IR 05000244/1981008
| ML17258B130 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 06/08/1981 |
| From: | Kister H, Zimmerman R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17258B128 | List: |
| References | |
| 50-244-81-08, 50-244-81-8, NUDOCS 8106230604 | |
| Download: ML17258B130 (14) | |
Text
U.S.
NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region I R
8
. ~50-R 8-08 Docket No.
50-244 50244-810323 50244-810402 50244-810414 License No. DPR-18 Priority Category Licensee:
Rochester Gas and Electric Cor oration 89 East Avenue Rochester, New York 14649 Facility Name:
R.
E. Ginna Nuclear Power Plant Inspection at:
Ontario, New York Inspection conducted:
April 1-30, 1981 Inspectors:
R.
P,
'rman, Senior Resident Inspector I el dat igned date signed
.
date signed Approved by:
H.
B. Kister, hief, Reactor Projects Section 1C, Division of Resident
&
Project Inspection at signed Ins ection Summar
'ns ection on A ril 1-30 1981 Re ort No. 50-244 81-08}
reas ns ecte
outlne, onslte, regu ar, and'ac s
haft jnspection by the resident inspector
ours},'reas inspected included plant operating'ecords; suryeillance testing; IE Bulletin response; operational event followup; periodic and special reports; Licensee Event Reports; and accessible portions of the facility during plant tours, Results:
Of the seven areas inspected; three items of noncompliance were identified sn two areas (Failure to ensure redundant component operability prior to surveillance testing - Paragraph 4; Failure to adequately verify isolation of primary coolant leak-age - Paragraph 5; Failure to make written reports of degraded mode operation -'ara-graph 2}.
8>06gs Oh>~
Region I Form 12 (Rev. April 77)
DETAILS 1.
Persons Contacted The below listed technical and supervisory level personnel were among those contacted:
W. Backus, Operations Supervisor J. Bodine, QC Engineer L. Boutwell, Maintenance Supervisor W. Dillion, Supervisor of Nuclear Security C. Edgar, I & C Supervisor D. Filkens, Supervisor Health Physics and Chemistry D. Gent, Results and Test Supervisor G. Larizza, Technical Engineer R. Morrill, Training Coordinator T. Meyer, Nuclear Engineer J.
C. Noon, Assistant Plant Superintendent C. Peck, Operations Engineer B. Quinn, Health Physicist B. A. Snow, Plant Superintendent S. Spector, Maintenance Engineer The inspector also interviewed and talked with other licensee personnel during the course of the inspection.
2.
Review of Plant 0 erations a.
General-The inspector reviewed plant operations through direct inspection through-out the reporting period.
Activities in progress included routine power operations with coastdown beginning 4/10; recovery from a loss of offsite power on 4/14 (paragraph 6); and a shutdown on 4/18 to commence the annual refueling, modification, and maintenance outage.
b.
Shift Lo s and 0 eratin Records l.
Operating logs and records were reviewed against Technical Speci-fications and administrative procedure requirements.
Included in the review were:
Control Room Log Daily Surveillance Log Shift Supervisor's Log daily during control room surveillance daily during control room surveillance daily during control room surveillance
3.
Plant Recorder Traces daily during control room surveillance Plant Process Computer Printout
-'aily during control room surveillance Station Event Reports 4/1/81 through 4/31/81 The logs and records were reviewed to verify that entries are properly made; entries involving abnormal conditions provide sufficient detail to communicate equipment status, deficiencies, corrective action res-toration and testing; records are being reviewed by management; oper-
. ating orders do not conflict with the Technical Specifications; logs and event reports detail no violations of Technical Specification or reporting requirements; logs and records are maintained in accordance with Technical Specification and administrative procedure requirements.
On two occasions, the inspector noted that the control room log re-quired additional information/clarification concerning off-normal plant events.
In the first instance, 0n April 14 an Unusual Event was declared shortly after a loss of offsite power; however, the control room log entry read that the Unusual Event was declared approximately three hours after the occurrence.
In the second instance, a log entry had not been made to document securing from a local radiation emer-gency which occurred on April 20.
The licensee representative acknowl-edged the inspector's comments and stated that additional effort would be made to maintain accurate information in the control room log.
The inspector observed that the following previously reviewed Station Event Reports, describing conditions leading to operation in a de-graded mode permitted by Technical Specification limiting conditions for operation, were not incorporated into written reports to the NRC Regional Office.
On January 16, 1981, for a period of approximately nine hours, the primary coolant leak detection systems sensitive to radio-activity were not in service.
Operation in this condition is permitted by Technical Specification 3.1.5.3 for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> without requiring further action.
On February 13, 1981, during surveillance testing, the Press-urizer Level Transmitter (LT-428) output signal failed low due to a faulty channel amplifier.
In accordance with Technical Specification 3.5.1, the number of channels to initiate a re-
'actor trip remained in a one-out-of-two instrumentation logic until the failed amplifier was replaced.
Failure to submit reports of the above degraded mode operation with-in thirty days of each respective event is contrary to Technical Spec-ification 6.9.6.2 and is considered an item of noncompliance (81-08-01).
c.
Plant Tour
During the course of the inspection, tours of the following areas were conducted:
Control Room Containment Auxiliary Building Intermediate Building (including control point)
Service Building Turbine Building Diesel Generator Rooms Battery Rooms Screenhouse Yard Area and Perimeter The following observations resulted from the tours:
a
~
b.
Honitorin instrumentation.
Process instruments were observed or corre ation between c annels and for conformance with Technical Specification requirements.
Annunciator alarms.
Various alarm conditions which had been received an ac nowledged were observed.
These were discussed with shift personnel to verify that the reasons for the alarms were understood and corrective action, if required, was being taken.
c ~
d.
e.
Shift mannin
.
Control room and shift manning were observed or conformance with 10 CFR 50.54 (K), Technical Specification, and administrative procedures.
Radiation rotection controls.
Areas observed included control point operation, posting o
radiation and high radiation areas, compliance with Radiation Work Permits and Special Work Permits, personnel monitoring devices being properly worn, and personnel frisking practices.
Equi ment lineu s.
Control board indications for valves and e ectrica rea ers were observed, on a daily basis, to verify that the position or condition required by Technical Specifications and plant lineup procedures was being satisfied.
One item of non-compliance was identified and is discussed in paragraph f.
E ui ment ta in
.
Selected equipment, for which tagging re-quests a
een snitiated, was observed to verify that tags were in place and the equipment in the condition specified.
g.
Fire rotection.
Fire detection and fire fighting equipment was o serve or conformance with Technical Specifications and administrative procedures.
h.
~Securit
.
Areas observed for conformance with regulatory re-quirements, the site security plan and administrative procedures, included vehicle and personnel access, protected and vital area integrity, escort and badging.
Plant housekee in controls.
Plant conditions were observed for con ormance wst a msnsstrative procedures.
Storage of material and components was observed with respect to prevention of fire and safety hazards.
Housekeeping was evaluated with respect.to controlling the spread of surface and airborne contamination.
3.
Ins ector llitnessin of Surveillance Tests a.
The inspector witnessed, the performance of surveillance testing of selected components to verify that the surveillance test procedure was properly approved and in use; test instrumentation required by the procedure was calibrated and in use; Technical Specifications were satisfied prior to removal of the system from service; test was performed by qualified personnel; the procedure was adequately de-tailed to assure performance of a satisfactory surveillance; and test results satisfied the procedural acceptance criteria, or were properly dispositioned.
b.
The inspector witnessed the performance of:
Periodic Test (PT)-22.4, Equipment Hatch Between Door Volume Leak Rate Test, Revision 6, September 8, 1978.
PT-37.2, Containment Vent Mass Air Flow Check, Revision 2, September 29, 1980.
No items of noncompliance were identified.
4.
Effect of Preventive Maintenance on E ui ment 0 erabilit During observation of control board indications on April 7, the inspector noted that the 'B'iesel Generator (D/G) had not been placed in operation when the 'A'/G electrical breaker to Bus 18 was tagged out of service for preventive maintenance.
The function of the 'A'/G is to supply emergency shutdown power to Buses 14 and 18 in the event of loss of all other a.c.
auxiliary power.
Failure to run the 'B'/G continuously with the 'A'/G
s.
unable to perform an intended function, and thus inoperable, is contrary to Technical Specifications 1.4 and 3.7.2.b and is considered an item of noncompliance'(81-08-02).
On-site licensee management had considered the 'A'/G operable based on past operating practice; maintenance was of a preventive rather than corrective nature; and the 'A'/G, itself, was unaffected by the maintenance.
The
'B'/G was immediately started after the inspector informed the licensee repre-sentative of the misinterpretation of the Technical Specification requirement concerning equipment operability.
The following actions have been taken by the licensee to prevent recurrence.
Administrative Procedure-52.4,
"Control of Limiting Conditions for Operating Equipment" has been revised to reflect the definition of op-erability as stated in D. Eisenhut's (NRC) letter to All Power Reactor Licensees, dated April 10, 1980.
Maintenance Procedure-32,
"Use of Circuit Breaker Multi-Amp Test Unit" and Mai ntenance Procedure-32.1,
"DB-25, DB-50 and DB-75 Circuit Breaker Maintenance and Overcurrent Trip Device Test and/or Maintenance" were revised to require the appropriate diesel generator declared inoperable when performing maintenance on the respective bus tie breaker.
Primar Coolant Leaka e Investi ation At 1:50 P.M, 4/10, a '1B'urge Exhaust Fan relay in the Diverse Containment Isolation System failed due to a short circuit, resulting in a blown fuse to the associated relay rack and a subsequent containment isolation.
Upon auto-matic isolation of the Letdown System, the Letdown Relief Valve 203 lifted to the Pressurizer Relief Tank and failed to reseat.
Relief Valve 203 was iso-lated by placing excess letdown in service with normal letdown secured.
The faulty containment isolation signal was removed, and the Letdown Suction Valve 427 and the Letdown Orifice Isolation Valves (200A, 200B and 202), located up-stream of Relief Valve 203, were closed.
Containment particulate activity which increased following the event from a nominal 6.9E-10 uc/cc to 6.3E-8 uc/cc; re-turned to normal levels shortly after shutting the letdown suction and orifice isolation valves.
At 6:20 P.M., 4/10, the licensee initiated a primary coolant leakage inspection of the Letdown System inside Containment to verify system integrity.
The walk-down inspection did not identify the source of the particulate activity.
The
'source of leakage was believed to be from either a reactor coolant pump upper seal or a leak to atmosphere from Relief Valve 203, which is inaccessible during power operation.
On April 15, following review of the above occurrence, the inspector informed on-site licensee management that the leakage investigation had not conclusively indicated that the, leak location had been acceptably isolated.
Basis for the inspector's comment involved a potentially suspect section of Reactor Coolant
7.
System (RCS) pressure boundary piping located between the Letdown Suction Valve 427 and Regenerative Heat Exchanger that had not been verified intact.
A leak in that section of piping would not have been identified through the investigation performed by the licensee.
Further, valve 427., which is designed to fail open, served as the means of isolation between the RCS and the above section of piping.
Valve 427 fails open as a direct result of a loss of in-strument air, and indirectly on a containment isolation signal, which secures instrument air to containment.
The ability to verify that this section of RCS pressure boundary piping was intact was available from the control room.
Fail-ure to ensure that within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of discovery, the leak location had been identified, properly isolated, and no reactor coolant pressure boundary leak-
.
age existed, is contrary to Technical Specification 3.1.5.1, and is considered an item of noncompliance (81-08-03).
At approximately 5:40 P.M., 4/15, the licensee representative, at the in-spector's request, opened the Letdown Suction Valve 427 to ascertain reactor coolant system pressure boundary integrity.
The piping was verified intact based on containment activity remaining at normal levels.
During the present outage, Relief Valve 203 has been replaced.
The original relief valve was disassembled and determined to have stuck open due to mechan-ical binding.
Leakage past a reactor coolant pump upper seal is believed to have been the source of increased containment activity following Letdown System isolation.
No leakage paths to atmosphere were found to exist from Relief Valve 203.
The licensee intends to individually fuse each relay associated with the Diverse Containment Isolation System to preclude a similiar inadvertent iso-.
lation occurrence.
This modification is planned prior to unit startup.
Licensee Event Re orts LER's The inspector reviewed the following LER's to verify that the details of the event were clearly reported, including the accuracy of the description of cause and adequacy of corrective action.
The inspector determined whether further information was required, and whether generic implementations were in-volved.
The inspector also verified that the reporting requirements of Tech-nical Specifications and Station Administrative and Operating Procedures had been met, that appropriate corrective action had been taken, that the event was reviewed by the Plant Operations Review Committee, and that the continued operation of the facility was conducted within the Technical Specification limit.
81-06:
Crack in Immediate Boration Line-March 23, 1981.
During a plant tour an auxiliary operator identified leakage from the immediate borate valve outlet piping.
The line was isolated for repair, with two flow paths from the boric acid tanks to the Reactor Coolant System remaining operable.
Upon further investigation, a pinhole leak was found on the two inch, schedule 10, 304 stainless steel piping near a welded flange.
When in use, the piping
contains12-131 by weight boric acid.
Following use, the line is flushed with reactor makeup water and the line remains stagnant.
The portion of leaking pipe has been removed, and the section forwarded to Westinghouse for metallurgical analysis.
The licensee intends to take all'ecessary corrective actions during the current shutdown, following results of the analysis to determine the failure mechanism.
The inspector will follow the licensee's actions.
81-07:
Loss of Off-Site Power - April 14, 1981.
At 7:39 A.M., 4/14, a loss Wo ttte 34.5 KV supply to the Station Auxiliary Transformer (12) occurred creating a loss of off-site power.
The 4160 volt buses 12A and 12B, and the 480 volt safeguard buses 14, 16, 17 and 18 deenergized.
Both diesel generators started on undervoltage signals and supplied power to the safeguard buses.
The temporary interruption to the power supply (Bus 14) to the 1B Instrument Bus caused numerous alarms and resulted in a turbine runback from 96K to 854 power.
Plant parameters were stabilized and power operation continued while cause of the loss of the 34.5 KY supply was investigated.
Inspection of the Spare Station Auxiliary Transformer (12B) revealed a fault in the 'B'hase vacuum breaker switch.
Cause of the fault is believed to have been a short to ground resulting from rain entering the breaker through a leaking gasket.
The 12B Transformer was isolated from the 34.5 KY bus and power was restored to the 12 Transformer.
The electrical distribution was returned to normal status at 12:17 P.M., 4/14.
The licensee had been maintaining the 34.5 KV line side of the 12B Transformer energized to keep the transformer dry.
Consideration is being given to main-taining the 12B Transformer isolated to prevent a possible recurrence.
Further, an inspection of the 12 Transformer vacuum breaker switches will be conducted dur'ing the current outage.
81-08:
Inoperative Containment Radiation Monitors April 2, 1981.
The cdn-tainment air particulate (R-11) and radiogas (R-12) sample pump failed due to a seized bearing.
The pump was repaired and placed back in service within approximately two hours, Administrative Procedure (A)-52.4 has been revised to require the containment iodine monitor (R-10A) be placed in;operation when R-11 and R-12 are out of service.
Review of Periodic and S ecial Re orts a e Upon receipt, periodic and special reports submitted by the licensee pursuant to Technical Specification 6.9.1 and 6.9.3 were reviewed by the inspector.
This review included the following considerations:
the report included the information required to be reported by NRC require-
9'ents; test results and/or supporting information were consistent with design predictions and performance specifications; planned corrective action was adequate for resolution of identified problems; determination whether any information in the report required class-ification as an abnormal occurrence; and the validity of reported information.
Within the scope of the above, the following periodic report was reviewed by the inspector.
Monthly Operating Report for March, 1981.
No items of noncompliance were identified.
8.
Followu on IE Bulletins IEB)
The inspector reviewed facility records, interviewed licensee personnel and observed facility equipment/components to verify that:
licensee management received and reviewed the bulletins in accor-dance with administrative procedures; information discussed in the licensee.',s bulletin response was accurate>
corrective action was taken as discussed in the reply; and, the licensee.'s response was within the time period required.
IEB 80-06, Engineered Safety Feature (ESF)
Reset Controls The licensee performed a review, at the schematic level, to determine whether
,all ESF equipment remains in the emergency mode following reset of the re-spective actuation signal.
Installed instrumentation and controls were tested in May, 1980 to verify consistency with the schematics reviewed.
Safety Injection Circuit:
~ Actuation of the safety injection reset, switch does not change the state of any equipment, but does permit the operator to place equipment affected by the safety injection signal to the desired position.
Containment Ventilation Isolation Circuit:
Actuation of the containment ventilation isolation reset switch does not change the state of any equip-ment.
The operator must then operate the control module switch/indicator on the containment isolation reset pushbutton panel, on an individual basis, to change equipment position.
Containment Isolation Signal; Following actuation of the containment isolation reset switch, required action is the same as for the Containment Ventilation Circui e
~
Containment Spray Circuit:
Actuation of the containment spray reset switch would permit the operator to place the containment spray pumps and discharge valves in the desired position.
The spray additive tank discharge valves open two minutes after receiving the containment spray actuation signal.
Actuation of the containment spray reset; however, will automatically return the spray additive tank discharge valves to the position called for by their controllers (normally closed).
Emergency Procedure (E)-1.2, Loss of Reactor Coolant, Revision 24, March 13, 1981 precludes resetting containment spray until low level is reached in the spray additive tank; at which time the operator can lineup the contain-ment spray system in the recirculation mode, if necessary.
The inspector considered the procedural control an acceptable means of maintaining equipment control of the spray additive tank discharge valves.
This bulletin is closed.
9. 'xit Interview At periodic intervals during the course of the inspection, meetings were held with senior facility management to discuss the inspection scope and find-'ng l S