IR 05000245/1987021
| ML20235J555 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 09/21/1987 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20235J529 | List: |
| References | |
| 50-245-87-21, NUDOCS 8710020013 | |
| Download: ML20235J555 (10) | |
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LS. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.-
50-245/87-21 Docket No.:
50-245 License No:
DPR-21 Licensee:
Northeast Nuclear Energy Company Facility:
Millstone Nuclear Power Station, Waterford, Connecticut Inspection At:
Millstone Unit 1 Dates:
August 11, 1987 through September 8, 1987 Inspectors:
Geoffrey E. Grant, Resident Inspector Approved:
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_9 /21/(7 E. C. McCabe, Chief, Reactor Projects Date Section 3B, DRP Summary:
Report No. 50-245/87-21
_(A_ugust 11 to September 8, 1987)
Area: 7;.spected:
This inspection included routine NRC resident (89 hours0.00103 days <br />0.0247 hours <br />1.471561e-4 weeks <br />3.38645e-5 months <br />)
inspet.. ion of plant operations, surveillance, maintenance, radiation protec-tion, physical security, fire protection, periodic and special reports and Instrumentation and Control surveillance.
Results: One violation was identified:
it indicates a potential trend toward inadequacies in I&C surveillance testing (Details 4 and 5.b). There were three reactor scrams during this report period (Detail 5); contributing factors
included procedure weakness, personnel error, and faulty equipment performance.
The Regional Administrator's site inspection confirmed the licensee's commit-ment to quality maintenance and plant cleanliness (Detail 7).
Planning and management of the 1987 refueling outage was evaluated as excellent, but the inspection also noted a potential for improving outage workload by performing more turbine safety committee reviews before an outage begins (Detail 8).
8710020013 870924 PDR ADOCK 05000245 G
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TABLE OF CONTENTS Page
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Persons Contacted................
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2.
Summary of Facility Activities........................
3.
Operational Safety Veri fication.....................
4.
Inadequate Control of Surveillance Testing.............
5.
Reactor Scrams.........................................
6.
Review o f Periodic and Special Reports................
7.
Regional Administrator Inspection......................
8.
Outage Planning and Management..........
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9.
Management Meeting.
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DETAILS 1.
Persons Contacted Mr. S. Scace, Station Superintendent Mr. J. Stetz, Unit 1 Superintendent The inspector also contacted other licensee employees including members of the Operations, Radiation Protection, Chemistry, Instrument and Control, Maintenance, Reactor Engineering, and Security Departments.
2.
Summary of Facility Activities During this report period the plant completed a 10-week refueling outage and recommenced power operations.
Initial startup and criticality for Cycle 12 occurred on August 14. A reactor trip was experienced early in
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the startup on August 14, with a subsequent restart on August 15 (see
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Detail 7).
Startup testing and power ramping were performed until the
. unit reached 100% power on August 20.
A reactor trip occurred on August 26 (see Detail 5). The unit returned to full power on August 27. Another reactor trip occurred on September 3 (see Detail 5).
Several maintenance actions were completed and the unit was returned to full power on September 6.
3.
Operational Safety Verification The inspector observed plant operations during regular and back shift tours of the following areas:
Control Room Cable Vault Reactor Building Fence Line (Protected Area)
Diesel Generator Room Intake Structure Vital Switchgear Room Gas Turbine Building Turbine Building Control Room instruments were observed for correlation between channels, proper functioning, and conformance with Technical Specifications. Alarm conditions in effect and alarms received in the control room were reviewed and discussed with the operators.
Operator awareness and response to these conditions were reviewed. Operators were found cognizant of board I
and plant conditions.
Control room and shif t manning were compared with Technical Specification requirements.
Posting and control of radiation, contaminated and high radiation areas were inspected.
Use of and compli-ance. with Radiation Work Permits and use of required personnel monitoring devices were checked. Plant housekeeping controls were observed including
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control of flammable and other hazardous materials.
During plant tours, logs and records were reviewed to ensure compliance with station proce-
- dures, to determine if entries were correctly made, and to verify correct communication of equipment status. These records included various opera-ting logs, turnover sheets, tagout and jumper logs, and Plant Information
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Reports. The inspector observed selected actions concerning site security including personnel monitoring, access control, placement of physical barriers, and compensatory measures.
Inspections of the control room were performed on backshifts on August 25, 26 and 27. Routine power operations were observed.
Operators were alert and attentive to plant conditions.
Logs, records and plant status were up to date. No abnormel conditions were observed.
4.
Inadequate Control of Surveillance Testing As part of pre-startup surveillance testing on August 12, Instrumentation and Control (I&C) personnel were performing SP 413B, " Auto Blowdown Logic Test". This procedure verifies the logic sequence of the Automatic Blow-down System (ADS) by a simulated automatic initiation.
An I&C tech-nician's disconnection of test equipment during performance of this sur-veillance caused the
"A" and
"C" Low Pressure Coolant Injection (LPCI)
pumps to start and the LPCI injection valve to open.
The reactor was in cold shutdown at the time, and this actuation resulted in a LPCI discharge to the reactor vessel of approximately 10,000 gallons. The injection was I
terminated by operators in approximately one minute. The licensee stopped the surveillance, reviewed the sequence of events and counseled the tech-nicians involved. Testing was resumed after these actions were complete.
The resident inspector's review of this event identified concerns beyond those addressed by the licensee. Step 7.3 of SP 4138 directs the connec-tion of test equipment at connector 287-J1A on the blowdown test panel in CRP 932-F.
This test panel is clearly identified as the Automatic Blow-down test panel and connector 287-J1A is likewise clearly identified.
The I&C teciinician mistakenly connected the test equipment to connector 1530-J1A on the LPCI test panel, also located in CRP-932-F.
This test panel and connector are also clearly identified and are vertically sepa-rated from the other test panel by about two feet.
The inspector noted that step 7.2 of SP 4138 includes a NOTE:
" Installation and removal of jumpers and test equipment require acknowledgement and verification initials of person accomplishing same on data sheet." Surveillance form j
4138-1 has required performance and verification initial blocks for con-
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necting the test equipment to 287-J1A.
In this instance, independent verification of the test equipment connection was not accomplished by the I&C technician assisting in the surveillance test.
This contributed to the subsequent LPCI actuation.
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. Another example of inadequate control of surveillance. occurred ~ the next day, August 13.
In this instance, SP 408E, " Main Steam Line Isolation Valve Closure Functional Test". was to be performed to test the MSLIV closure scram logic.
The Initial Conditions section of SP 408E requires in Step 5.2 that relays 590-102A through H (inboard and outboard MSLIV position scram relays) be visually verified as energized.
Step 5.4 requires that all MSLIVs.be verified open by observing CRP 903. The sur-veillance form associated with this test requires ' the technician to initial when the test Prerequisites and Initial Conditions have been met.
The technicians performing this surveillance failed to accomplish these required verifications.
Because the MSLIVs were in fact closed, subse-quent commencement of the test resulted in a full Reactor Protection System scram signal.
As the reactor was shutdown and in a refueling outage, no rod motion occurred.
A third example of inattention during surveillance testing occurred on August 26.
While performing SP 404 C,
" Average Power Range Monitor i
Calibration and Functional Test," an I&C technician mistakenly placed two APRMs in a test condition simultaneously.
This caused a reactor scram i
from 100% power.
(See Detail 5.b)
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Failure to provide required independent verification as prescribed in the l
procedure (first example) and failure to establish required initial condi-
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tions (second example) are violations of Technical Specification 6.8.1 which requires that written procedures covering surveillance of safety-related equipment be implemented.
This also violates 10 CFR 50, Appendix B, Part V, which requires that activities affecting quality be accom-plished in accordance with documented instructions and procedures.
(VIO 50-245/87-21-01)
5.
Reactor Scrams a.
On August 14, while performing a reactor plant heat-up, a reactor scram occurred on Intermediate Range Monitor (IRM) Hi-Hi flux. At the time, the reactor was at 296 degrees F, 48 psig and Range 7 on the IRMs. The high flux condition was created by operator-initiated excessive control rod withdrawal. An increasing power trend result-ing from previous rod withdrawal coupled with the continuous with-drawal of rod 26-31 resulted in too large a reactivity addition, too fast a reactor period, and a scram on high flux.
Rod 26-31 was the first rod in its group to be withdrawn from the notch 12 group position.
Procedures allowed continuous rod withdrawal from this position to full out (notch position 48).
However, caution must be excercised by the operator during continuous withdrawal to avoid a reactor period that is too short.
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4-This was the ~ initial startup of the newly refueled Cycle 12 core.
Refueling placed 196 new GE-8B fuel assemblies in the core. These
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assemblies are zoned axially with Gadolinium (Gd) poison that changes concentration at the notch 12 level. The fuel is also zoned axially with a lower enrichment in the area of the top few notches.
These'
two factors cause a control rod near new GE-88 fuel assemblies to have a high worth as it is being withdrawn through the mid portion of the core (beyond notch position 12).
The magnitude of ' this response was not anticipated by the operator.
A three notch with-drawal of the rod was sufficient to cause a high flux condition..The licensee changed operating procedures to require single notch rod withdrawal from the notch 12 to notch 24 positions. The inspector had no further questions in this area.
b.
On August 26, the reactor tripped from 100% power.
The trip was caused by an I&C technician error while performing SP 404C, " Average Power Range Monitor Calibration and Functional Test." The technician was testing APRM #2.
Completion of step 7.5.2 lef t the APRM mode switch in a test position (ZERO) which automatically inputs a scram signal to "A" RPS.
He looked away from the panel to check the next step in the procedure, 7.5.3, which places the APRM mode switch in another test position (POWER / FLOW). When he turned back to the panel to accomplish this action, he mistakenly performed it on APRM #6.
This caused a scram signal in the
"B" RPS and a reactor scram. The technician was being observed by a senior technician who was unable to react quickly enough to prevent the incorrect action.
Plant response to the trip was normal and all safety systems operated cor-rec tly '.
The resident inspector re-emphasized the I&C surveillance concerns which had been previously discussed with licensee manage-ment (See Detail 4.) The licensee is reviewing these events.
c.
On September 3, the reactor tripped from 100s power.
The automatic scram was due to low pressure in the scram pilot air header.
The header pressure dropped below the trip setpoint of 53 psi due to a loss of instrument air system pressure.
A combination of events resulted in the loss of air pressure.
The normal supply to the instrument air system was lost when the Sullair compressor tripped on overload.
The standby Instrument air compressor started auto-matically on decreasing header pressure but subsequently failed to load as required.
Although the station air system can supply the instrument air system through a cross-connect line containing check valves, in this instance the check valves failed to open. Operators responded quickly to the lowering header pressure condition, but were unable to take sufficient corrective actions fast enough to avoid a reactor scram.
The plant response to the trip was normal and all safety systems operated correctly.
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l The licensee's investigation identified several items that require j
correction or more detailed study.
These included:
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The crcss-connect check valves were found to have failed due to an accumulation of water and debris in the piping.
This con-
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dition was corrected and daily blow-down of the line was estab-i lished.
No anticipatory alarms preceded the. loss of the Sullair com-O
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prassor.
The licensee is studying the advisability of having I
alerting alarms available to the operator.
The failure of the instrument air compressor to load was cor-
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rected within minutes of the reactor trip.
The licensee is studying the current system line-up that allows j
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the instrument air system to supply the station air system. The l
large demand that caused the loss of the Sullair compressor was due to station air demand.
The resident inspector reviewed the licensee's post-trip and correc-tive actions and found them acceptable.
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6.
Review of Periodic and Special Reports j
i Upon receipt, periodic and special reports submitted pursuant to Technical l
Specifications were reviewed.
This review verified that the reported
.information was valid and included the NRC required data; that test i
results and supporting information were consistent with design predictions and performance specifications; and that planned corrective actions were adequate for resolution of the problem.
The inspector also ascertained whether any reported information should be classified as an abnormal occurrence. The following reports were reviewed:
Monthly Operating Report for Millstone Unit 1 for the month of July, l
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j Monthly Operating Report for Millstone Unit 1 for the month of June,
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Monthly Operating Report for Millstone Unit 1 for the month of May,
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Special Report submitted in accordat.ce with Technical Specifications 3.12.E.2.b and 6.9.2.d.
This report detailed conditions resulting in a fire watch being posted in the Emergency Diesel Generator room for greater than 14 days.
This report was submitted as Licensee Event Report (LER) 87-12.
No deficiencies were noted during these reviews.
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7.
Regional Administrator Inspection On August 18,1987,.a material condition inspection was performed by the USNRC Region I Regional. Administrator.
Areas inspected included the Reactor Building, Emergency Gas Turbine Building, Emergency Diesel Gener-ator Room, portions of.the Turbine Building, Cable Vault and Control Room.
Instrumentation, cabling, control systems, valves, valve motor-operators, motors, pumps, electrical switchboards, batteries, and - motor generators.
Vital and emergency equipment were inspected to determine overall material condition and impact of the licensee's maintenance program. Additionally, housekeeping, general cleanliness and control of radiological areas were also observed.
l The detailed results of the inspection were discussed with licensee management on August 19 during the exit meeting and was provided to the licensee and are listed in Inspection Report 50-423/87-17 as an appendix.
The overall finding was that both the material condition.and cleanliness of the plant were good.
The general absence of debris and excellent housekeeping conditions were particularly noted in light of the plant's outage status.
Items noted for licensee followup will be reviewed during routine inspection activities (UNR 87-21-01).
8.
Outage Pianning and Management Planning for the 1987 refueling outage began shortly after the conclusion of the 1985 outage.
Such early planning helped to ensure that critical items were included in the outage work package and that long lead time procurement were initiated to avoid unnecessary impact on the outage schedule. This also smoothed pre-outage schedule development and suppor-ted early identification of safety significant issues. Early and increas-ingly frequent formal outage planning meetings coupled with extensive multi-disciplinary attendance and participation aided in early problem identification and resolution.
These meetings also promoted inter-department cooperation, and a disciplined and cohesive team existed at the commencement of outage activities.
The licensee committed personnel and financial resources to computer-based outage planning.
The detail provided by this system proved to be a key to successful outage planning.
As project needs were identified, requisite plant conditions and materials were determined and expected durations were projected by experienced personnel. This data was incor-porated, with the aid of critical path management sof tware, into a master outage schedule. The schedule provided graphical representations of out-age activities based upon task logic ties, plant conditions and sensitivity
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to impact on the critical path. The schedule was refined in an iterative process based upon technical as well. as management reviews, and reflected the effects of changing project data and shifting priorities. The flexi-bility of the system was tested when senior management determined shortly before the outage that two weeks needed to be trimmed from the schedule and outage commencement was required one week earlier than previously planned.
Due to the extensive pre planning and management familiarity with all aspects of outage activities, these changes were incorporated with minimal impact.
During detailed outage activity reviews, the resident inspector concluded that schedule compression and early commence-
ment had not adversely impacted work quality or proper attention to safety
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significant issues.
i Outage staffing was designed to respond to the increased pace and complex-ity of outage activities.
Operations Department shift staffing was increased. to ensure adequate activity coverage and coordination, and j
maintenance of a safety perspective.
Establishment of an Outage Coordi-l nator early in the planning phase strengthened the scheduling process.
i During the outage, the coordinator provided supervisory level oversight of
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activities, plant evolutions and conditions, and inter-departmental
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liaison.
Augmenting the Outage Coordinator during the outage was a l
Management Representative.
This position was filled on a shift basis by unit department heads and other management level personnel.
This repre-sentative brought a management perspective to outage activities and imple-
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mented problem identification, resolution and expediting activities. The overall staffing plan proved highly effective in ensuring the quality of safety-related activities, Real-time management of outage activities was provided during regularly scheduled twice-daily status meetings. Current project progress as well as an expanded time-base printout of the projected events during a one week window was provided daily to. supervisors.
These meetings were characterized by accurate assessments of work in progress and resolution of conflicts.
Special meetings were scheduled as necessary to focus sufficient and appropriate resources on specific problems.
During these
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meetings the licensee displayed cooperation and a very positive attitude j
toward both nuclear safety and high quality work.
As discussed in Inspection Report 50-245/87-12, Detail 21, the Plant Operations Review
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Committee (PORC) provided excellent oversight of outage activities and
issues. The inspector noted, however, that valuable PORC time was spent reviewing routine procedure changes and other items that could have been accomplished prior to the outage. Although a certain amount of emergent review is expected, efforts should be made to clear routine work prior to outage commencement.
Only isolated instances of less effective control occurred durin'g the l
outage.
These are described in Detail 4 and IR 50-245/87-12, Detail 12.
These appear to be minor perturbations in a successful outage program.
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Management Meetings At periodic irtervals during this inspection, meetings were held with senior plant management to discuss the findings.
No proprietary information was identified as being in the inspection coverage, f
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