IR 05000400/1987040
| ML18005A265 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 01/28/1988 |
| From: | Burris S, Fredrickson P, Maxwell G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18005A263 | List: |
| References | |
| 50-400-87-40, IEB-87-002, IEB-87-2, NUDOCS 8802080271 | |
| Download: ML18005A265 (10) | |
Text
~p,lt RE0I (4 P0 cs
'p***+
UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323 January 29, 1988 Report No.:
50-400/87-40 Licensee:
Carolina Power and Light Company P.
O.
Box 1551 Raleigh, NC 27602 Docket No.:
50-400 Facility Name:
Shearon Harris
Inspection Conducted; October 27 - November 27, 1987 Inspectors:
xwe Approved by:
re r>c son, ect>on le Division of Reactor Projects License No.:
NPF-63 ate gne ate igne te gne SUMMARY Scope:
This routine, announced inspection involved inspection in the areas of Followup on Items of Noncompliance, IE Bulletins, Operational Safety Verifica-tion, Monthly Surveillance Observation, and Monthly Maintenance Observation.
Results:
Three violations were identified,
"Repeat Violations of Breach of Containment Integrity" - Paragraph 5. b, "Failure to Follow Operations Proce-dures"
- Paragraphs 5.c.(1)
and (2),
and "Failure to Take Prompt Corrective Action on Conditions Adverse to Quality" - Paragraph 5. d.
Additional examples of a previous violations were identified in paragraph 3,
0g29 880208 ocy, 050oPDR 027>
8800O+OO
@DR AD
REPORT DETAILS Persons Contacted Licensee Employees G.
G. Campbell, Manager of Maintenance J.
M. Collins, Manager, Operations G.
L. Forehand, Director, QA/gC L. I. Loflin, Manager, Harris Plant Engineering Support G.
A. Myer, General Manager, Milestone Completion D.
L. Tibbitts, Director, Regulatory Compliance R.
B.
Van Metre, Manager, Harris Plant Technical Support R.
A. Watson, Vice President, Harris Nuclear Project J.
L. Willis, Plant General Manager, Operations Other licensee employees contacted included technicians, operators, mechanics, security force members, engineering personnel and office personnel.
Exit Interview The inspection scope and findings were summarized on November 30, 1987, with the Plant General Manager, Operations.
No written material was provided to the licensee by the resident inspectors during this reporting period.
The licensee did not identify as proprietary any of the materials provided to or reviewed by the resident inspectors during this inspection.
The violations identified in this report have been discussed in detail with the licensee.
The licensee provided no dissenting information at the exit meeting.
Followup on Items of Noncompliance (92702)
(Open) Violation 400/87-31-02
"Failure to Report an ESF Actuation Within Four Hours."
This violation identified that the licensee had failed to report under
CFR 50. 72 that the Auxiliary Feedwater system sustained an automatic initiation on August 5, 1987.
Subsequent to the issuance of this violation, the inspectors identified three other instances where the licensee did not appropriately report plant events.
These three events were reported under
CFR 50.73 as License Event Report (LER)87-034, 87-052 and 87-055 and are discussed in paragraph 5.b.
As these events appear to be similar in nature and the corrective action from violation 400/87-31-02 was not complete when the last event occurred, no additional violation is being issued; however, these events are being considered as additional examples of violation 400/87-31-02.
(Withdrawn)
Violation 400/87-37-02
"AFW Logic Design Deficiency".
Region II management has been informed that a determination has been made by the NRC Office of General Counsel that a previously-identified
violation,
"AFM Logic Design Deficiency",
was not appropriate.
This decision was made on the basis of recent interpretations from related cases since Appendix A design criteria are considered to be guidance.
This item is withdrawn.
Our records wi 11 be adjusted to reflect this decision due to the enforcement decision on this matter, the enforcement conference scheduled for November 20, 1987, was cancelled.
4.
Inspection and Enforcement (IE) Bulletins (92703)
(Open)
"Fastener Testing to Determine Conformance with Applicable Material Specifications".
During the week of November l5, 1987 the licensee received IE Bulletin 87-02, which required that the licensee perform verification testing on the identified material currently in their storage facilities.
The licensee implemented the recommended proportional sampling criteria outlined in the Bulletin and-have sent these samples to the CP8 L Energy and Environmental Center for composition and identifica-tion testing.
Verification of the method of sampling was confirmed by the on-site NRC personnel with no discrepancies noted.
The inspectors will follow-up on this item in a future inspection period.
5.
Operational Safety Verification (71707, 71710)
a.
Plant Tours The inspectors conducted routine plant tours during this inspection period to verify that the licensee's requirements and commitments were being implemented.
These tours were performed to verify that systems, valves and breakers required for safe plant operations were in their correct position; fire protection equipment, spare equipment and materials were being maintained and stored properly; plant operators were aware of the current plant status; plant operations personnel were documenting the status of out-of-service equipment; security and health physics controls were being implemented as required by procedures; there were no undocumented cases of unusual fluid leaks, piping vibration, abnormal hanger or seismic restraint movements; and all reviewed equipment requiring calibration was current.
Tours of the plant included review of site documentation and inter-views with plant personnel.
The inspectors reviewed the shift foreman's log, control room operator's log, clearance center tag out logs, system status logs, chemistry and health physics logs, and control status board.
During these tours the inspectors noted that the operators appeared to be alert and aware of changing plant conditions.
The inspectors evaluated operations shift turnovers and attended shift briefings.
They observed that the briefings and turnovers provided sufficient detail for the next shift cre The
'inspectors.verified that various plant spaces were not in a condition which would degrade the performance capabilities of any required system or component.
This inspection included checking the condition of electrical cabinets to ensure that they were free of foreign and loose debris, or material.
Site security was evaluated by observing personnel in the protected and vital areas to ensure that these persons had the proper authori-zation to be in the respective areas. 'he security personnel appeared to be alert and attentive to their duties and those officers performing personnel and vehicular searches were thorough and syste-matic.
Responses to security alarm conditions appeared to be prompt and adequate.
b.
Breach of Containment Building Integrity On September 17, while the plant was operating in Hot Shutdown, plant health physics personnel breached the containment building integrity.
The event occurred at 2: 15 p.m.
and lasted for about two minutes and resulted from both the inner and outer personnel access doors being opened at the same time.'he inspectors interviewed plant personnel and evaluated the licensee's documentation related to the event; as a
result the following information was noted:
(1)
The Event was documented by the licensee on Licensee Event Report (LER)87-055 pursuant to 10 CFR 50. 73 (a).(2).(i).
(2)
The Event was not reported by the licensee by telephone to the NRC Operations Center Duty Officer.
CFR 50. 72. b.(2).(iii)
requires the licensee to report via the emergency notification system, within four hours, any event or condition that could have prevented the fulfillment of the safety function of a structure needed to mitigate the consequences of an accident.
(3)
The breach of integrity occurred when personnel opened the inner containment access door when the outer door was not fully closed; this required the personnel to override an interlock.
(4)
The cause of the event was attributed to personnel not under-standing the access door air lock indicating system.
Personnel opened the inner door even though the indication light for the outer door did not indicate that the outer door was closed, i.e.
the closed light was not lit.
(5)
Breaching containment integrity while operating in Hot Shutdown is a violation of Technical Specification (TS) 3.6. 1.3.a which requires that an operable personnel access air lock door be closed or locked closed when the other door is inoperabl Further review revealed that on August 31, while the p'1ant was operating in Hot Standby, the outer containment building personnel access door was declared inoperable, due to its 0-ring seals. becoming loose and falling out of its grooves.
Even though the shift foreman announced the inoperable status of the doors over the site public announcement system, two personnel opened the inner door when the outer door was inoperable.
This event.lasted. for about one minute and resulted in a violation of containment integrity requirements of TS 3.6. 1.3, Action Item a.,
The event was not reported by telephone to the NRC Operations Center Duty Officer, as required by 10 CFR 50. 72. b.(2).(iii); the event was documented on LER 87-052.
Both of the preceding events (LERs87-052 and 87-055) occurred within
days of each other.
The licensee stated that the corrective action for LER 87-052 was not in full effect until after the event documented on LER 87-055 had occurred.
In addition to the corrective action documented by the licensee on the above LERs, plant management issued a Plant Special Order 87-003 on October 23, 1987.
The order mandates that only specified personnel are allowed to operate the containment access doors.
Those personnel were provided detailed qualification training on the correct door operation and the govern-ing procedures.
Each plant operations shift has been required to have at least two qualified door operators.
,The door operators'ill be supplied by the plant maintenance organization.
Mhen a contain-ment entry is made, the qualified door operator is required to operate the access doors and remain on station until all personnel have exited.
A previous violation (50-400/87-21-01)
identified similar instances where personnel entered the containment building on June ll, 1987, through the inner personnel access door subsequent to the outer door being declared inoperable.
The licensee also failed to report this condition to the NRC Duty Officer in accordance with 10 CFR 50.72.-
b.2.(iii) but did document the event under LER 87-034.
The inspectors evaluated the licensee's response to violation 50-400/
87-21-01 and found that the "corrective steps taken to avoid further noncompliance" failed to prevent repeated violations of TS 3.6. 1.3.a, as demonstrated by the events described above.
Failure to take corrective action to prevent the repeated violations of TS 3.6. 1.3.a is a violation of 10 CFR 50, Appendix B, Criterion XVI, FSAR Section 17. 2. 16 and CP8L Corporate gA Program Section 15. 2. 5, which collectively require corrective action to preclude recurrence of significant conditions adverse to quality.
This violation will be identified as
"Repeat Violations of Breach of Containment Integrity",
50-400/87-40-0 Reactor Trips Due to Operator Error (1)
On November 7, with the plant at about three and one half percent power, approximately 1000 pounds steam header pressure, main steam isolation bypass valves open equalizing pressure across the main steam isolation valves (MSIV) and the steam dump control (SDC)
system in pressure mode control with the cont-
. roller set at 74K (960 pounds),
the plant experienced a reactor trip/safety injection.
Operations'ersonnel started opening the 1A steam generator MSIV when the pressure drop across the 1A MSIV reached the minimum required pressure after signing the procedural step which acknowledged that the SDC system cont-roller was set at 84K (1092 pounds) control.
Within one minute operations personnel noted a "swell" condition in the lA steam generator which indicated that the level was increasing.
After approximately another minute the operators reported that they heard steam flow increasing and decreasing rapidly, indicating main steam header pressure/flow fluctuations.
Since the SDC was in the pressure mode of control when the MSIV started opening, the SDC system opened the steam dump valves in response to the increased steam pressure.
In addition to the pressure control function, as sensed on the common header, interlock P-12 (low-low TAVG 553 degrees F.) would shut these steam dump valves if the average coolant temperature dropped to 553 degrees F.
With the steam dump valves open removing heat from the secondary system, primary temperature was lowered to the P-12 setpoint which automatically shut the Group 1 and 2 steam dump valves.
Pressure and temperature increased to the point at which the steam dump valves reopened with a subsequent closing due to a decrease in the primary temperature.
Opening and closing osci llations occurred three times until the pressure in the common header decreased to the low pressure safety injection (SI)
and reactor trip setpoint, The SI was secured in approximately 12 minutes after injecting 2925 gallons of water to the reactor coolant system.
Pressurizer level increased from approximately 25K to 55K.
The inspector determined that the licensee properly reacted to this event during the restoration of the unit to stable plant conditions.
This event was caused by the operator's failure to correctly set or verify that the steam pressure controller was set at the 84K control setting, as required by Operations Procedure (OP)-126 Rev.
2, Hain Steam, Extraction Steam and Steam Dump Systems.
The inspectors were informed that although the operator signed the procedural step which set or verified the correct setting of the steam dump pressure controller, the operator apparently misread the 74K setting to be 84K.
Failure to have the steam dump control system set to the proper steam header pressure resulted in the subsequent low pressure SI and reactor trip.
This item is one example of a violation for failure to establish and implement procedures for activities affecting quality,
"Failure to Follow Operations Procedures",
50-400/87-40-0 (2)
On November 8, with the plant operating in Node 1 at 22X power, the licensee manually tripped the reactor due to loss of the running main feedwater pump.
Prior to the event the licensee had been using the condensate system to clean up the feedwater which required that the recirculation valve from the condensate pump discharge (1CE-293)
be placed in the full open position versus the modulate position required during power operations.
With recirculation valve 1CE-293 fully open at this power level, the condensate pump was operating near run-out capacity.
Therefore, when the operator increased the loading requirements on the turbine generator, it created an increase'eed flow demand on the feedwater and condensate systems.
With the condensate system discharge pressure near its low trip setpoint prior to the increased feed demand, when the discharge pressure dropped to the low trip setpoint the condensate pump tripped, which also tripped the condensate booster pump and running main feedwater pump on electrical interlocks.
When these feedwater system pumps tripped, the operators manually tripped the reactor in accordance with their emergency procedures.
The licensee reviewed the circumstances leading up to the trip and found that the operator should have placed condensate recirculation valve 1CE-293 back in the modulate mode of control when recirculation clean up was terminated.
The licensee's procedure, OP-134, Rev.
2, Condensate System, had the operator place valve 1CE-293 in full open but never stipulated to return the valve to its normal at-power lineup.
This inadequate procedure is an additional example of violation, "Failure to Follow Operations Procedures",
50-400/87-40-02.
d.
Reactor Coolant System Vent (RCSV) Valves As discussed in Inspection Report 50-400/87-37, an event on October 9, 1987, involving RCSV valves resulted in a loss of RCS inventory.
Continuing the analysis of this event, the inspectors reviewed the historical records associated with the RCSV valves and have determined the the following sequence of events occurred:
In late 1981 the licensee's H.
B. Robinson facility experienced problems with spurious opening of Target Rock RCSV valves.
The vendor subsequently recommended that the Robinson site modify the system by inverting the valves, which was done, thereby eliminating the problem.
In late 1984 Target Rock generated an internal memorandum discussing the spurious opening phenomena, spurious openings of these valves are described in more details in a December 1980 Target Rock Report ¹2866, and a 1981 ASME pape In mid-to-late 1985 RCSV valves experienced reverse flow problems while the licensee was conducting reactor coolant system fill and venting evolutions during the preoperational testing phase at the Harris Plant.
In early 1986 the licensee's Joint Test Group (JTG) met to discuss the preoperational test results associated with the reverse flow problems, and the JTG voted unanimously to accept these test results with the exception of a test exception concerning addition of a precaution in General Procedure GP-008 regarding the inability of Target Rock valves to isolate the RCS from pressurizer relief tank pressure.
Following the commencement of power range testing activities in ear ly 1987, the licensee experienced problems with RCSV valves spuriously opening while opening RC 900, 901 and 904.
The licensee generated a
Work Request to check valve position indication and control circuits (MR-87-AGDU1 issued on February 27, 1987),
but did not declare any of the malfunc-tioning valves as inoperable.
In April 1987 the work requested by WR-87-AGDU1 was accomplished with no electrical or control deficiencies identified.
Licensee personnel issued WR-87-AGDU2 on April 10, 1987, requesting mechanical assistance in reworking these valves.
Over the next six months the RCSV valves were tested in accor-dance with Maintenance Surveillance Test (MST) requirements with no further problems being identified.
However, on October 9, 1987, while conducting the RCSV valve NST, the valves operated improperly, discharging water inventory to the containment'nd pressurizer relief tank.
The licensee issued a new Work Request 87-BSFAl implementing Plant Change Request PCR-2397 and cancell-ing WR-87-AGDU2.
Again, none of the valves were declared inoperable.
In October 1987 the licensee implemented PCR-2397 during its planned outage.
This PCR disassembled and inspected valves RC-901 and 903 for damage and reassembled these valves.
Addi-tionally the PCR inverted valves RC-904 and 905 and installed reverse flow check valves.
These modifications should eliminate future spurious opening problems similar to those observed.
Based on the extended time from or'iginal identification to the final resolution of the spurious opening problem, it is evident that the problem did not receive prompt attention and corrective action.
This is based on the previously identified chronology of events, in which all phases of Shearon Harris plant management were aware of the specific problems with RCSV valves as early as June 1985, for the reverse flow problem and February 1987 for the valve configuration
problem.
Although the 1981 spurious opening event occurred at another site, it was a
CP8L facility and, as such, the information should have been disseminated to Shearon Harris.
This item will be identified as a violation, "Failure to Take Prompt Corrective Action on Conditions Adverse to Quality", 50-400/87-40-03.
As the failure to properly declare the spurious operating valves as inoperable appears to be interrelated with the failure to take prompt corrective action on these problems, a separate operability violation is not being issued.
Three violations were identified in the areas inspected.
6.
Monthly Maintenance Observation (62703, 62700, 37700)
The inspectors reviewed the licensee's maintenance activities during this inspection period to verify the following:
maintenance personnel were obtaining the appropriate tag out and clearance approvals prior to commencing work activities, correct documentation was available for all requested parts and material prior to use, procedures were available and adequate for the work being conducted, maintenance personnel performing work activities were qualified to accomplish these tasks no maintenance activities reviewed were violating any limiting conditions for operation during the specific evolutions; the required QA/QC reviews and QC hold points were implemented; post-maintenance testing activities were completed, and equipment, was properly returned fo service after the completion of work activities.
The following activities were evaluated during the inspectors'outine monthly maintenance observations:
Fire protection maintenance activity was accomplished in accordance with FPT-3003, Rev.
1, Attachment 2, Page 6 of 16.
This is an annual test of all fire main isolation valves.
RSCV valves were tested following rework which was authorized by PCR-2397.
OST-1043 was completed with the reactor system in Hot Shutdown (Mode 4) at approximately 260 degrees F. at 330 psig.
MST-10045, Rev.
1, was conducted on excore nuclear instrument system N-42.
The MST was a calibration of the nuclear instrument system authorized by WR 87-BFFF1.
No violations or deviations were identified in the areas inspected.