a S/G dry, restoring main feedwater to a dry S/G at power, and the adequacy of the operating procedure. The licensee's evaluations will be reviewed by the inspector and are identified as inspector followup item (50-206/89-18-02).
No violations or deviations were identifie !
_ _ _ _ - _ _ _ _
g '
, .,
~
J f
~~
~1~ --- T hk.l .h ~~ '
r...
..
w- 6
-
.
7 - .-
,
,
^
,
15F Monthly' Surveillance A'ctikities (61726)
<!' lDuring.this< report period, the inspectors observed or conducted-
.
< inspection ofL11censee surveillance activities as discussed belo ' Observation of Routine Surveillance Activities '(Unit.1)
S01-12.3-10 Diesel Generator Load Test-(31 Day)-
.' S01-12.1-4 Control Room Shift and' Daily Log Readings
. ObservationofRoutineSurveillanceActivitics(Unit 2)
S023-3-3.25 -Once a Shift Surveillance (Modes 1-4)
~ Observation of Routine Surveillance Activities (Unit 3).
S023-II- Nuclear Instrument (NI) Safety Channel' O. Drawer Test (31. Day).
S023-V-3. LPSI In-Service Test (Quarterly)
A Failure to develop Test Program to Implement Final' Safety Analysis Basis (Unit 2&3)
OnJuly9,1989,whilereviewinglicenseeactivitiesrelated-to equipment. qualification deficiencies (to be addressed in a licensee event report), the inspector found that the licensee had not been performing ground checks on the instrument leads of the Reactor Protection System (RPS) and the Engineered Safety Feature Actuation System'(ESFAS) since initial operation of the units.1The licensee had indicated in the FSAR discussion of the single failure design
' basis that this would be performed periodicall In response to the inspector's concern,' the -licensee checked 46 -
Foxboro instrument -loops from the transmitter to the first device in: the cabinet and did not identify any grounds. Furthermore, the licensee began.to. define the total scope of all RPS and.ESFAS instruments and logic circuits to be' included in the ground check program..,The root cause of the failure to translate this element
"
of the design basis into plant surveillance p'rocedures is als being determine i
'
- This item remains unresolved pending additional informatio ,
(S0-361/89-18-02)
No violations or ' deviations were identified. -
s o
- 6; Monthly Maintenance Activities (62703)
.g '
During this_ report period, the inspectors observed or conducted inspection of licensee maintenance activities as diseassed belo ,2 h
T
_ _ . - . - _ . _ _ . - . _ _ _ _ _ . _ _ - _ - _ _ - - _ -
_ - _ . -
QQ W , -
f , ,,.
ff .
-
h[, , -
i
, , .
.
>
= ObservationofRoutineMaintenanceActivities(Unit 1)
&,
'
-
, ,
M089070276000 Reinstall Thermocouple on RCP-A Motor-
.g- M089070650000 Diesel Generator Inspection
'
M089072277000 ' Clean Ch cu_1ating Water Boxes
<
.M089069327000 Auxiliary Feedwater Pump Predictive Maintenance
<
b.- Observation of Routine Maintenance Activities (Unit 2)
M089071085000 Ground Check of Pressurizer Pressure Transmitter
..' 2PT01031 e Observation of Routine Maintenance ' Activities (Unit 3)-
M088082765000 - Appendix R Fuse Replacement for 3HV9371 Improper Temporary Cable Installation (Unit I'd On July 19, 1989,sthe inspector found four irstrument cables /that were' tied to the-vertical cable trays on both trains of the 4KV-Class IE cables. Located in these trays are the IE cables going to breaker' 11C02 and 12002 for Train 'A" and Train "B" safety related
-
- 4KV buses, respectivel The temporary cables were connected from a recorder to the Train
"B" 4KV Bus "2C" undervoltage auxiliary relays, 'This installation was performed under maintenance. order (MO) 85031690001 which was issued to. troubleshoot spurious diesel generator starts. The'
recorder cart was tied to a building post betind the Train "A" 4KV breaker cabinet in order to comply with the seismic tie-down requirement. However, the temporary instrument cables were routed
'across both trains of 4KV distribution by tying them to the cable trays. - Thus, the separation criteria specified in IEEE 384-1977-
>
were not met. This condition existed for more than 90 days, which
..
included Unit operation in Modes 1,2 and ,
In response to the inspector's concern, the licensee removed the-temporary installation and initiated an investigation. -The licensee noted in the review that temporary cable installation was
, controlled by procedure S0123-I-1.36 " Cables Installation of Temporary Cables", P.evision 0, dated May 28, 1986, applicable to
'
.
.all three units. This procedure stated that " Temporary applies to power support for the. duration of a maintenance activity unless
." '
designated as otherwise." In addition it specified.that " Temporary
,
cable shall not be routed across both trains of a redundant ~ safet system." Because the procedure did not specifically mention instrument and control cables, the planner dic! not believe that it applied to this MO, and no controls for routir,9 of the temporary cables were established. Thus, the technician routed the cables
,
Y _ - . . - - . _ - _ _ _ _ _ _ - - - - - . _ _ - - - - - - - -
,
,
,
, . . -
- \ ." 8
+
via the most direct path which resulted in crossing'both trains of the redundant 4KV system.
>
The inspector reviewed procedure 50123-I-1.36 and found that it was essentially based on a previous Bechtel construction procedure, WPP/1802, Revision 0, " Temporary Construction Aids and Utilities."
It did not include all the criteria specified by IEEE 384-1977,
" Criteria for Independence of. Class IE Equipment and Circuits."
However, the FSARs for Unit I and Units 2/3 commit the licensee to implement the requirements of IEEE 384-197 This finding is similar to a previous inspection finding (362/85-26-02). In that instance, the licensee did not properly justify separation criteria that were less conservative than thos' e established by the FSAR. In response to that. finding, the licensee provided the evaluation of separation criteria for. temporary construction power cables during plant outage / shutdown. However, temporary instrument cables and temporary power cables for plant operation above mode 5 were not specifically addresse In response to the inspector's finding, the licensee committed to conduct a comprehensive review of related procedures for necessary update and to review personnel training for implementation of the appropriate separation standard The licensee's failure to develop and implement an adequate maintenance procedure.for safety.related ec;uipment is an apparent violation (50-206/89-18-03).
One violation was identifie . Engineered Safety Feature Walkdown (71710) (Unit 1)
The inspector walked down the Auxiliary Feedwater (AFW) piping system after the July 24, 1989 Unit 1 trip and AFW initiatio No~ violations or deviations were identifie . Decay Heat Removal with Reduced Reactor Coolant Inventory (TI 2515/101)
Review of the licensee's program for control of reactor vessel level and decay heat removal activities during periods of reduced reactor coolant inventory (mid-loop operations) were previousl,y examined, as documented in Inspection Reports 50-206/88-28 (for Unit 1) and 50-361/89-14 (for Units 2 and 3). NRR r views of the licensee's program were documented in letters to the licensee dated July 12 and August 2,1989 for Unit 1 and Units 2/3, respectively. Licensee actions in response to concerns '
and observations included in the NRR letters will be examined during future inspections. (Followup Item 50-206/89-18-04)
No violations or deviations were identifie _ _ _ _ _ _ _ _ _ .
? ,l~ I9
.
9. Evaleetion Of Licensee Ouality Assurance Program Implementation (35502)
Onsite Review Committee And Nuclear Safety Group Review (40500-1)
Management Safety Involvement (35502-1)
The purpose of this review was to continue with an evaluation of the effectiveness of the licensee's implementation of the quality assurance program. This effort consisted of a comprehensive review of program surveillance, audits, periodic finding reports, and other documents concerning quality oversight group activities. In addition, discussions with Southern California Edison's (SCE's) quality oversight groups were
.
held in order to determine whether recent changes in personnel and pro-gram direction were enhancing the effectiveness of the quality programs to find problems-and weaknesses in SCE activities. During this inspec-tion, the site OA, Quality Programs Section, Nuclear Safety Group (NSG)..
and Independent Safety Engineering Group (ISEG) were specifically
.
reviewed. This effort consisted of a followup to activities documented in inspection report 50-206/89-1 _
The 'nspector' discussed the functions of the Quality Programs, NSG, and ISEG groups with responsible licensee personnel to understand their areas of' responsibility. Their functions are as follows:
-
The Quality Programs Section consists of a quality engineering (OE)
group, program support group, and a training and certification group.' The OE group's functions are to perform design process monitoring of design documents such as design change packages, per-form surveillance of corporate organizations, and perform internal audits. The program support group functions to oversee procedure and document changes, perform 10 CFR Part 21 evaluations, and provide other support functions. The training and certification group is chartered with training program development for 0A and administration of certification program The NSG is chartered with performing independent reviews of such items as proposed facility changes, license amendments. LERs, etc.;
surveillance of plant activities; equipment deficiency trending; probabilistic risk assessment (PRA) trending; and area monitoring program participatio The ISEG is mainly tasked with performing operating experience reviews of NRC, INP0, and vendor issuances. In addition, ISEG '
performs a number of other activities such as surveillance of plant activities, review and approval of " accept-as-is" NCRs, and "
area monitorin As a result of these discussions and previous inspection effort (as mentioned above) the inspector considered that the licensee had clearly defined the responsibilities of these groups, provided feedback mechanisms, and established interface points between the group j The inspector discussed the training and certification programs for the 4 quality assurance engineers. During these discussions, the licensee !
indicated that all OA engineers had been tested and that as a result, a 1 i
_ _ _ _ - _ _ _ - _ - -
- - _ _ _ _
_ _ _ _ -
[${:*. Y
\ .
,
y l1 , . number of engineer certifications' had to be downgraded. This indicated .
' the need for additional training in certain areas. The trainingLprogram as detailed by training program description'(TPD) 0AE-001 consisted of, basic. technical specific, and personnel development training. The1 3 training appeared to be based upon established American National-Standar# Institee (ANSI). standards and, Institute for Nuclear Power-Operations (INPO) document i
. Audits Reviewed-
.SCES-001-89 to evaluate the Unit.1 ISI program
, SCES-003-89 to evaluate: activities affecting design change iimpiementatio .SCES-005-89 to check safety injection tank operability
.SCES-006-89Lto verify compliance with Technical Specifications'
SCES-018-89 to verify compliance with Technical Specifications These, audits. appeared ido be well planned and documented, 'and SCES-001
' had a 'significant finding concerning implementation of the ISI-progra Summary
, Based upon the documents reviewed and discussions: held'with SCE person -
nel ~the.. inspector considered that the licensee had laid the. proper groundwork.for an effective quality oversight organization and!that the oversight function was'still headed in a positive direction as indicated
~
'
-in inspection report 50-206/89-11. The inspector will continue to review the licensee's implementation of program upgrades to enhance
'
l quality oversight group effectiveness during future inspection ,
No violations or deviations'were identifie . Nonlicensed Staff Training (41400)
Technical Training /for Engineering Staff 1 ~ Design Engineer Training
,
The;. licensee is participating in industry efforts to assess the scope and content of training that is desirable for the design ,
engineering staff...Conjunctively, the licensee had developed a
'
' draft Nuclear Engineering Design Organization Training Plan, which was in the review and comment stage at the time of the inspectio The. inspector reviewed this. plan and discussed certain aspects with licensee management. The' plan, as envisioned, was composed of four major elements: orientation, job fundamentals. training, continuing ,
training, and on-the-job training. The licensee was in-the early phase of defining the content of each element.-
The licensee was using the INPO Guidelines for the development of their comprehensive training program. In addition, the licensee's independent task force arrived at several conclusions and
- - - -______ _ - b
~
-
<
[ ,. 11
. -
[
(. 's i
recommendations as a result of a training assessment. These recommendations were being factored into the training program development. The licensee also considered the comments regarding training provided as a result of the NRC's Safety System Functional Inspectio t The inspector concluded that the licensee's efforts to produce a design engineering training program appeared to be proceedin However, progress in implementing the design engineering program appears to have been inappropriately slow.
l- Technical Training for Site Technical Staff and Managers The. inspector reviewed the technical training program for technical staff and managers, as provided by Training Program Description TST-01. The program consists of two phase Phase #1 is general training designed to provide general information for all disciplines regarding NRC regulations, industrial stan-dards, corporate and site policies, radiological health and safety, and site procedures and requirements. Phase 1 is generally ,
completed within one year of entry into the progra I
' Phase 2 is designed to provide site specific training in plant systems, core damage mitigation, fundamentals, plant transient analysis, and simulator systems integration. Phase 2 training is generally completed within five years of entry into the progra Phase 1 and 2 training will be provided by self-study, classroom training, computer training, and/or videotaped trainin Cognizant engineer training is a lengthy process recognizing that ;
there is a high level of experience in the Station Technical Orga-nization. Therefore, the cognizant engineer training is designed to focus on continuing training provided by self-study, specialized discipline technical schools, etc. New hires (about two per year)
participate in orientation type training and work as a discipjine engineer for about two years. Following this, new hires attend approximately 1.5 years of shift technical advisor (STA) training and work as a licensed STA for approximately four years before returning to station technical as a cognizant enginee !
No violations or deviations were identifie . Review of Licensee Event Reports (90712, 92700)
Through direct observations, discussion with licensee personnel, or i review of the records, the following Li e nsee Event Report (LER) was closed:
Unit 1 j 206/88-18, Improper Control of Chemistry Monitoring Activity
I
_____________________________i
,-
'
x m gg ,
- -
--- - - --
M ; ..
- *
. .je: : A
. , . ;12 y ,
x .
ge
'
?The inspectors. reviewed.the..following LER and found that the licenseel W r still had corrective; actions to perform for' the problemssidentified in-
- , the LER.L Thertfore, this LER willc remain open pending.. completion ofE t 111censee corrective actions and additional NRC review:
0 '
' Unit 11 f ,
. 88-04-01, Excess Opening Time for. Safety Injection Valve HV851B;
.,
' ~
-
No violations or deviations were identified, t x ,
312. . Follow-Up on Items of No' incompliance L(92702). 3:
'm '(Closed) Violation 206/88-28-01, Improper Contrdi of Reactivity 9~ ._
' Monitoring Activity t
. - ,
,This item concerned the performance of reactor * coolant chemistry , f, , '
sampling without using the' latest procedure for performance of the
, ,
,
' activity.-~'* x
.
's '
- For corrective action, the licensee removed superseded procedure
<c
'
,: .50123-III-1.6.1, Revision 2, from the sample room and replaced-it y' ' with Revision 3. . In addition, the copies of the uncontrolled
procedure were removed from other locations and replaced with the
'
updatedgprocedure. < Administrative controls were instituted to;
'
i ~
t ensure that revisions and temporary changes were identified and -
'
provided to chemistry personnel. The. inspector considered that the
licensee's actions on'this matter were appropriate and that this-
item is closed.- (Closed) Violation 206/89-01-01, Inadequa'te Impleinentation of Foreign Material Exclusion Controls ,
,
This item identified that. foreign' material" exclusion (FME)
boundaries and postings'~were not established for. maintenance on an:
, emergency diesel generator (D/G).
For corrective' action on this matter, the. licensee reestablished-the FME boundaries for the D/G maintenance work, and FME monitor training was. enhanced to reaffirm requirements for challenging personnel entering a FME boundary. The inspector considered that the : licensee's actions were adequate. This item is close ' (0 pen) Violation ~ 362/85-26-02, Improper Routing'of-Temporary Cables:
.This item identified. problems with existing procedures for controlling the use'and placement of, and separation criteria for, temporary cables routed near Class-1E equipment and cable tray ,
, .This is similar to, and will be reviewed with, the concerns y discussed in paragraph 6.d of this repor l/
<
- _ - . .
'
['- ., [ .13
, o-13. Exit Meeting (30703)
~On' July 28, 1989 an exit meeting was conducted with the' licensee representatives identified in Paragraph 1. -The inspectors summarized-the inspection scope and findings as described in previous sections of this repor The. licensee acknowledged the inspection findings and noted that appropriate corrective actions would be implemented where warrante The licensee did not. identify as proprietary any of the.information provided to or reviewed'by the inspectors during this inspection.,
-
$
6