IR 05000267/1986006
| ML20205M899 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 03/28/1986 |
| From: | Farrell R, Jaudon J, Snow M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20205M519 | List: |
| References | |
| 50-267-86-06, 50-267-86-6, NUDOCS 8604150461 | |
| Download: ML20205M899 (7) | |
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APPENDIX B U. S. NUCLEAR REGULATORY COMMISSION
REGION IV
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NRC Inspection Report:
50-267/86-06 License: DPR-34 Docket:
50-267 Licensee: Public Service Company of Colorado (PSC)
P. O. Box 840 Denver, Colorado 80201 Facility Name:
Fort St. Vrain Nuclear Generating Station Inspection At:
Fort St. Vrain (FSV) Nuclear Generating Station, Platteville, Colorado Inspection Conducted: February 1 - 28, 1986 Mbte / '
3 ~ [id Inspectors:
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'R. E. Farrell, Senior Resident Inspector (SRI)
Date 0"l67L-AS 5 f
. Project Engineer, Project Section A Cate
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g M{Rea[cto E. Sk Projects Branch Approved:
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f. P/ Jau
,~ Chief, Project Section A Ifate *
Reicto rojects Branch-I kok#IMjCK 61 860409
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05000267
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-2-Inspection Summary Inspection Conducted February 1-28, 1986 (Report 50-267/86-06)
Areas Inspected: Routine, unannounced inspection of operational safety verification, licensee action on previously identified inspection items, and reserve shutdown system.
Results: Within the areas inspected, two violations were identified in one area (failure to conduct a safety evaluation and failure to take prompt corrective action) and no violations or deviations were identified in the remaining two areas.
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DETAILS i
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Persons Contacted Principal Licensee Employees P
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- T. Borst, Support Services Manager / Radiation Protection Manager
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- W. Craine, Superintendent, Maintenance l
- R. Craun, Site Engineering Manager l
- J. Gahm, Manager, Nuclear Production
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- M. Holmes, Manager, Nuclear Services J. Hunter, Shift Supervisor -
- D. Evans, Superintendent, Operations
- H. Ferris, QA Operations Manager
- C. Fuller, Station Manager S. Hofstetter, Nuclear Licensing Engineer
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- F. Novachek, Technical / Administrative Services Manager l
- #*L. Singleton, Manager QA
~ *J. Gramling, Supervisor, Nuclear Licensing i
L M. Kasten, Equipment Operator
- 'D. Warembourg, Manager, Nuclear Engineering The NRC inspectors also contacted other licensee and contractor person-nel during the inspection.
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- Denotes those attending the exit. interview conducted by Mr. M. E. Skow on February 13, 1986.
- Denotes those attending the exit interview conducted by Mr. R. E. Farrell l
on February 28, 1986.
2.
Operational Safety. Verification The NRC inspectors reviewed licensee activities to ascertain that the
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facility is being operated safely and in conformance with regulatory i
requirements and that the licensee's management control system is effectively discharging its responsibilities for continued safe opera-tion.
The review was conducted by direct observation of activities, tours of the facility, interviews and discussions with licensee personnel, independent verifications of safety system status and limiting condi-l i'
tions for operations, and review of facility records.
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Logs and records reviewed-included:
Shift supervisor logs
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Reactor operator logs
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Equipment operator logs
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Auxiliary operator logs
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Technical Specification compliance logs
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Operations order book
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Operations deviations reports
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Clearance log
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Temporary configuration reports
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Station service requests (SSR)
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During tours of accessible areas, particular attention was directed to the following:
Monitoring instrumentation
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Radiation controls
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Housekeeping
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Fluid leaks
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Piping vibrations
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Hanger / seismic restraints
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Clearance tags
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Fire hazards
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Control room manning
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During the course of plant tours, the'NRC inspectors noted a conduit -
containing two instrument air lines connected to a junction box suspended by a rope tied to structural steel. The rope was serving the function of a hanger support and was providing support in the vertical direction l
only. The rope support in question was located on the snubber deck below the. reactor pressure vessel. The instrument air lines and junction box were, upon further investigation, found to be associated with Helium Circulators A and B of Reactor Coolant Loop 1.
Investigation by PSC site nuclear engineering revealed that the air lines were associated with level control switches 21303 and 21305, which actuate the level control valves for the high pressure moisture separa-
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-5-tors associated with these helium circulators. According to PSC's piping and instrumentation Diagram 82-6, these lines are safety-related.
l The rope was replacing support No. N-221 which was a Category I, seismic hanger.
PSC has thus far been unable to determine why, how, or when support N-221 was removed and replaced with a rope. There was no record of a safety evaluation or an authorization for this change found. This is considered to be a violation of NRC regulations' 10 CFR 50.59 (50-267/8606-01).
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The tour of the plant on which the above described rope replacing Support N-221 was identified, took place on the afternoon of Febru-ary 20,1986. The tour was conducted by the NRC SRI, a PSC shift supervisor, one NRC consultant, and four personnel from the Office of,
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Nuclear Reactor Regulation. The NRC personnel requested of the PSC shift supervisor that PSC investigate the matter and advise them by the following day, (February 21,1986) of. actions planned or taken.
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February 21, PSC personnel did not yet have an answer as to the adequacy of the rope replacing the hanger; however, it was indicated that the matter was still under investigation.
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On February 24, 1986, the NRC SRI contacted FSV plant management to determine the status of the investigation into the rope.
It was deter-
, ' mined that the investigation had been dropped without resolution and that the rope was still in place supporting the safety-related instru-ment air tubing. Additionally, a documentation search by PSC site nuclear engineering, in response to the NRC SRI's question as to why this rope had not been identified during the course of environmental qualifi-
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cation walkdowns, produced documentation indicating that the rope had indeed been identified during an EQ walkdown in November of 1985. The
plant was critical when the~ rope was identified by NRC personnel, and
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had been known to PSC since November 1985.
The absence of a prompt investigation and resolution of questions regarding the operability of this rope as a safety-related instrument tubing support is an apparent violation of NRC regulations-(50-267/8606-02).
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_The licensee, on the evening of February 24, 1986, installed an engi-
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neered support replacing the rope. The NRC SRI has inspected this support which appears to adequately support the ' instrument air lines and instrument junction t,ox previously supported by the. rope. Corrective action actively continues in the snubber deck area. The licensee has conducted independent inspections in the snubber deck area and identi-
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fied numerous deficiencies which are now being corrected. Many of these
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deficiencies are in the area of_ housekeeping, some involve insulation,
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and at least one involves an inadequate conduit support. The NRC SRI will continue to monitor these activities.
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Du: ing the' inspection period, the NRC SRI observed a plant startup authorized under the Commission order allo"ing 35% of full power opera-tion through May 31, 1986. On February 14,1986, at 1157 hours0.0134 days <br />0.321 hours <br />0.00191 weeks <br />4.402385e-4 months <br /> MST with i
a core _ outlet' temperature of 132 F, a reactor pressure vessel pressure
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of 230 pounds per square inch absolute'(psia),150 parts per million moisture content in the reactor coolant which corresponds to a dew point l
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-6-of 13.2 F and 8.9% of maximum reactor coolant flow, the reactor opera-tors commenced operations to take the plant critical. The reactor was taken critical at 1330 hours0.0154 days <br />0.369 hours <br />0.0022 weeks <br />5.06065e-4 months <br /> MST after withdrawing control rod groups 2A, 48, 4D, and 4F. The critical rod group was 2B. The reactor went critical at 63 inches on group 28. The reactor engineers criticality prediction was 50 to 81 inches out on the critical group. Criticality was well within the prediction. At 1330 hours0.0154 days <br />0.369 hours <br />0.0022 weeks <br />5.06065e-4 months <br /> MST the startup channels indicated 2000 counts per second with a core outlet temperature 135 F, reactor pressure vessel pressure of 229 psia and a 161 parts per million reactor coolant moisture content corresponding to a dew point of 13.4 F.
This was the first time in 1986 that the reactor was taken critical.
The NRC SRI noted that the startup was trouble free, the plant manager and operations superintendent were in the control room observing, and a quality assurance supervisor was also present to observe the startup.
l The SRI noted that access of personnel into the control room has been tightened, consistent with NRC recommendations and that plant housekeep-
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ing has been consistently improving in recent months.
An NRC inspector observation (documented in NRC Inspection Report 50-267/86-03) that compressed gas bottles in the plant were not always l
properly secured was monitored during this reporting period. During l
this inspection period, improvement was noted. Compressed gas bottles
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are secured properly with chains with only occasional lapses.
Plant operations personnel ~ appear to be monitoring this closely and are sensitive to the hazard presented by improperly secured. compressed gas bottles.
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Licensee Action on Previously Identified Inspection Items (0 pen) Violation 50-267/8531-01: System Lineup Procedures. The viola-tion stated that the diesel generator procedures did not include valve positions for some valves. The licensee has taken some corrective action.
Procedure S0P 92-04, Section 2.2, "Startup Check List for Automatic Operation," instructs the operators to confirm the availability
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of service water, to check the lube oil level, and to verify that the I
engine water jacket heaters are operating. The corrective action still l
does not address engine water jacket. heaters, nor does it specify what valves, plugs or similar components to check in verifying system avail-f ability. By extension, it also does not provide a means for the opera-tor to document that all components were checked. The procedure still relies on the operator's familiarity.with the system.
In addition, the corrective action does not address independent verification. This
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violation remains open.
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Reserve Shutdown System'
l The objective of the inspection was to determine whether the licensee
has satisfactory procedures for assuring that the reserve shutdown i
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c system will perfonn its intended function if required, and to-provide
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_ assurance that the licensee has satisfactorily implemented those proce-!'
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dures.
l The NRC inspector reviewed the following surveillance procedures:
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l Procedure Number'
Date Accomplished l
SR 4.1.9.D.1-R 7/19/85
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SR 4.1.8.C.-1/2/3-Q-1/29/86 10/31/85
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SR 4.1.8.A/B-W/4.1.9.A/8-W 2/9/86 2/3/86
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'1/26/86 1/19/86 1/11/86 1/7/86 The'NRC inspector reviewed the procedures to determine.whether they met the. requirements of the Technical Specifications, included adequate acceptance criteria, and specified appropriate handling requirements and test conditions.
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No violations or-deviations we'reinot'ed.-
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5.
Exit interviews were held on February 13 and 28,1986, with those
- personnel denoted in paragraph 1 of this report. At the meeting, the scope of the inspection and findings were sununarized.
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