IR 05000341/1987020
| ML20234C936 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 06/29/1987 |
| From: | Greenman E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20234C924 | List: |
| References | |
| 50-341-87-20, NUDOCS 8707070024 | |
| Download: ML20234C936 (13) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
i Report No. 50-341/87020(DRP)
Docket No. 50-341 License No. NPF-43 Licensee:
Detroit Edison Company
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2000 Second Avenue Detroit, MI 48226 I
Facility Name:
Fermi 2 Inspection At:
Fermi Site, Newport, Michigan Inspection Conducted:
May 5 through June 8, 1987 Inspectors:
W. G. Rogers M. E. Parker W. Gunther D. Beckman Approved By.3 G Gr t im n, D p Director 6/ s /r 7
Division of Reactor Projects Date'
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Inspection Summary Inspection on May 5 through June 8,1987, (Report No. 50-341/87020(DRP))
Areas Inspected:
Routine, unannounced inspection by resident inspectors of
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previously identified items, LERs, events, regional requests, operational safety, maintenance, surveillance, startup test witnessing and observation, inanagement meetings, report review, and plant trips.
Results:
One violation was identified (failure to provide specific action for modification of software in surveillance procedure, Paragraph 8).
Two unresolved items were identified (Paragraphs 6.a. and 6.b.) and three open items were identified (Paragraphs 5.a.(1), 5.a.(2), and 6).
No deviations were identified.
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{DR ADOCK 05000341 PDR
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a DETAILS
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Persons Contacted i
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Detroit Edison Company
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F. Abramson, Operations Engineer
+F. Agosti, Vice President, Nuclear Operations
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L. Bregni, Compliance Engineer
- +S. Cashell, Licensing
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G. Debner, Startup Engineer, Test Phase l
+K. Earle, Technical Engineer, Nuclear Production
- +L. Esau, Maintenance Engineer
- S. Frost, Licensing
+C. Gelletly, Nuclear Projects and Plant Engineering
+J. Green, Systems Engineering
R. Kelm, Director, Nuclear Security
- J. Leman, Nuclear Production, Director Plant Safety
- R. Lenart, Plant Manager, Nuclear Production
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+A. Lim, Systems Engineering R. May, Superintendent, Maintenance and Modification
+W. Orser, Vice President, Nuclear Engineering and Services J. Plona, Operations Support Engineer E. Preston, Director, Plant Safety
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+T. Randazzo, Director, Regulatory Affairs
+B. Sheffel, Nuclear Production, Technical Engineering ISI
- M. Sierra, Technical Engineering
- B. R. Sylvia, Group Vice President, Nuclear Operations
+J. Thorpe, Systems Engineering
+G. Trahey, Director, Quality Assurance W. Tucker, Superintendent, Operations b.
U.S. Nuclear Regulatory Commission D. Beckman, NRC Contractor
- W. Gunther, NRC Contractor
+M. Parker, Resident Inspector i
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- +W. Rogers, Senior Resident Inspector
- Denotes those who attended the exit meeting on May 22, 1987.
+ Denotes those who attended the exit meeting on June 5, 1987.
2.
Followup on Inspector Identif.ied Items (92701)
(0 pen) Unresolved Item (341/86034-02(DRP)):
This item concerns the controls placed on the reactor building corner room doors as they relate to meeting flood protection and separation requirements to satisfy single and common mode failure criteria for the emergency core cooling system (ECCS).
The inspectors reviewed the licensee's Technical Specification Interpretation Number TS-87-004, Revision 1, which was written to address this item.
The inspectors concluded that this interpretation adequately addressed the inspectors' concerns.
Subsequently, the licensee revised the interpretation (Revision 2) to allow an eight-hour grace period prior
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to declaring the associated ECCS equipment inoperable.
The inspectors
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consider this grace period inappropriate and not consistent with ECCS
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Limiting Condition for Operation time limitations.
Through discussions with the licensee, the licensen has agreed to revise the interpretation to delete the eight-hour grace period.
This item is to remain open pending revision of the Technical Specification Interpretation by the licensee.
3.
Licensee Event Reports Followup (92700)
During this report period, a contractor from Brookhaven National
Laboratory (BNL) completed an inspection of LERs associated with the reactor water cleanup (RWCU) system.
The resident inspector reviewed the activities of the contractor and concluded that the LERs were recdy to close or status as snown below.
Specific followup accomplished by the BNL contractor is documented in Section 6.1 of Attachment 1 to this inspection report.
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(Closed) LER 85-024 RWCU system isolation valve closure.
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(Closed) LER 85-025 Isolation of RWCU system.
(Closed) LER 85025-01 Isolation of RWCU system.
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(Closed) LER 85-027 RWCU pump trip du,e to G33F004 going closed.
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(Closed) LER 85-028 RWCU spurious isolation, (Closed) LER 85-031 RWCU isolation.
(Closed) LER 85-034 RWCU isolation.
(Closed) LER 85-050 RWCU system spurious isolation.
(Closed) LER 95-061 RWCU isolation.
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(Closed) LER 85-063 RWCU isolation due to high differential i
flow.
l (Closed) LER 85-064 RWCU steam leak detection system isolates.
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(Closed) LER 86-003 RWCU isolation valve closure.
(Cicsed) LER 86-021 RWCU isolation due to a faulty test cable.
(Closed) LER 86-046 Personnel error resuited in reversed thermocouple leads and inoperable RWCU heat exchanger room differential temperature instrument channel.
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(Closed) LER 86-049 RWCU system gasket failure causes a high differential flow resulting in an ESF actuation.
(Closed) LER 87-001 Inadequate communication results in isolation of RWCU during calibration.
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(Closed) LER 87-005 Ga';ket failure results in manual ESF isolation of RWCU.
(0 pen) LER 85-046:
This LER will remain open pending completion of a
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I flow calibration at full operating temperature.
The licensee has scheduled the calibration for June 1987.
(0 pen) LER 85-066:
This LER is closely associated with LER 85-046 and will remain open pending completion of the above stated flow calibration.
(Closed) LER 86-012:
RWCU failure to meet safety analysis report criteria, identified a violation of 10 CFR 50, Appendix A, Criterion 21, in that the RWCU leak detection / isolation system was not designed to withstand a single failure and still perform its protective function.
The inspector evaluated this violation under 10 CFR 2 Appendix C.V.A.
The inspector concluded that the licensee met all the crit ~ ria as stated e
below:
a.
The design error was identified by the licensee.
b.
The safety significance of the design error would have resulted in the isolation of the RWCU system in 44 seconds versus 20 seconds.
The ramifications of the 24 second differer.ce would have been to reduce the environmental qualification lifetime of some of the i
equipment in the reactor building.
All safety-related systems would have performea their safety function when called upon and no safety iimit would base been violated.
Therefore, given the miniinal safety j
significance of the violation it would be characterized as a Severity Level IV.
c.
An LER wat, subniitted within the specified time requirement.
d.
The design error was identified during a maintenance cutage. A facility modification was accomplished to the system to assure that it would properly isolate the RWCU systa with a single failure present.
That modification was performed during the maintenance outage and prior to declaring the system operable.
e.
No corrective action to a previous violation issued within the iast two years could have reasonably corrected this violation.
(Closed) LER 86-034-00 and LER 86-034-01:
The BNL contractor recommended leaving this LER open pending the implementation of a one-second time delay in the system.
The licensee does not intend to do this, i
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Environmental qualification of equipment in the area would be impacted
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by the additional time delay. Based upon the licensee's reasons, the
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inspector concluded that the LER would be closed without further followup.
(0 pen) LER 86-050:
This LER will remain open pending implementation of PDC-6949 which will install some additional check valves in the RWCU system to preclude water intrusion into the essential instrument air system. The licensee will provide a commitment date by the end of the
June routine inspection period as to when that facility change will be l
performed.
No violations or deviations were identified in this area.
4.
Followup of Events (93702)
Duric:9 the inspection p6riod, several events occurred, some of which required prompt notification of the NRC pursuant to 10 CFR 50.72. itse inspectors pursued the events onsite with licensee and/or other NRC officials.
In each case, the inspectors verified that the notification was correct and timely, if appropriate, that the licensee was taking prompt and appropriate actions, that activities were conducted within regulatory requirements and that corrective actions would prevent future recurrence. The specific events are as follows:
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May 10, 1987 - Engineering Safety Feature Actuation, resulting from false low reactor vessel water level signal.
May 13, 1987 Reactor scram on Level 3 as a result of turbine feed
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pump trip.
May 20, 1987 Notification of Unusual Event as a result of pressure
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boundary leakige.
May 20, 1987 - Engineering Safety Feature Actuation resulting from RWCU heat exchanger outlet temperature high.
May 21, 1987 - Engineering Safety Feature Actuation resulting from RWCU heat exchanger outlet temperature high.
J May 28, 1987 - Engineering Safety Feature Actuation causing emergency diesel generators to autostart.
i June 1, 1987 - Engineering Safety Feature Actuation causing CCHVAC to shift to recirculation mode.
During the inspection period, the inspector selected two DERs to overview the licensee's evaluation and corrective actions. The two DERs were:
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DER 87-171:
Identification of incorrect data in the dynamic portion
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of the NSS process computer. The inspector ascertained that the incorrect data was not used in any surveillance or by operators to verify compliance with thera.al inits. The technical section is presently reviewing the controls established over data inputs into the process computer. The licensee committed to keep the inspector informe of the corrective actions. Therefore, this matter will be reviewed et a future date upon establishment of those corrective actions.
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DER 87-174:
Inappropriate use of the EECW override switches.
The inspector ascertained that the licensee has suspended all uses of the override switches until it can be determined what the design intent and regulatory allowances are for the switches. The inspector will review this matter in a future inspection report period.
The inspectors also reviewed security-related reportable events during the inspection period.
No violations or deviations were identified in this area.
5.
Followup on Regional Requests (92705)
a.
EX0 Sensor, Inc. Quclity Assurance
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In the spring of 1985 an I&E Vendor Branch inspection was performed of the EX0 Sensor, Inc. The inspection identified numerous concerns with the quality assurance aspects of EX0 Sensor. Upon receipt of the inspection findings, NRC regional management requested all licensees who had received equipment from EX0 Sensor, Inc. to confirm the acceptability of the procured equipment.
In a letter dated March 17, 1986, the resident inspector was requested to evaluate the licensee's response and have the licensee resolve any problems found. The resident inspector assigned initial review of this matter to a Brookhaven National Laboratory (BNL) contractor.
The inspection results of the contractor are presented in Section 6.4 of Attachmer.t 1.
The resident inspector reviewed the BNL contractor inspection results and determined:
(1) The action plan implementation to correct identified problems with ths hydrogen / oxygen analyzers is considered an open item (341/87020-01(DRP)).
The licensee has committed to keep the inspector informed on a monthly basis as to the progress of the action plan.
(2) The licensee had been untimely in performing a 10 CFR 21 evaluation of the hydrogen / oxygen analyzer sensor failures.
Also, this evaluation was only one of a family of equipment failures where 10 CFR 21 evaluations had been untimely.
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i licensee has committed to completing the 10 CFR 21 evaluation
by June 30, 1987, and performance'of that evaluation is considered an open. item (341/87020-02(DRP)).
The. failure to perform the 10 CFR 21 evaluations in an appropriate time frame is considered a violation.
The inspector evaluated this-i violation under 10 CFR 2 Appendix C.V. A.
The inspector
.l concluded that the licensee met all-the criteria as stated j
below:
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(a) The situation had been previously. identified by the:
licensee.
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(b) The violation would be characterized as a Severity Level IV under 10 CFR 2 Appendix C,_ Supplement VII.D.1.
(c) Deportability was not applicable in this instance.
(d) The 10 CFR 21 evaluations were performed with only two -
outstanding at the end of.the inspection period.
Procedural changes were made to assure 10 CFR 21 evaluations were performed within 60 days of their identification.
(e) There have been no violations associated with the 10 CFR 21 process within the last two years.
b.
I&C Overtime As a result of concerns expressed in a May 11, 1987, meeting with Detroit Edison on the Maintenance Survey Team findings, the resident inspectors were requested to followup on the use of excessive overtime by the I&C organization.
NUREG 0737 overtime limitations have been incorporated into Technical Specification * and implemented in Plant Operation Manual (P0M) 12.000.114, _" Overtime Guidelines."
The inspectors verified, through a review of work hours, that the I&C technicians were complying with POM 12.000.114 and,' as such, NUREG-0737.
Even though the technicians have been routinely performing overtime work, this overtime is within the limitations-specified in NUREG 0737.
The licensee has taken action to reduce the overall overtime for the I&C organization.
c.
Backshift Coverage During the inspection period, the inspectors performed a backshift coverage inspection as a result of concerns identified at other licensed facilities. This inspection was performed during the I
afternoon and morning shifts to observe control room decorum and activities, During this time, the inspectors considered the personnel on shift alert and knowledgeable of plant status. No abnormalities were noted during the backshift coverage.
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6.
Operational Safety Verification (71707)
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The inspectors observed control room operations, reviewed applicable logs and conducted discussions with control room operators during the period from May 5 through June 8, 1987.
The inspectors verified the operability
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of selected emergency systems, reviewed tagout records and verified
proper return to service of affected components.
Tours of the reactor building and turbine building were conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, and excessive q
vibrations and to verify that maintenance requests had been initiated for i
equipment in need of maintenance.
The inspectors, by observation and direct interview, verified that the physical security plan was being implemented in accordance with the station security plan.
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The inspectors observed plant housekeeping / cleanliness conditions and-l verified implementation of radiation protection controls.
During the inspection, the inspectors walked down the accessible portions of the high pressure coolant injection (HPCI) system to verify operability by l
comparing system lineup with plant drawings, as-built configuration or present valve lineup lists; observing equipment conditions that could degrade performance; and verified that instrumentation was properly valved, functioning, and calibrated.
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During the walkdown of the HPCI system, the inspector identified several items which have been brought to the attention of the licensee but which need additional followup by both the inspector and the licensee.
It should be noted that the HPCI system was declared inoperable per Technical Specifications pending resolution of problems identified during the performance of startup testing.
a.
E41-F011, HPCI Test Return Line to CST Isolation Valve, was closed but in the energized position.
This valve, as a condition of the license, is required to be closed and deenergized and only operable during HPCI and RCIC testing periods.
No testing was in process at the time.
The valve condition was identified to the oncoming Nuclear Shift Supervisor on June 4,1987, at approximately 11:00 p.m. EDT.
Previous startup testing on the HPCI system was suspended at approximately 3:00 a.m. EDT on June 4, 1987.
The licensee took immediate action to deenergize the valve in the closed positian.
This is considered an unresolved item (341/87020-03(DRP)) pending further review by the inspector.
b.
E41-F022, HPCI Turbine Exhaust Line Drain to Suppression Chamber Valve, was founJ in the closed position but is required by the system operating procedure to be open.
No abnormal lineup sheet I
(ALS) was generated identifying this abnormal position.
Review of
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the sequence of event recorder revealed that the valve was secured at 2:15 p.m. EDT on June 4,1987, approximately nine hours eadier.
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The licensee reviewed the conditions and could determine no reason
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for the valve to be in the closed position and subsequently placed the valve in the open position.
This is considered an unresolved item (341/87020-04(DRP)) pending further review by both the NRC and the licensee.
c.
Review of the out-of-service log (OSL) identified that the HPCI system was declared inoperable as a result of the E41-F011 being j
energized.
The log did not, however, identify that, (1) HPCI was currently tagged out to repair E41-F020, suction relief valve, and (2) HPCI was inoperable pending resolution of E41-F006, HPCI Pump Discharge Inboard Isolation Valve problems identified during startup testing.
d.
Miscellaneous procedure valve nomenclature and hardware problems.
These reviews and observations were conducted to verify that' facility operations were in conformance with tne requirements established under technical specifications, 10 CFR, and administrative procedures.
During the inspection period, the BNL contractor performed a limited walkdown of some safety-related panels.
The results of that walkdown are presented in Section 6.5 of Attachment 1.
The resident inspector reviewed the findings and determined that'the labeling of the Riley
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temperature monitoring modules would be considered an open item (341/87020-05(DRP)).
The licensee committed to relabel the modules by August 1, 1987.
No violations or deviations were identified in this area.
7.
Monthly Maintenance Observation (62703)
Station maintenance activities on safety-related systems and components listed below were observed to ascertain that they 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 were inspected as applicable; functional testing and/or calibrations were performed prior to returning components or systems to service; quality I
control records were maintained; activities were accomplished by
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qualified personnel; parts and materials used were properly certified; j
radiological controls were implemented; and fire prevention controls were l
implemented.
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Work requests were reviewed to determine the status of outstanding jobs
and to assure that priority is assigned to safety-relatea equipment maintenance which may affect system performance.
The following maintenance activities were observeti:
Disassembly of south reactor feedpump.
- Residual heat removal (RHR) B motor termination box repair.
- Following completion of maintenance on the RHR pump motor, the inspectors verified that these systems had been returned to service properly.
a.
On May 25, 1987, the licensee identified that the mounting bolts on i
the RHR B motor termination box were found sheared.
The sheared i
bolts were sent to engineering research for evaluation.
The analysis determ'ned that the cause of failure was '/ibration induced fatigue failure of the bolts.
RHR B motor termination box is unique to the RHR motors in that the 8 motor termination box weighs 450 pounds verses 45 pounds for the other'three RHR motor termination boxes.
The RHR B motor termination box was specifically ordered from the vendor to accommodate environmentally qualified motor termination splices.
In reviewing the circumstances of the failure, it was identified that a seismic evaluation was performed but did not include a vibration analysis.
The licensee has subsequently completed their engineering analysis aiad issued an Engineering Design Package (EDP) 7440.
This EDF consists of increasing the motor termination box mounting bolt size from 1/2-fnch diameter to 1-inch diameter to eliminate any vibratory concerns.
This item has been referred to Region III Division of Reactor Safety to review the licensee's seismic and vibration analysis.
b.
During the inspection period, the BNL contractor reviewed portions of the motor operated valve (MOV) repair procedures / training and the preventative maintenance program.
The results of the inspection are presented in Section 6.3 of Attachment 1.
The problems identified in the preventative maintenance program will be followed under the followup to the mainten?nce survey team and will be considered only as another facet of unresolved item 341/35024-01.
With regard to the MOV training the licensee committed to keep the resident inspector abreast of any schedule slippage of the training.
c.
During the inspection period, a consultant was engaged by NRC Region III to perform a followup and overview of the licensee's investigation and corrective actions concerning the failure of the south reactor feedwater pump turbine on May 13, 1987.
The consultant's inspection results and observations are presented in i
Attachment 2 to this inspection report.
No viclations or deviations were identified in this area.
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8.
Monthly Surveillance Observation (61726)
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The inspectors observed surveillance testing required by Technical Specifications and verified that:
testing was performed in accordann
with adequate procedures, test instrumentation was calibrated, limiting conditions for operation were met, removal and restoration of the affected components were accomplished, test results conformed with Technical Specifications and procedure requirements and were reviewed by personnel other than the individual directing the test, and any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel.
The inspectors also witnessed portions of the following test activities:
24.609 Rod Soquence Control System Functional Test.
- 24.307.14 Emergency Diesel Generator No. 11 - Start and Load Test.
- 44.010.129 Reactor Protection System - Average Power Range Monitur F Channel Functional Test.
The inspectors performed a record review of completed surveillance tests.
The review was to determine that the test was acccuplished within the
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reouired Technical Specification time interval, procedural steps were properly initiated, the procedure acceptancs criteria were met, independent verifications were accomplished by people other than those l
performing the test, and the tests were signed in and out of the control J
room surveillance log book.
The surveillance tests reviewed were:
54.000.07 Core Performance Parameter Check.
- 71.080.01 Chemistry Surveillance.
During the inspection period the BNL contractor reviewed numerous surveillance procedures as identified in Sections 6.2 and 6.4 of Attachment 1 to the inspection report.
The resident inspector reviewed the contractor results and determined:
a.
The minor procedural discrepancies had been handled appropriately.
b.
The licensee considered the current procedural circuit checks as adequate and did not accept the recommendation to perform a final loop check of instrumentation.
c.
The licensee considered adding the process parameters in the data j
sheets as not being necessary.
d.
Surveillance procedure 44.120.028 did not appropriately procedurally control the manipulations of the hydrogen / oxygen analyzer sof tware.
Failure to provide the specific actions of how to modify the
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software in the procedure is considered a violation (341-87020-06(DRP))
of Technical Specification 6.8.1.d.
The licensee committed to
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revise the surveillance procedure and the one for the companion j
channel prior to use of the procedure or September 1, 1987,
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whichever comes first.
Additionally, the licensee reviewed all surveillance procedures associated with microprocessors and did not note the same procedural inadequacies.
Therefore, no further corrective action need be taken by the licensee in this area.
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Startup Test Witnessing and Observation (72302)
i The inspectors witnessed attempts by the licensee to perform HPCI Hot Vessel Injection Test, STUT.03E.015.
Followup to the problems encountered and witnessing a successful test completion will be done in a future inspection report period.
10.
Management Meetings (30702)
On May 11, 1987, the Regional Administrator was on site.
He attended a meeting with the licensee and NRR representatives to discuss observations of the maintenance survey team.
Later in the afternoon he introduced himself to some of the Monroe County Commissioners and discussed matters of mutual interest.
11.
Report Review (90713)
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During the inspection period, the inspector reviewed the licensee's
Menthly Operating Report for April 1987. ' The inspector confirmed that the information provided met the requirements of Technical Specification i
6.6. A.3 and Regulatory Guide 1.16.
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No violations or deviations were identified in this area.
12.
Plant Trips (93702)
Following the plant trip on May 13, 1987, the inspectors ascertained the
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status of the reactor and safety systems by observation of control room indicators and discussions with licensee personnel concerning plant parameters, emergency system status and reactor coolant chemistry.
The inspectors verified the establishment of proper communications and reviewed the corrective actions taken by the licensee.
All systems responded as expected, and the plant was returned to operation on May 14, 1987.
During the followup into the reactor scram on May 13, 1987, as a result of the loss of the south reactor feed pump, the inspectors expressed concern regarding the philosophy and training provided to reactor
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operators concerning operator actions to be taken on loss of feedwater.
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As a result of these concerns, the licensee has evaluated and provided
additional guidance to the operators on loss of feedwater.
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has modified A0P 20.107.01, " Loss of Feedwater or Feedwater Control," to provide this guidance and has enforced this guidance in required reading i
and night orders.
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No violations or deviations were identified.
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i 13.
Unresolved Items Unresolved items are matters about which more information is requir'ed in order to ascertain whether they are acceptable items, violations or deviations.
Unresolved items disclosed doring the inspection are
discussed in Paragraphs 6.a and 6.b.
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Open Items Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involve some action on the part of the NRC or licensee or both.
Open~ items disclosed during
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'the inspection are discussed in Paragraphs 5.a.(1), 5.a.(2), and 6.
15.
Exit Interview (30703)
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The inspectors met with licensee representatives (denoted in Paragraph 1)
on May 22 and June 5, 1987, and informally throughout the inspection period and summarized the scope and findings of the inspection activities.
The inspectors also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspectors during the inspection.
The licensee did not identify any such documents / processes as proprietary.
The licensee acknowledged the findings of the inspection.
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