IR 05000295/1981009
| ML20009F948 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 06/18/1981 |
| From: | Hayes D, Kohler J, Waters J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20009F937 | List: |
| References | |
| 50-295-81-09, 50-295-81-9, 50-304-81-05, 50-304-81-5, NUDOCS 8108030186 | |
| Download: ML20009F948 (18) | |
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U.S. NUCLEAR REGULATORY COMMISSION OFF1CE OF INSPECTION AND ENFORCEMENT
REGION III
Report No.
50-295/81-09; 50-304/81-05 Docket No.
50-295/304 License No. DPR-39, DPR-48 Licensee:
Commonwealth Edison Company P. O. Box 767 Chicago, IL 60690 Facility Name:
Zien Nuclear Power Station, Units 1 & 2 Inspection At:
Zion, IL Inenection :onducLJ:
April 16, 1981 through May 29, 1981 c
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Inspectors:
J. E. Kohler
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Waters i
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A Approved By:
M. Hayes', C ef Reactor Projec s Section 1B
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Inspection Summary Inspection on April 16-May 29, 1981 (Report No. 50-295/81-09; 50-304/81-05)
Areas Inspected: Routine unannounced resident inspection of licensee actions on previous items, reactor operations, operator logs, operational safety verification, monthly maintenance observation, monthly surveillance observation, LER follow-up, instrument inverter failures, loss of heat tracing channels.
Unit 1 primary to secondary leakage, part 21 notification on volume control tank level, radiation chemistry foreman coverage, radioactive releases, diesel generator operability with less than twelve cylinders, TMI modifica-tions, IE Bulletin followup. The inspection involved 404 hours0.00468 days <br />0.112 hours <br />6.679894e-4 weeks <br />1.53722e-4 months <br /> onsite by two NRC inspectors including 41 hours4.74537e-4 days <br />0.0114 hours <br />6.779101e-5 weeks <br />1.56005e-5 months <br /> on site during off shifts.
Results: Of the areas inspected, no items of noncompliance were identified.
8108030186 810720 PDR ADOCK 05000
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1.
Persons Contacted K. Graesser, Station Superintendent
- L. Soth, Operating Assistant Superintendent
- C. P11ml, Administrative and Support Services Assistant Superintendent E. Fuerst, Unit 1 Operating Engineer
- J. Gilmore, Unit 2 Operating Engineer R. Budowle, Assistant Technical Staff Supervisor D. Howard, Rad-Chem Supervisor J. Marianyi, Operating Engineer R. Shannon, ISI Co-ordinator F. Ost, Health Physics Engineer
- B. Harl, Quality Assurance Engineer
- L. Pruett, Technical Staff Engineer
- B. Choate, Technical Staff Technician B. Schramer, Rad-Chem Engineer
- Denotes those present at management exit of May 29, 1981.
2.
Licensee Action on Previous Inspection Finding (closed) Unresolved Item (295/80-17-05): Review of Radiation Monitors Taken Out of Service. The licensee was requested to analyse the effects of taking various radiation monitors out of service when these monitors had specific control functions.
"he licensee responded with a review of all existing radiation monitors that have control fur :tions and the ef fect on plant safety if these monitors are out of service.
In cases where the monitors are required to stop radioactive releases to the environment, these monitors must by procedure be in service during the release or the release must be terminated. Other monitors with control functions could be supplemented by shiftly grab samples or monitors that alert the operator of abnormal radioactive conditions. In these cases manual action would be taken by the operator.
This item is closed and no items of noncompliance were identified.
3.
Summary of Operations Unit 1 The unit was made critical April 16, 1981 following a ninety-one day refueling outage. After low power physics testing the turbine was latched April 18, 1981.
Upon rolling the turbine up to 1800 rpm the No. 7 bearing failed. This was attributed to a J oose nut on the coupling immediately adjacent to the bearing.
The turbine was tripped in response to the bearing failure. While in hot standby the reactor tripped on April 18, 1981 due to low-low level in the IB steam generator. The low-low level resulted from difficulty in controlling level at low power and small moderator temperature coefficient. Bearing repairs were completed and the unit was made critical and tied to the grid April 22, 1981.
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A Additional unscheduled shutdowns and load reductions were as follows:
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a.
On April 26, 1981 the unit tripped from 81% power on an OP AT signal.
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Instrument mechanics had been resetting power range high flux trip points and in doing so induced an OPo T trip signal by decreasing the OPA T set point. The trip signal was reset but came back in a few seconds later. The mechanics moved on t3 another channel and once again the OPS T set point was reduced by increasing the NI
signal.
Since the trip signal still remained from the first in-strument the 2/4 logic for OP4iT was satisfied and the unit tripped.
The unit was made critical and restored to the grid April 27, 1981.
b.
Power was ramped down and the unit placed in hot standby May 2, 1981 for turbine balancing to correct high vibration readings. Numerous weight adjustments were required. The unit was restored to the grid
May 5, 1981.
On May 9, 1981 the unit was again removed from the grid and placed in c.
hot standby for turbine balancing. The weight adjustments were com-l i
pleted and the unit was restored to the grid the sa a day, d.
The licensee discovered both channels of heat tracing failed on a section of baron 1ijection tank piping on May 27, 1981.
(See para-
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graph 7 for details of the heat trace problem.) Power reduction i
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from 100% was commenced in accordance with technical specifications j
limiting condition for operation. The heat tracing was replaced and the power reduction stopped at approximately 46%. The unit was sub-l sequently returned to full power.
e.
On May 28, 1981 the unit tripped from 100% power and the safety in-jection systems activated. This resulted from loss of an instrument
bus coincident with tripped OP4sT and OTA T channels due to instru:nent
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maintenance.
(See paragraph 4 for details.) The unit was made critical
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and restored to the grid May 29, 1981.
Unit 2 The unit operated at power levels up to 93%. Reactor power was limited because of isolation of a low pressure feedwater heater string due to tube leakage.
One unscheduled reactor trip occurred: On May 7, 1981 the unit tripped from 92% power and the safety injection systems activated. This resulted from the loss of an instrument bus.
(See paragraph 5 for details). The unit was made critical May 8, 1981 and restored to the grid May 9, 1981.
4.
Zion Unit 1 Safety Injection At approximately 0847 on May 28, 1961 Zion Unit 1 tripped from 100% power and a safety injection signal was experienced. The safety injection re-sulted in ECCS equipment starting and the injection of the Boron Injeccion-3-
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Tank. The safety injection was caused by high steam flow coincident with low steam line pressure. However, the low steam pressure indication was the direct result of the loss of an inverter to instrument bus 114. The reactor did not experience a pipe break and all vital system parameters remained nominal. Resident inspectors were in the control room during the recovery.
The safety injection with boron injection tank injection was reset when vital parameters were verified to be normal.
The licensee investigated the cause of the reactor trip. The trip occurred before the safety injection and computer first out indicated over powersS T i
and over temperature /S T in two out of four loops. No nuclear safety para-meters were exceeded. The cause of the over power /1 T and over temperature
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j[ T reactor signals being developed is as follows:
Instrument mechanics were working in loop C inputing new 100% loopsi T's based on the recent startup
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l from refueling. In order to do this, bistables were tripped on loop C OPd T channels. During this work in progress, the plant experienced a loss of in-
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strument inverter supply to bus 114. Loas of the inverter developed OPdiT and OT/AT signals in another loop. The plant now had OPdLT and OTliT signals
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in two out of four loops and experienced a reactor trip as designed. The inverter failure also developed low steam line pressure signals in two out of four loops.
The reactor trip caused the high steam flow alarm-because the set point for j
high steam flow drops to a zero power setpoint immediately following the trip.
j Steam flow dropped off as a result of the trip but is always above the zero power setpoint initially.
The high steam flow coincident with the low steam
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pressure from the inverter failure resulted in safety ejection. No water
hammers or pipe damage were reported.
(See paragraph 6, Instrument Bus Inverter Failures, for the specific areas receiving further station review'.)
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5.
Zion Unit 2 Safety Injection i
l At approximately 1600 on May 7, 1981 Zion Unit 2 tripped from 92% power and i
a sa ety injection signal was experienced. The safety injection resulted in r
ECCS equipment starting and the injection of the Boron Injtetion Tank. The safety injection was caused by high steam flow coincident with low steam line pressure. However, the low steam pressure indication was the direct result of the loss of an inverter to instrument bus 213, two seconds before the safety injection.
The reactor did not experience a pipe break and all vital system parameters remained nominal. Resident inspectors were in the control room during the recovery. The safety injection with boron injection tank was permitted to run for ten minutes before it was terminated. Pressurizer level reached 65%.
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The licensee inve;tigated the cause of the reactor trip. The trip occurred before the safety injection and computer first out indicated low steam generator level and a steam flow feed flow mismatch on A steam generator. It was determined that instrument bus 213 also fed some feedwater pump control circuits which caused the C feedpump to go to idle. When the feedpump went to idle, a steam flow feedflow mismatch resulted. The loss of instrument bus 213 also resulted-4-L J
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in tripping bistables for low steam generator A level. The net result was a reactor trip due to steam flow feed flow mismatch coincident with low steam level. The reactor trip caused the high steam flow alarm because the set point for high steam flow drops to 40% at zero power.
Steam flow was dropping off as a result of the trip but is always above the zero power setpoint initially.
The high steam flow coincident with the low steam pressure trip from the in-verter failure resulted in a safety injection. No water hammers or pipe damage was reported.
(See paragraph 6, Instrument Bus Inverter Failures, for the specific areas receiving further station review.
6.
Instrument Bus Inverter Failures During this inspection interval, each of the units experienced an instrument invertet failure which led, in both cases, to a reactor trip and safety in-jection. Both inverter failureq were caused by transformer failures internal to the inverter.
Following each event, a station review was conducted prior to start up in order to address the actual plant response and the underlying cause of the event.
Based on this review, it was concluded that four specific items would be in-vestigated in detail la order to determine whether any plant modifications would be appropriate:
a.
Performance of the inverters, particularly the Sola Transformers which are part of the inverters.
b.
The safeguards logic which has two separate loop low steam pressure bistables on each of the four inverters.
Instrument bus power for the feedwater pump control circuits.
c.
d.
NRC criteria regarding automatic switchover to dirty power following failure of an inverter.
This item is designated Open Item 295/81-09-01, 304/81-05-01.
7.
Loss of Two Channels of Ileat Trace, While operating at 100% power on May 27, 1981 Zion Unit 1 entered a limiting condition for operation when technicians trouble shooting an alarm signaling a failure in one of two redundant heat tracing circuits of the Boron Injection Tank piping, discovered that the redundant heat trace circuit, while not alarmed, was drawing insufficient current to be considered operable. Since both circuits were inoperable, the plant was required to be in a hot shutdown condition within four hours of discovery.
Realistically this meant beginning a power reduction of about 1% per minute within two hours of discovery so
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that the plant can be brought off line without being tripped.
While technicians reported with reasonable confidence that at least one hect trace circuit could be made operable within four hours so that a total plant-5-
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shutdown could be avoided, the plant reduced load to 46% during the first four hours of the shutdown.
Subsequently the shutdown was stopped at plus three hours into the event because the heat trace was repaired, and a power ascent ton was begun.
No items of noncompliance were identified.
8.
Unit 1 Primary to Secondary Leakage Primary to secondary leakage was discovered on the 1B and 1C steam generators prior to the Unit 1 refueling outage. Eddy current testing was performed and tubes were plugged in the 1B and 1C steam generators.
Inspection Reports
50-295/81-01 and 50-304/81-01 Section 5 provides details of these evolutions.
Since the return of Unit 1 to power operation on April 23, 1981, primary to
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secondary leakage has again been seen on the 1B steam generator. By securing
blowdown and monitoring build up rate of Iodine-131 in the steam generator, the licensee is able to calculate the leak rate as follows:
5-12-81 4.2 gal / day 5-15-81 14.6 gal / day 5-21-31 32.2 gal / day
5-26-81 24.5 gal / day
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l 6-2-81 14.6 gal / day
j The pre-outage leak rate was calculated to be 45 gal / day and the technical specification limit is 500 gal / day. The equilibrium gross activity with
continuous blowdown is between 10-5 and 10-6 uci/ml. No detectable activity has been found down stream of the blow down demineralizers or in the condensate
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i systems. The licensee is continuing to monitor the leak rate and barring any significant increase plans no further action until the next Unit 1 refueling outage during March of 1982.
No items of noncompliance were identified.
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Part 21 Notification on Volume Control Tank Level On bby 21,1981 Inspection and Enforcement Headquarters was notified of a potential for adverse control and protection system interaction in certain
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Westinghouse plants. In the absence of proper operator action, a single random failure in the volume control tank (VCT) level control system could adversely affect the high head safety injection system for certain Westinghouse designed plants. A failure of the VCT level control system could:
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a.
Cause the letdown flow to be diverted to the liquid holdup tank.
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Cause the VCT liquid inventory to decrease due to normal charging flow without any makeup to the VCT via letdown.
Without operator intervention, the VCT could empty causing the operating centrifugal charging pump (high head safety injection system pump) to be damaged due to loss of suction fluid.
Upon a certain set of postulated conditions, the adverse interaction could lead to a system not meeting the single failure criterion.
It was determined by the station that the operator has various additional annunciators to warn him/her of a failure of VCT level controller. These alarms were determined to be sufficient to ensure that manual operator action would be taken:
Type of Plant Alarms High VCT Level Low VCT Level Full Letdown Flow Divert Refueling Water Transfer Automatic Makeup Start in VCT Low-Low VCT Level Low Charging Flow Lew RCP Seal Injection Flow High Temperatures Low Pressurizer Level Operator action is acceptable unless a generic position to the contrary is obtained from NRC Inspection Enforcement Headquarters.
No items of noncompliance were identified.
10.
Radiation Chemistry Foreman Round the Clock Coverage The licensee committed in response to Inspection Reports 50/295/80-05, 50-304/80-04 to provide radiation chemistry foreman on each shift. This-7-(
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commitment has been modified because the need for foreman coverage between midnigh; and six A.M. has been re-evaluated. Due to changes in staffing policy, the licensee can guarantee that radiation chemistry technicians meet the applicable ANSI 18.1 requirements for experience and training during periods of time when no foreman coverage exists (ANS1 18.1,1971).
No items of noncomp.11ance were identified, i
11.
Radioactive Release :
a.
Releasa of May 7, 1981 During recovery f rom the safety injection of May 7,1981 (see para-graph 5) it was necessary to divert large quantities of coolant to the hold up tanks. The divert was required to reduce pressurizer level and dilute the RCS back to the concentration required for start up.
During che divert the liberated gases escaped from the waste gas system and were released via the auxiliary building vent stack.
The total activity released was 1.54 curies and the maximum instantaneous release rate was 1.6% of the techaical specifications limit.
No items of noncompliance were identified.
b.
Release of Aptil 27, 1981 At approximately 11:00 A.M. April 27, 1981, the inspector observed an increasing trend on the recorder for 0-R-14, the auxiliary building vent gaseous activity monitor. The operators were notified and commenced efforts to determine the source. The increase continued and approximately one hour later 0-R-14 reached its alarm point of 10E4 cr2. While checking various other potential sources, the waste gas analyser was de-energized closing the thirteen solenoid inlet isolation valves. Approximately twenty minut(s later the 0-R-14 trend started d:_ creasing. By 1:45 P.M.
the 0-R-14 reading was back to normal.
Subsequent investigation showed that a trap on the waste gas analyser drain line designed to allow only water to pass to the auxiliary building equipment drain tank was allowing gas flow also. The auxiliary building equipment drain tank ie vented to the auxiliary building vent header. The licensee calculated that the total release was five curies and the maximum instantaneous release rate was 4.5% of the technical specifications limit.
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No items of noncompliance were ider tified.
Spread of Contqmination Off Site c.
On April 14, 1981 a contractor employee became contaminated on his hands (50,000 cpm right, 25,000 cpm lef t) while wor king on a boric acid evaporator.
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His hands were decontaminated except that 400 cpm remained on his right j
thumb which could not be removed. He was issued a glove for that hand l
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and allowed to let.ve the site with the intention that he would be re-checked the following day to see if he had " sweated it out".
In the process of leaving the site he set off the portal activity monitor but it was assuned to be caused by the 400 cpm on his thumb. When he was rechecked the next day it was discovered that he had 12,000-13,000 com on his clothes (the same c3cthes he had worn the previous day.'
Surveys performed by the licensee showed contamination in his car (13,000 cpm max), house (700 cpm max), and friend's car (3,000 cpm max).
Surveys of his friend's house showed no contamination. The house and cars were decontaminated by licensee personnel. This matter is discussed further in IE Inspection Report Nos. 50-295/81-15 and 50-304/81-11.
No items of noncompliance were identified.
12.
Diesel Generator Operability with Less Than Twelve Cylinders During a review of station deficiency reports the inspector found DVR-22-1-81-3 which documented an occurrence where one cylinder of the 1A diesel generator was inoperable due to a broken rocker arm and bent push rods. The occurrence was classified nonreportable. The licensee's position is that the diesel was operable since:
a.
The malfunction _ould not have affected other cylinders b.
The diesel was carrying rated load with the cylinder inoperable and is designed to do so with two inoperable cylinders.
The licensee is attempting to obtain documentation from the vendor of the diesel's capability with less than all cylinders operable.
Resolution as to diesel operability with less than twelve operable cylinders is designated as Open Item 295/81-09-02, 304/81-05-02.
13.
Unit 1 TMI Modifications and Technical Specifications Modifications to the Unit 1 sampling system which were required ~aecause of requirements contained in NUREG-0737 necessitated the insta11ntion of automatic sampling system valves in place of the existing manual valves. The theory is that during a radiological emergency, prohibitively high radiation field may make manual sample system valves inaccessible.
The licensee has made the following valve changes:
Removed Replaced by 1PR0020 ISOV-PR26A ISOV-PR'25A 1PR0008 ISOV-PR26B ISOV-PR25B-9-
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1PR0016 ISOV-PR26C ISOV-PR25C 1PR0012
'ISOV-PR26D ISOV-PR25D Unfortunately, the valves removed are referenced in the technical specifications while new valves are not.
The licensee has submitted a technical specification change to reference the new valves but the revision has not been received at the station. In effect, work was begun on a system bcfore the technical specifi-cations were issued.
The inspector determined that all procedures referencing the old valves had been revised to reflect the new valves. These procedures are as follows:
PT-10 RP-1610-7 S01-18-APP-A-1 ZEP GOP Drawing will be revised when the modificat'lon is closed out, i'
This item is open pending NRC issuance of revised-technical specifications and is denoted (295/81-09-03, 304/81-05-03).
14.. Operational Safety Verffication The inspector observed control room operations reviewed applicable logs and conducted discussions with control room operators during the months of April and May. The inspector verified the operability of selected emergency systems, reviewed tagout records and verified proper return to service of affected components. Tours of the auxiliary building and turbine building were con-ducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, and excessive vibrations and to verify that maintenance requests had been initiated for equipment in need of maintenance. The inspector by observation and direct interview verified that the physical security plan was being implemented in accordance with the station security plan.
The inspector observed plant housekeeping / cleanliness condicions and verified implementation of radiation protection controls. During the month of May, the inspector walked down the accessible portions of the auxiliary feed system to verify operability. The inspector also witnessed portions of the radioactive waste system controls associated with radwaste shipments and barreling.
These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established under technical specitications, 10 CFR, and administrative procedures.
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15.
Monthly Maintenance Observation Station maintenance activities of safety related systems and components listed below were observed / reviewed 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 accom-plished using approved procedures and were inspected as applicable; functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; and, fire prevention controis were implemented.
Work requests were reviewed to determine status of outstanding jobs and to assure that priority is assigned to safety related equipment maintenance which may affect system performance. Maintenance activities were observed on the "0" diesel generator.
Following completion of maintenance, the inspector verified that the "0" diesel generator had been returned to service properly.
16.
Monthly Surveillance Observatioa The inspector reviewed technical specifications requiring surveillance testing on the diesel generators and verified that testing was performed in accordance with adequate procedures, that test instrumentation was calibrated, that limiting conditions for operation were met, that removal and restoration of the affected components were accomplished, that test results conformed with technical specifications and procedure requirements and were reviewed by per-sonnel other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel.
17.
Licensee Event Reports Followup Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine that re-portability requirements were fulfilled, immediate corrective action was ac-complished, and corrective action to prevent recurrence had been accomplished in accordance with technical specifications:
LER NO.
UNIT 1 81-19 Failure of ORT-PR-25 81-11 ORT-PR-17 Found Out of Service-11-
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LFR NO.
UNIT 1 81-12 Snubbers Failed Functional Test 81-13 Standby Instrument Air ~ Compressor Failed to Start 81-14 Standby Instrucent Air Compressor Failed to Statt 81-15 Failure of GRE-0005 81-16 Ft s. lure of RHR Mini-flow Valve 81-17 Failure of Heat Trace Circuit Breaker 81-18 Failure et Heat Trace 81-19 Failure of ORT-PR-25 81-10 Out of Specifications Boric Acid Concentration LER MO.
UNIT 2 81-04 Trip of "0" Diesel Generator 81-05 Water in "0" Diesel Generator Lube Oil 81-06 Class I Pipe Tunnel Rad Monitor
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IE Bulletin Followup
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For the IE Bulletin listed below the inspector verified that the written response was within the time period stated in the bulletin, that the written response included the information required to be reported, that the written respouse included adequate corrective action commitments based on information presenta-tion in the bulletin and the licensee's response, that licensee management forwarded copies of the written response to the appropriate onsite management representatives, that information discussed in the licensee's written response was accurate, and that corrective action taken by the licensee was as described in the written response:
IES NO.
81-02 Failure of Gate Type Valves to Close Against Differential Pressure 19.
Three Mile Island Requirements The inspector reviewed the licensee's implementation of Three Mile Island lessons learned requirements contained in NUF.EG-0737. These requirements were inspected in November 1980, March 1981 and June 1981. Variances dis-
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covered between NUREG positions and Zion Station field implementation were identified to NRC management. The following is a summary of the items in-spected.
No items of noncompliance were identified.
l.A.l.1 Shift Technical Advisor -
on Duty Inspector reviewed implementing procedure ZAP l-51-1.
The following variance was noted: The SIA at Zion is a SRO position that involves routine and emergency functions. Routine activities involve control room supervision equivalent to a SRO licensed shift foreman.
Emergency activities involve being the technical advisor to the shift. NRR has been made aware of the Zion STA job descriptions in previous correspondence, l.A.l.2 Shift Supervisor Responsibilities Inspector reviewed implementing procedure ZAP l-51-1 specifying the management and safe operation function of the shift engineet. No var-iances with TAP requirement was noted.
1.A.l.3 Shift Manning -
Overtime Limits Specified Inspector reviewed ZAP 10-52-3, " Shift Manning Relief and Turnover" which endorsea the NRC guidelines. The inspector noted the following variances:
(1) Ceco guidelines only apply to NRC licensed individuals on shift work who have the control roon as their duty station.
(2) The existing union collect.ve bargaining agreement conflicts with the NRC overtime limits.
(3) The station disagrees with the guidelines requiring a break of at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> between work periods.
I.C.I Short Term Accident and Procedures Review
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Shift Relief Turnover l
Procedures i
Inspector reviewed implementing procedure ZAP 10-52-3 specifying turn-l over sheets for shif t engineer, radwaste foreman, nuclear station
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operator and center desk equipment operator.
No variances were noted.
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I. C. 3, Shift Supervisor Responsibilities Inspector reviewed ZAP 1-51-1, " Station Organization". No var'ances were noted.
I.C.4 Control Rogm Access Inspector reviewed implementing procedure ZAP 5-51-8.
No variances were identified.
I.C.5 Feedback of Operating Experience Inspector reviewed ZAP 2-52-3.
No variances were noted.
I.C.6 Verify Correctness of Operating Activities
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Inspector reviewed ZAP 1-51-1, "Stetion Organization", and ZAP 10-52-3,
" Shift Manning Relief and Turnover". The station has taken steps to include SRO shift foreman independent verification of all periodic testing (surveillance) that is.related to safety. The station has taken steps to include jumper and lif ted leads under independent verification.
II.B.3 Post Accident-Sampling and III.D.3.1.3 In Plant Radiation Monitoring The inspectors reviewed ZCP-500, ACP-123A, ACP-123B, RP 1740-1, and RP 1740-3. These procedures were also reviewed by the Health Physics Appraisal Team (IE Inspection Report Nos. 50-295/80-05 and 50-304/83-04).
The following variances were identified during the Health Physics Appraisal.
Corrective actions for these variances will be reviewed by regional radiation protection specialists. No further variances were identified by the inspectors.
(1) Containment sample line RE 11/12 has a low point and could fill with water in the event of a steam environment in containment.
(2) Procedure ZCP-500 does not contain precautions or limitations on containment sampling in the e"ent of a positive pressure in containment.
(3) Procedures for use of a lead cask and dumbwaiter for sample mov;-
ment appear to need improvement.
(4) Procedures RP 1740-1 and RP 1740-3 fail to warn against purging charcoal canisters in the counting room.
The licensee has received on site two SAM-2 Stabilized Assay Meters for measure of high level radioactive samples. The SAM-2's are cali-brated and a procedure is written for their use.
Silver Zeolite Cartridges are on site. No variances were identified.
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II.D.3 Valve Position Indication Relief f
Valve and Safety Valves Inspector reviewed annuciator location 10A " Pressurizer Safety Valve, PORV Open".
The system relies on accoustic monitoring of flow through safety and relief valves. Control room operators demonstrated that alarm is operable by injecting a test signal and verifying alarm annunciated. No variances were noted.
Inspector verified through review of design documents that PORV position indication is powered from a vital bus.
t Inspector vnrified that PORV block valves are operable during normal operation.
PORV block valves are opened for low temperature over
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pressure protection of the RCS.
II.E.1.2.
Auxiliary Feed System Initiation
and Flow j
The licensee has responded to NRC in a letter dated December 14, 1979 that the flow indication for the AFWS was not safety grade and t' hat no modifications were planned.
The AFWS flow indication is powered through inverters which receive j
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power from station batteries as well as a dirty feed for back up.
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battery charges receive emergency power from the diesel generators.
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II.E.3.1 P - cuncy Power for
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rizer Heaters i
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No variances were noted. The inspector reviewed changes made to the
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following Emergency Operating Procedures to remind the operator to check i
to restore pressurizer heaters involving transients where pressurizer
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level is lost.:
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i E0P-10 Steam Generator Tube Rupture E0P-7 Station Blackout E0P-8 Main Steam Feedline Break
(Procedure changes will be made to include the reminder to
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restore pressurizer heaters)
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II.E.6.1 Recombiner 't'roceaures Reviews and Ug rap i
Inspector reviewed S01-9 section 4.6 for revise.1 recombiner, procedures.
Procedures are in place.
II.E.4.2 Isolation D_ependebilit.y l
Diverse containment isolation is provided through SI signal which has
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diverse input s.
Diverse parameters are used for MSIV closure. No vari-l
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ances noted.
.1 The inspector reviewed IIcensee justification submitt ed t o NRR for non-
essent tal syst ems that are not isolat. d.
The inspector reviewed the licensee commitments made to dedicate a person I
to close manually open valves af ter an accident. NRC Safety Evaluat!on
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Report (SER) of February 28, 1930 describing the status of the 11ccesee's
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j compliance with TMI-2 lessons 1carntd requirements state that the licensee will delegate an individual to close such manually open valves in the
event of an emergency or when operation is complete.
l The inspector's review determined that the licensee's actual commitment is at variance with the SER document. Manual awl throttle valves in this category are in the RCP seal injection lines. These valves are located in ?he pipe chase. The licensee has committ ed wJ documented in E0P page 14 to isolat.e these valves manually and initiate isolation seal
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water if radiation Icvels in the pipe chase perait operator entrance.
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Inspector reyf ewed jumper and lif t ed Icad log which documents modifications made to ensure prevention of automatic reopening of containment isolation
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valves upon rest of isolatite signal.
No varia u es were noted.
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II.E.4.2.(Sa)
Containment Pressure Setpoint_
l The licensee has not determined that any setpoint changes are required.
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l II.F.2 Instrumentation to Detect Inadequate Core Cooling Inspector re'.elewed installed subcooling meters on both Unit 1 and Unit 2.
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Inspecter reviewed dOP-9 Appendix B which describes use of subcooling meters. No variances were noted.
E0P-9 includes references on how to compute subcaoling if monitor talls.
11.0.1 Power Supplies for Pressurizer Relief Valves Pressurizer level instrument channels are powered independently through inverters from the station batteries with a dirty power backup. No variences were noted.
Power supplies for PORV solenoids are energized from safety reirted power sources. Associated block valve is powered from a safety related power supply different from the PORV solenoids. Plant has modification pending which has not been completed which would further diversify powe? supplica of I'ORV and its associated block valves. Ne variances noted.
II.K.3.9 PID Controller No variances were noted by the inspector.
III.A.l.2 Upgrade Emergency Support Facilities Inspector reviewed EPIP-300-3. This procedure describes clcsed off area in meeting room as interim TSC. TSC was inspected for presence of telephone cx munications.
Inspector witnessed startup and operation of PING-1, particulate iodine and noble gas monitor outside OTSC. No variances were noted.
.Onsite_ Operational Support Center Inspector reviewed EPIP-300-4 which designates lunchroom as Interim Onsite Oparational Support Center.
III.D.l.1 Primary Coolant Gatside Containment Inspector reviewed ZAP 10-52-4 which establishes a leak reduction program.
No variancer were noted.
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- 20P500 " Post Accident Sampling and Analysis (Containment and Reactor Coolant)"
- ZCP123A " Hydrogen Analysis of Gas Samples" j
- RP1740-1 " Monitoring High Activity Releases During an Accident"
- RPl740-3 " Radioactive Samplings Under Accident Cenditions"
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- ZCP23A " Boron Analysis of Gas Samples" i
20.
Meetings and Offsite Functions l
The inspectora attended the fellowing offsite functions during the inspection period:
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May 5, 1981 Meeting with Mayor Spencer, Zion, Illinois May 13-14, 22, 1981 OIE Region III Office 21.
Unresolved Items
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Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, items of noncompliance or deviations. Three unresolved items were disclosed during this inspection.
22.
Exit Interview The inspector met with licensee representatives (cenoted in Paragraph 1)
throughout the month and at the conclusion of the inspection on May 29, l
1931 and summarized the scope and findings of the inspection activities.
The licensee acknowledged the inspector's comments.
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