IR 05000315/1989023
| ML17328A162 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 09/13/1989 |
| From: | Burgess B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17328A160 | List: |
| References | |
| 50-315-89-23, 50-316-89-23, GL-87-12, IEB-87-002, IEB-87-2, NUDOCS 8909270167 | |
| Download: ML17328A162 (25) | |
Text
U. S.
NUCLEAR REGULATORY COMMISSION
REGION III
Report Nos. 50-315/89023(DRP);
50-316/89023(DRP)
Docket Nos. 50-315 50-316 License Nos.
American Electric Power Service Corporation Indiana Michigan Power Company 1 Riverside Plaza Columbus, OH 43216 Facility Name:
Donald C. Cook Nuclear Power Plant, Units 1 and
Inspection At:
Donald C. Cook Site, Bridgman, MI Inspection Conducted:
July 19 through August 29, 1989 Inspectors:
B. L. Jorgensen D. G. Passehl Approved Byg
.
.
B ge s, Chief
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g Projec Section 2A Ins ection Summar Da Ins ection on Jul 19 throu h Au ust 29, 1989 (Re ort Nos.
50-315/89023(DRP);
reas Ins ecte
Routine unannounced inspection by the resident inspectors o : actions on previously identified items; plant operations; reactor trips; maintenance; surveillance; emergency preparedness; reportable events; Bulletins; and, NRC Region III requests.
One Safety Issues Management System (SIMS) item (No. BL-87-02, NRC Bulletin 87-02)
was reviewed and closed.
Results:
Of the nine areas inspected, no violations or deviations were
~sent17ied in eight areas.
One violation was identified (Level IV - failure to follow administrative procedures for shift and relief turnoyer - Paragraph 10.c) in the remaining area.
Both units operated routinely du'ring this inspection period except for a four day period (August 14 - 18, 1989)
when Unit 2 was shutdown following a reactor trip caused by an electrical component failure.
No specific strengths or weaknesses were noted during this inspection.
pbR ADOCK 05000
DETAILS 1.
Persons Contacted
- W. Smith, Jr., Plant Manager,
~A. Blind, Assistant Plant Manager - Administration
- J. Rutkowski, Assistant Plant Manager - Production L. Gibson, Assistant Plant Manager - Technical Support B. Svensson, Licensing Activity Coordinator
- K. Baker, Operations Superintendent
- J. Sampson, Safety and Assessment Superintendent E. Morse, gC/NDE Genera'1 Supervisor T. Beilman, 18C Department Superintendent
- J. Droste, Maintenance Superintendent
- T. Postlewait, Technical Superintendent
- Engineering
- L. Matthias, Administrative Superintendent
- J. Wojcik, Technical Superintendent
- Physical Sciences M. Horvath, guality Assurance Supervisor D. Loope, Radiation Protection Supervisor The inspector also contacted a number of other licensee and contract employees and informally interviewed operations, maintenance, and technical personnel.
- Denotes some of the personnel attending Management Interview on August 30, 1989.
2.
Actions on Previousl Identified Items (92701 92702)
a.
(Closed)
Generic Letter 87-12 (315/87012-GL; 316/87012-GL):
Loss of Residual Heat Removal (RHR) While the Reactor Coolant System (RCS)
is Partially Filled.
The subject matter has been superseded by an NRC 10CFR50.54(f) letter, Generic Letter 88-17.
Inspection of the more recent, broader based letter will be performed pursuant to published Temporary Instruction TI 2515/101.
b.
(Closed)
Open Item (316/87026-02):
undocumented isolation of operating air to valve 2-NRV-164.
The Problem Report 87-960 package associated with this item summarizes the investigation and corrective actions.
The investigation found no documented activity (Job Order history, testing, operations)
to which the condition could be ascribed.
It is presumed the air valve was isolated by mistake, due to either a component identification error or a
restoration error.
To prevent a recurrence, startup procedures were changed to perform functional and full-stroke valve testing following a refueling outage.
This is now accomplished at Step 6.18 of Unit I procedure
- 1-OHP 4021.002.001,
"Filling and Venting the Reactor Coolant System,"
and at Step 6.19 of the analogous Unit 2 procedur c ~
d.
e.
(Closed)
Open Item (316/87026-03):
weld repair apparently not done in accordance with administrative requirements.
This item was subsequently upgraded to a violation for which a Notice of Violation (NOV) was issued with Inspection Report No. 50-315/87029(DRP);
316/87029(DRP).
Corrective and preventive actions were deemed adequate at the time the NOV was issued, such that a written response to address such actions was not required.
II (Closed)
Open Items (315/87026-04; 316/87026-06):
control of
"Caution Tags" in main control rooms.
The licensee performed a
followup audit of all "Caution Tags" posted throughout the plant, and reduced the number of effective tags from 155 to 46.
Subsequently, a revision to procedure PMI-2110, "Clearance Permit System" incorporated more concise guidance on the use of tags.
Also, a bi-monthly review was established to re-examine all effective "Caution Tags" to ensure they are still valid and current.
These actions have served to effectively control proliferation of such tags.
(Closed)
Open Items (315/87026-05; 316/87026-07):
mechanism for controlled distribution of design change information to end users.
This matter was initially noted during a site guality Assurance (gA)
audit, No. 87-27.
The inspector reviewed closeout actions for the audit finding.
This involved deletion of requirements to prepare Design Change Guidelines from the governing procedure.
These Guidelines had served to inform/train operators on design changes, but they were not "controlled" documents of verified accuracy.
Closeout of the gA finding did not include developing an acceptable substitute for the Guidelines document.
The licensee subsequently developed such a "controlled" document (covered by procedure, verified, updated when required) entitled Design Change Overview.
This document satisfied the inspector's concern.
No violations, deviations, unresolved or open items were identified.
3.
0 erational Safet Verification (71707 71710 42700)
Routine facility operating activities were observed as conducted in the plant and from the main control rooms.
Plant startup, steady power operation, plant shutdown, and system(s)
lineup and operation were observed as applicable.
The performance of licensed Reactor Operators and Senior Reactor Operators, of Shift Technical Advisors, and of auxiliary equipment operators was observed and evaluated including procedure use and adherence, records and logs, communications, shift/duty turnover, and the degree of professionalism of control room activities.
Evaluation, corrective action, and response for off-normal conditions or events, if any, were examined.
This included compliance to any reporting requirement Observations of the control room monitors, indicators, and recorders were made to verify the operability of emergency systems, radiation monitoring systems and nuclear reactor protection systems, as applicable.
Reviews of surveillance, equipment condition, and tagout logs were conducted.
Proper return to service of selected components was verified.
a.
Unit 1 was in routine power operation during this inspection period.
Operation of the unit at reduced primary coolant temperature and pressure but at full rated thermal power did not cause any significant difficulties.
Unit 2 was in routine power operation except for the period August 14-18, 1989 following a reactor trip due to instrument electrical problems; see Paragraph 4.
b.
The inspector performed a partial walkdown of the Unit 1 auxiliary feedwater system using Data/signoff Sheet 5.1 from procedure 1-OHP 4021.056.001,
"Filling and Venting Auxiliary Feedwater System and Placing System in Standby Readiness."
Approximately 100 components were inspected as to proper identification and position.
(1)
Valve 1-CPX-253-V1, designated to be closed, was found open.
This valve is the inboard isolation valve on a small suction pressure test tap.
The licensee was informed of the discrepancy and initiated a Problem Report (No.89-918) to document investigation of the cause and significance of the item.
The investigation found the valve had probably been opened for routine monthly testing and not reclosed.
The test procedure'ontained no specific reference to the valve (by number), directing test personnel only to isolate the line before disconnecting the test pressure gauge at the end of the test.
This step had apparently been accomplished by closing the outboard valve, which is not included on the lineup sheet.
As to significance, the evaluation found other similar valves also positioned "open", so the checksheet was revised to have them all "closed" and the issue was evaluated generically.
(2)
Valve 1-FW-174, the emergency leakoff line isolation from the West motor-driven auxiliary feedwater pump (MDAFP), is physically located in the East pump room, adjacent to valve 1-FW-175, the East pump isolation valve.
These adjacent valves are listed on pages 9 and 2 of the lineup sheet, respectively.
This is an extreme example among several of the lineup sheets not logically following component location, thus being unnecessarily inconvenient.
(3)
High-point vents were not treated consistently.
Some were included on the checksheet, but others were not.
Vents not included were:
1-FW-187, I-FW-212, 1-FW-189, 1-FW-190, 1-FW-210, and I-FW-211.
The above listing may be incomplete for Unit I.
Unit 2 was not checked.
The licensee was requested to review the treatment of vents.
c.
Some items were identified during tours for which the inspector requested licensee review and appropriate followup.
(I)
The post-outage auxiliary building cleanup had begun, but several staging areas (for various tools or equipment) of recent vintage were not designated for cleanup.
Instead, they were approved as
"permanent" laydown areas.
This was discussed with plant management from the perspective of verifying that new permanent laydown areas are really necessary and, if so, they are in appropriate, optimum locations.
(2)
A previous inspection report (Report No. 50-315/89018(DRP)
50-316/89018(DRP))
noted the procedure present at the ice-making machine monitoring and control console was not
"controlled" and lacked some pages while other pages were disordered.
The inspector reviewed the procedure again during a tour of the area during this inspection.
He found a
controlled procedure copy, with all pages present, ordered, and protected in plastic sleeves, posted in a new binder at the console.
d.
e.
On July 20, 1989, a routine sample disclosed the Unit I Boron Injection Tank (BIT) had a boric acid concentration of 19,596 ppm vs.
a Technical Specification minimum of 20,000 ppm.
The inspector was informed of the commencement of unit shutdown as required by Technical Specifications (See Paragraph 7.b.)
and he observed licensee corrective actions from the main control room.
The BIT recirculation was switched to an alternate boric acid storage tank and acceptable concentration restored within less than two hours of the original finding.
The shutdown was then terminated and the unit restored to full power., Problem Report 89-891 was initiated to document cause and corrective actions for this event.
The licensee's review, which the Plant Manager directed to include generic sampling and analytical implications, is continuing.
On August 7, 1989, switchyard breaker 12AB tripped open.
This breaker provides one source of offsite power to each unit.
An inspection showed a destructive failure of one insulator phase, and damage to another from flying fragments.
Technical Specifications directing breaker alignment checks and a start test on the AB diesel generator were met.
The cause of the failure was concluded to be moisture intrusion leading to a hot short.
Both insulators were replaced and offsite power fully restored within the time limits of the Technical Specifications.
No violations, deviations, unresolved or open items were identifie '.
Reactor Tri s(s) or ESF Actuations (93702)
Unit 2 tripped from 100 percent power on August 14, 1989, due to a failure of the Control Room Instrumentation Distribution (GRID) IV Inverter Static Switch.
The trip occurred when operators were attempting to restore the normal power supply to GRID IV, which had earlier transferred automatically to its normal alternate power source.
When normal power restoration was attempted, the failed static switch caused CRID IV to provide a low output voltage (high current) to its loads.
Numerous fuses and breakers were blown and tripped, which resulted in the loss of power to many control room instrument indications.
On August 15, an NRC Augmented Inspection Team (AIT) was established to review the event, determine cause, assess plant response, and monitor equipment repair and testing.
The results of. the'nvestigation can be found in NRC Inspection Report No. 50-315/89025(DRP);50-316/89025(DRP).
Unit startup was performed on Friday, August 18 after repairs and was uneventful.
No violations, deviations, unresolved or open items were identified.
Maintenance (62703 42700)
Maintenance activities in the plant were routinely inspected, including both corrective maintenance (repairs)
and preventive maintenance.
Mechanical, electrical, and instrument and control group maintenance activities were included as available.
The focus of the inspection was to assure the maintenance activities reviewed were conducted in accordance with approved procedures, regulatory guides and industry codes or standards and in conformance with Technical Specifications.
The following items were considered during this review: the Limiting Conditions for Operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures; and post maintenance testing was performed as applicable.
The following activities were inspected:
a0 Job Order JO AOl0986: support job for design change RFC-12-2901 involving replacing the range capsule in instrument 1-CFA-459 (Reactor Coolant Pump Seal Heat Exchanger Component Cooling Water Flow).
This change increased instrument range to a maximum 120 gpm vs. the'original 50 gpm.
Actual flows were typically up to 75 gpm, over-ranging the original instrument.
b.
Plant Modification PM-698: replace non-nuclear auxiliary supply cooler on ventilation unit 1-HV-AS-2.
c In response to continuing generic concerns from NRC Region III and the Office of Nuclear Regulation (NRR) regarding check valve reliability and operability, particularly in those systems where check valves are safety-related or important to safety, the inspector reviewed select programmatic aspects of the licensee's check valve maintenance and surveillance progra To summarize, the licensee has incorporated the Significant Operating Events Report (SOER) 86-03 into a procedure (12 THP 5070 ISI.003) for the disassembly and visual examination of check valves.
There is an additional procedure (12 THP 5070 ISI.002) which implements the in-service examination and testing requirements of ASME Section XI into the valve program.
The inspector provided a list of specific check valve problems and their associated corrective actions to Region III.
Included were those which resulted in an NRC Information Notice (IN 88-85)
on Anchor Darling check valves, and those which failed their containment leakage rate testing.
Other Problem Reports reviewed and submitted dealt with: material differences found for two bolts in 2-RK-108 East and West, which resulted in replacement of the bracket and bolts called for by the vendor's bill of material; and the problems with seat leakage on I-SI-150 and 2-SI-150, where the nut holding the disc on the clapper arm was found missing and the disc lying on the bottom of the valve.
The last was resolved by staking the nut in three places to preclude loosening.
The inspector noted that once having been addressed by the licensee's program, there have been no repeat problems with specific check valves.
d.
Job Order JO 8013045: troubleshoot lack of turbine-driven auxiliary feedwater (TDAFP) pump speed indication observed after Unit 2 reactor trip of August 14, 1989.
The pump tachometer magnetic pickup had a broken lead, which.was resoldered.
The failure eliminated'ot only the speed indication, but the electronic overspeed protection; mechanical overspeed protection was unaffected.
The electronic overspeed was known to be functioning properly as recently as July 26, 1989.
The time of the failure could not be precisely determined.
Pump performance in response to demand signals subsequent to the reactor trip was not adversely affected.
e.
A routine licensee evaluation of data in his corrective action program disclosed eight instances of wrong fuses or breakers being installed.
Four events were identified in a four-month period in 1988.
As a consequence, a potential
"adverse trend" Problem Report (No.89-907)
was wr itten to investigate why improperly sized fuses or breakers came to be installed.
No violations, deviations, unresolved or open items were identified.
6.
Sur veil lance (61726 42700)
The inspector reviewed Technical Specifications required surveillance testing as described below and verified that testing was performed in accordance with adequate procedures, that test instrumentation was calibrated, that Limiting Conditions for Operation were met, that removal and restoration of the affected components were properly accomplished, that test results conformed with Technical Specifications and procedure
requirements and were reviewed by personnel other than the individual directing the test, and that deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel.
The following activities were inspected:
a ~
b.
C.
- 12 THP 6040 PER.001,
"Centrifugal Pump Performance Test Procedure."
This test was performed on the Unit 1 East Motor-Driven Auxiliary Feedwater Pump (MDAFP) on August 2, 1989, to verify previous test results.
The earlier results, which provided a
"baseline" pump performance curve for a newly installed impeller, indicated the new impeller was little better than the worn component it replaced.
Further, the East pump produced lower discharge pressures across the range of flows than the West NDAFP, with which it shares certain emergency leak off piping.
The test of August
confirmed these results.
The licensee has determined that emergency and safety functions of the pumps are not compromised by the discharge pressure differences (up to about 40 psig) but evaluation of potential problems for non-safety uses of the pumps is continuing.
- I OHP 4030 STP.017E,
"East Motor Driven Auxiliary Feedwater System Test."
When this routine monthly test was performed on August 20, 1989, the pump failed to produce 1375 psid while at 200 gpm flow through the recirculation test line.
The failure was documented by Condition Report 12-08-89-1455.
Investigation of the situation established the procedure criterion involving the 1375 psid as unnecessarily conservative, so it was revised.
The value of 1375 should apply to discharge gauge pressure while the pump is aligned through its emergency leakoff (mini-flow) line.
The fact that this was a first failure to meet this conservative criterion, since its incorporation in a June 1988 procedure revision, was considered a
result of slightly elevated condensate storage tank temperature.
It was not clear this was the case.
This was discussed with the licensee.
The inspector's concern that
=
the Technical Data Book Figure 15.2 used to evaluate pump performance was not current to the new impeller did not turn out to be the case.
A new pump curve had been established for the new impeller per IST requirements.
The procedure revision was necessary to end confusion over testing using IST values and testing using Techni ca 1 Specification values.
- 12 THP 4030 STP.208-BI,
"ECCS Flow Balance - Boron Injection System."
While performing the subject test on Unit 1 West Centrifugal Charging Pump (CCP), it tripped on motor overload.
The inspector reviewed the resultant Problem Report (PR 89-704)
and it'
supporting close-out documentation.
The root cause of the event was attributed to setpoint drift.
The overload re'lay was found to actuate at 3.05 amps instead of the correct value of 3.5 amps.
It was last checked in February l989 and
required no adjustment.
The relay was subsequently calibrated to, its correct setpoint per I2 THP 6030 IMP.014, "Protective Relay Calibration."
d.
To satisfy generic concerns, the CCP overload setpoints as presently established were reviewed for adequacy.
The review determined that the setpoint calculational methodology did not appear to present a
generic problem based partly on the operating history of critical pump motors, where there have been few conditions of motor trips due to over load, and since test flows and accident flows are essentially equivalent for the various pump motors.
- 12 MHP 4030 STP.029,
"Test of Grinnell Hydraulic Snubbers."
When snubber No.
1-GMS-S377 was observed during a routine inspection to have a low hydraulic fluid reservoir level, it was removed and a
spare installed.
It subsequently failed a "bleed rate" performance criterion during functional testing using the subject procedure, and a Job Order was written for a rebuild of the snubber.
Further, the failure was documented on Condition Report 1-07-89-1319 for purposes of tracking and adjusting inspection schedules relating to the fai lure.
Coincidentally, the inspection noted a separate Condition Report (No. 12-07-89-1304)
requesting a Technical Specification interpretation that snubbers found with low fluid level, on any occasion other than a scheduled visual surveillance to verify fluid level, would not be counted as affecting subsequent i'nspection schedules.
This interpretation was at variance with previous practice, and the inspector considered it to be nonconservative.
The frequency of snubber failures versus time cannot be accurately determined unless all "failures" are counted, including those which are randomly identified or which are self-identifying.
This was discussed with licensee management during the course of the inspection and at the Management Interview.
The licensee continues to apply a conservative approach while the proposed interpretation is under consideration.
The inspector will continue to monitor the issue until it is resolved.
e.
- 12 MHP 4030 STP.008,
"Inspection of Containment Sump."
The inspector reviewed the procedure and the results of the most recent inspections of the respective unit sumps.
The Unit 2 sump inspection to satisfy Technical Specification 4.5.2.d.2 (requiring inspection each 18-months)
was last performed on February 14, 1989.
Minor debris was found on the sump f'floor and was removed.
Drains, trash racks and screens were found to be clean.
The Unit I sump inspection for the same purpose and requirements was last performed on May 26, 1989.
Again, minor debris was removed and other conditions were found acceptable.
This information was reviewed to verify the licensee had measures in effect to keep his safety related sumps clean (debris free)
and with undamaged drains, racks and screens; another plant recently discovered debris in a safety related sump which was suspected of having been there, undetected, for over a decade since plant construction.
No violations, deviations, unresolved or open items were identifie Emer enc Pre aredness (82201 82203)
a ~
The licensee declared an Unusual Event on August 8, 1989 at 7:59 p.m.
(EDT) under their Emergency Plan as a result of being notified by the local FBI of a bomb threat that was received in Saginaw, Michigan, against an unspecified Michigan nuclear plant.
A suspect who allegedly admitted making the threat was later apprehended by police in Grand Rapids,'Michigan.
The Unusual Event was terminated at 8:43 p.m.
(EDT) the same day.
b.
The licensee declared an Emergency Plan Unusual Event on July 20, 1989 with regards to the Unit 1 Boron Injection Tank (BIT) discussed earlier (See Paragraph 3.d).
The inspectors were satisfied with the classification and notification of both events.
No violations, deviations, unresolved or open items were identified.
8.
Re ortable Events (92700)
The inspector reviewed the following Licensee Event Reports (LERs) by means of direct observation, discussions with licensee personnel, and review of records.
The review addressed compliance to reporting requirements and, as applicable, that immediate corrective action and appropriate action to prevent recurrence had been accomplished.
a.
(Closed)
Licensee Event Report (LER 316/86028:
Personnel error results in operation with incorrect rod insertion limit. Electrical jumpers required for normal operation of insertion limit computing modules were not re-connected properly on two of four modules after a calibration.
The modules were thus incorrect'ly dependent (vs.
independent)
on reactor coolant temperature.
The Control Bank 0 insertion limit ca'lculation was nonconservative'ly Iow by about 20 steps, but subsequent review indicated no rods violated what should have been the insertion limits.
The problem arose from a misunderstanding of the purpose and use of the jumpers during a
revision of the calibration procedure, such that their reconnection was not addressed.
The procedure was corrected, and the problem has not recurred.
b.
(Closed)
Licensee Event Report (LER 316/87009):
Radiation monitor inoperable without required compensatory sampling due to miscommunication (personnel error).
Monitor SRA-2800, the gland seal leakoff (GLSO) condenser effluent monitor used to monitor primary-to-secondary leakage (redundant monitor to condenser off-gas) "failed" when cooling to the GLSO isolated spuriously.
Thereafter, the monitor's inoperable status (from inadequate flow due to sample line excess condensation)
was not recognized in time
~
to perform an 8-hour compensatory grab sampling.
This despite a
monitor check which should have noted the problem but, due to miscommunication (personnel error) did not.
The problem was identified and corrected with only one compensatory sample not
C.
collected on time.
There were previous compensatory surveillances likewise omitted due to miscommunications, but the problem has not recurred subsequent to this event.
(Closed)
Licensee Event Report (LER 316/87014):
Missed surveillance due to personnel error in.process computer software programming.
The quadrant power tilt ratio annunciator program was inadvertently written over during troubleshooting, such that the alarm was inoperable but the inoperability was not recognized.
Compensatory surveillances for the inoperable function were not performed as required in the 8-day interval until the problem was recognized.
The tilt ratio continued to be calculated, however, and no limits were exceeded.
The root cause was programmer failure to verify there was adequate computer space available for his program installation.
All computer analysts authorized to perform troubleshooting were counseled concerning the event; there has been no recurrence.
No violations, deviations, unresolved or open items were identified.
9.
NRC Com liance Bulletins (92703 25027)
The inspector reviewed the NRC communications listed below and verified that: the licensee has received the correspondence; the correspondence was reviewed by appropriate management representatives; a written response was submitted if r'equired;.and, plant-specific actions were taken as described in the licensee's response.
(Closed)
NRC Bulletin No. 87-02, including Supplement No. I and Supplement No. 2, "Fastener Testing to Determine Conformance With Applicable Material Specifications."
Followup inspection was performed in accordance with NRC Temporary Instruction 2500/27, which initially identified Sections 04.05 and 04.06 as applicable to D. C. Cook.
Inspection disclosed that Section 04.05 did not actually apply.
This Section involved plants which had not tested a minimum of 20 safety-related and a
total of 40 fastener pieces; D. C. Cook was recorded as having tested only 18 safety related pieces among a total of 51.
The licensee's Bulletin submittals (AEP:NRC:1045 and 1045A dated January 12 and January 27, 1988 respectively)
document 22 safety related pieces analyzed.
The discrepancy involved two all-thread studs, each of which came with two hex nuts.
The six pieces were accounted for on only two individual
"Fastener Testing Data Sheets" incorporated as Attachment 1 to the referenced letters.
The first 18 pages of Attachment I thus account for 22 pieces, not 18.
All 22 pieces were individually analyzed, with results reported in Attachment 2 of the referenced letters.
Section 04.06 applied to D.
C. Cook.
This Section involved verification of adequate licensee root cause analysis and corrective action for out-of-specification non-safety related fasteners.
One such item (sample No. 31-772250, an ASTM A193 Grade B7 I/2 by 2 inch stud)
was found.
This
stud was actually carbon steel, not Grade B7 alloy.
The licensee's root cause evaluation found the discrepancy had not occurred at purchase..
Instead, several studs bought and issued as ordinary carbon steel, but not required to complete the job, were erroneously returned to stock with an alloy part number assigned.
The erroneously labeled pieces were all accounted for; none had been used in safety related applications.
Preventive action to preclude recurrence involved implementing a
verification for non-safety related material (as already existed for safety related items) to assure a match between "return" coding and the material ordered.
No violations, deviations, unresolved or open items were identified.
10.
Re ion III Re uests (92701)
The inspector did a follow up review concerning observations by NRC Region III Operator Licensing Examiners.
The results are summarized below.
a.
Procedure 12-OHP 4021.013.006:
"Operation of the Eberline Radiation Monitoring System Control Terminal," describes the R-5 radiation monitor as being located on the refueling crane bridge over the Spent Fuel Pit. It is actually located on the North wall adjacent to the Spent Fuel Pit.
b.
The Unit 2 Temporary Modification Log has an entry (No. 95) dated November 2, 1987 describing an.action where leads were lifted to defeat a standing problem alarm on the four bottle nitrogen system for the accumulators.
The examiner's concern was that the configuration was designated as a Temporary Modification longer than the 12 months the procedure mandates (ref.
The inspector reviewed the referenced procedure and found no problems with the licensee's log entry.
According to the procedure, plant management has the discretion to authorize extensions based on proper documentation and on discussions-with the responsible Department Superintendent.
The inspector reviewed the authorization documentation and found the close-out date for the Temporary Modification is June 1990 per Plant Nuclear Safety Review Committee (PNSRC)
No. 2278.
The plant intends to eliminate the four bottle nitrogen system, because it is not used, via Plant Modification No. 12-PM-491.
c Unit 2 Auxiliary Equipment Operator (AEO) logs lacked signoffs by oncoming AEOs (signifying that they have received a turnover);
and by Unit supervision, as specified by OHI-4012, "Conduct of Operations (Shift Turnover)."
A sample of AEO logs for the period June
through July 31, 1989 were reviewed.
The inspector saw many instances (over 20-percent of the logsheets)
where incomplete reviews were performed as indicated by lack of the proper signatures.
Following procedure OHI-4012 is required by Technical Specification 6.8.l.a
(through reference to Regulatory Guide 1.33 Appendix A).
Failure to perform the reviews and signoffs per the procedure is considered a
violation of the referenced Technical Specification (Violation 316/89023-01).
This item was brought to the attention of management and resulted in the issuance of two Operations Department memos dated July 28 and August 8, 1989, emphasizing the need for adherence to OHI-4012.
Subsequently, the inspector checked the logs from August 4-29, 1989 and noted only one instance of the Unit Supervisor failing to sign the log on August 18.
As a result, the Notice of Violation will not require a written response, since adequate corrective measures have been implemented and verified.
d.
At the request of NRC Region III, which had received an anonymous allegation to the effect that safety injection check valve 2-SI-158-Ll had a substantial bonnet leak (several gallons per minute) the inspector made a specific leak inspection of the subject valve on August 15, 1989.
Adjacent valve 2-SI-158-L4 was also inspected.
Both valves showed indications of past minor leakage in that both had small deposits of crystalized boric acid on or around them.
Neither valve was leaking at all at the time of the inspection.
The inspector considered his observations to be consistent with the licensee's determinations on an ongoing basis that leakage from either or both of the valves totalled substantially less than one gpm.
This information was provided to NRC Region III.
One violation, and no deviations, unresolved or open items were identified.
11.
Mana ement Interview (30703)
The inspectors met with licensee representatives (denoted in Paragraph I)
on August 30, 1989 to discuss the scope and findings of the inspection as described in these Details.
In addition, the inspector also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspector during the inspection.
The licensee did not identify any such documents/processes as proprietary.
The following items were specifically discussed:
a.
The violation described in Paragraph 10.c.
b.
The snubber inspection described in paragraph 6.d.
NRC 766 FORM NRC FORM 766 U.S.
NUCLEAR REGULATORY COMMISSION (6-83)
IE MC 0535 INSPECTOR'S REPORT OFFICE OF INSPECTION AND ENFORCEMENT asse PRINCIPAL INSPECTOR
B. L. Jorgensen R
ME
- B. L. Bur ess SITE NAME D. C.
Cook Units
8
TRANSACTION:
DOCKET NO
- REPORT (15-19)
TYPE (1): (2-14) 8-DIGITS: NO.:
.
- ~l-I I II: 0661 6-OOOIVV:
66666216 D-DELETE C
R-REPLACE
- NEXT INSPEC.
MO.
YR.
PERIOD OF INSPECTION
- INSPECTION PERFORMED BY (33)
FROM: (21-26)
TO :(27-32)
1-REGIONAL OFFICE STAFF
OTHER:
g: gY: Y$
- N,: gY: YR:X 2-RESIDENT INSPECTO 3-PAT
- 89
- 08
- 89
%%'++%%%'*+~NY*
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Yh'YY* *
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ORGANIZATION CODE OF REGION/HQ CONDUCTING. ACTIVITY:
REGIONAL ACTION (37)
III
~ 2-REGIONAL OFFICE LETTER TYPE OF ACTIVITY CONDUCTED (38-39)
X
-
Y-G
. ISI 1-PLA S
.
1-N IRY 03-INCIDENT 07-SPEC IAL 1 1-INVENT.VER.
15-INVESTI-04-ENFORCEMENT 08-VENDOR 12-SHIPMENT/EXPORT GATION 05-MGMT.AUDIT 09-MAT.ACCT 13-IMPORT A:
B :C: D:
- L I I-VIOL TIO
- 3-DEVIATION
- 4-VIOLATIONE DEVIATION INSPECTION/INVESTIGATION FINDINGS (40)
TOTAL NO OF
VIOLATIONS 8I
- DEVIATIONS (41-42)
A
'NFORCEMENT
CONFERENCE
HELD 1-YES (43)
- REPORT CONTAIN
- 2.790 INFO.
(44)
LETT R
NR RM
OR REG.
TO HQ.
FOR
LETTER ISSUED
- ACTION (45-50)
(51-56)
NRC FORM 766 MODULE INFORMATION
- N:
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T: M:
Y: B:
P: E:
E: R:
M P:
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A E:L RC-: MDL ORD BR I
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T I T N A S
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C N
E U:L:
D M:E U B:V:
R E:E:
E R:L:
B
- 3:
0 :7
- 0:3:
EXIT MEETING (MAP 2)
B
5 :4
- 7: 0:0:
A 0:0:3 0:0:3 0:1:3
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PROCEDURES (MAP 3)
0:1:4
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- 5:7:
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OPS SAFETY (MAP 2)
B
- 7:
1 :7
- 1:0:
B
5 :6
- 7: 2:6:
SURVEILLANCE (MAP 2)
B
5 :6
- 7: 0:3:
MAINTENANCE (MAP 2)
A 0:0:4 0 0'5 009 0:0:7 0:3:8 0:4:3 0:1:0
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WALKDOWN (MAP 2)
0:0:2
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B
- 7:
- 7: 1:1:
STARTUP AFTER REFUEL (MAP 2):
B
5 :7
- 7: 1:4:
WINTER PREPS (MAP 2)
B B
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B:
B:
5 :9
- 2
- 7: 0:0:
LER - SITE (MAP 2)
5 :9
- 7: 0:1:
FOLLOW UP A
A 0:0:7 0 0'7 0:1:0 0:0:8
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VIOLATION OR DEVIATION
NRC FORM 766 MODULE INFORMATION R
ORD T:
Y:
p o
E:
N:
UM:
8:
E:
R:
pH:
A:
S:
E:
R C:
M H:
C A A:
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D U T:
U A E:
R L R:
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- 13 1 :1 1 :19:
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26:
- 9:
2 :7
- 0:2:
A :
- 0:0:3
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FOLLOW N/C (MAP 2)
- 0:0:3
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5 :9
- 7: 0:3:
IEB/IAL/GL (MAP 3)
5 :9
- 7: 0:2:
A:
8 '
A :
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SIGNIF EVENT (MAP 2)': 8:
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8 :
5 :2
- 0: 2:7:
IEB 87-02 (MAP 3)
5 :6
- 7: 1:5:
- A:
8:
A '
popo]
- 0:0:2
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CONTAIN INTEGRITY (MAP 3): 8:
~
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- 8:
2 :2
- 0:1:
EMER DETECT/CLASSIF (MAP 3)
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A:
8:
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8 :
5 :8
- 2: 0:3:
- A:
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EMER NOTIF/COMMUN (MAP 3)
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5 :6
- 7: 0:5:
A '
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PREP FOR REFUEL (MAP 3)
8 '
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8 :
5 :6
- 7: 1:0:
REFUEL ACTIV'S (MAP 3)
A:
8:
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8:
5 :8
- 7: 0:0:
- A:
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REFUEL POOL ACTIV'S (MAP 3)
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NRC FORM 766 INSPECTOR'S REPORT (CONTINUATION)
DOCKET NUMBER
~
~
~'
~
~
D LE NUMBER
- 5:7 :1:7 :0 :7
- B:
- .
ViorA'TTON SEVERTTY
OR DEVIATION
~
x
~
- RELATED:
~
~
~
~
~
~
~
~
- B D:
(27)
(28) :(29):
VIOLATION OR DEVIATION(ENTER UP TO 2400 CHARACTERS FOR EACH ITEM)
l.
Unit 2 Technical Specification 6.8.1.a requires procedures be implemented as recommended in Appendix "A" of Regulatory Guide 1.33, November, 1972, which includes, at Item A.7, administrative procedures for shift and relief turnover.
Licensee procedure OHI-4012, "Conduct of Operations (Shift Turnover)", establishes requirements for review and evaluation of conditions associated with the position or station to be assumed by each shift operator prior to the start of the tour of duty.
The procedure specifically requires that, prior to relief, the oncoming Unit Supervisor review and sign the completed auxiliary equipment operator (AEO) log sheet.
The procedure additionally requires that, prior to the start of each shift, the oncoming AEO sign the AEO log sheet, which signifies the AEO has possession of tour keys, addition to indication of assumption of the tour of duty.
This is a Severity Level IV violation (Supplement 7).
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4 RP 0516B SALP FUNCTIONAL AREA ASSESSMENT AND PRELIMINARY INSPECTOR EVALUATION FORM Facility:
D.C.Cook-Unit I Ins ection Re ort No.:
315/89023 PLANT OPERATIONS
'
DI L G CAL C NTR L X
X
- ev:
n1
a an
~
~
~
~
~
~
~
M IN NANCE R
E LL NC X
- P:
~ ED
S RG NC P
P DN
X
- R:
GIN NGT
. SPPR
- 0:
AN ~
~
~
~
~
OTHER
~
~
~
~
CRITERIA FOR DETERMINING CATEGORY RATING 1.
Management Involvemen 1n ssur>ng ua 1 y.
2.
Approach to Resolution of Technical Issues from a Safety Standpoint.
3.
Responsiveness to NRC Initiatives.
4.
Enforcement History.
5.
Operational and Construction Events.
6.
Staffing (including management).
RATING KEY:
(For Categories 2 - Declining and 3, provide narrative basis for conclusion)
Category I Category
Category 2 - Declining Category
- X: Inspector(s)
concerns adequately addressed or Lead Inspection Evaluati n
rm being proce Inspector 7%4 g
sen Signa re)
a ssed.
ction Chi na ure a
e
RP 0516B SALP FUNCTIONAL AREA ASSESSMENT AND PRELIMINARY INSPECTOR EVALUATION FORM Facility:
D.C.Cook-Unit 2 Inspection Report No.:
316/89023 N
I L
- x
- I
- I:I
- ev: nit: attn
N P
~
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~
~
- RDIL L
NRLS
~
X
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~
~
~
~
~
~
~
~
~
- 0:
- MAINTENANCE/SURVEILLANCE
SEC RITY
~
v
~
~
Il
~
- P:
- R:
~ O
~
SAFET ASSESS/g ALI RI.:X::::::::1:S:
~
~
~
~
OTHER
~
~
~
~
~
~
CRITERIA FOR DETERMINING CATEGORY RATING 1.
Management Involvement
~n ssuring ua i y.
2.
Approach to Resolution of Technical Issues from a Safety Standpoint.
3.
Responsiveness to NRC Initiatives.
4.
Enforcement History.
5.
Operational and Construction Events.
6.
Staffing (including management).
RATING KEY:
(For Categories 2 - Declining and 3, provide narrative basis for conclusion)
Category
Category 2 - Declining Category
Category
- X : Inspector(s)
concerns adequately addressed or
- Inspection Evaluation Form being Lead Inspecto B.L. Jorgens Signature)
rocessed.
ection C
gnature a