IR 05000295/1981026
| ML20040B653 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 12/11/1981 |
| From: | Hayes D, Kohler J, Waters J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II), NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20040B639 | List: |
| References | |
| TASK-2.K.3.09, TASK-TM 50-295-81-26, 50-304-81-23, IEB-80-11, NUDOCS 8201260287 | |
| Download: ML20040B653 (12) | |
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U.S. NUCLEAR RECULATORY C05DfISSION OFFICE OF INSPECTION AND ENFORCEMENT
REGION III
Report No.:
50-295/81-26; 50-304/81-23 Docket No.:
50-295/304 License No.:
Commonwealth Edison Company P. O. Box 767 Chicago, IL 60690 Facility Name:
Zion Nuclear Power Station, Units 1 & 2 Inspection At:
Zion, TL Inspection Conducted: October 2, 1981 through November 30, 1981 7 E. /(As Inspectors:
J. E. Kohler
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J. R. Waters
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Date 1/0
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J et t M Approved By:
. W. Hayes, tief M/h'
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Reactor Projects Section IB
'Dat6 Inspection Summary inspection on October 2, 1981 through November 30, 1981 (Report No. 50-295/81-26; 50-304/81-23 Areas Inspected:
Routine unannounced resident inspection of licensee actior,on previous inspection items, reactor trips, radioactive releases, fuel assembly damage, primary leakage, DB-50 breaker cracks, containment spray discrepencies, foreign objects found in the primary and secondary side of steam generators, radiation monitor operability, S/G level set point changes, status of TMI items, primary to secondary leakage, refueling operations, preparation for refueling, operational safety verification, monthly surveillance, licensee event reports, IE Bulletin follow up, auxiliary feedpump inoperability and reportabflity of releases. The inspection involved a total of 510 inspector hours onsite by two NRC inspectors including 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> onsite during off shifts.
Results: No items of noncompliance were identified.
8201260287 820113 PDR ADOCK 05000295 G
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1.
Persons Contacted l
- K. Graesser, Station Superintendent
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Fuerst, Assistant Station Superintendent-Operations
- G. Plim1, Assistant Station Superintendent-Administrative
and Support Services j
R. Budowle, Unit 1 Operating Engineer J. Gilmore, Unit 2 Operating Engineer
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L. Pruett, Assistant Technical Staff Supervisor P. LeBlond, Assistant Technical Staff Supervisor
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A. Miosi, Technical Staff Supervisor B. Schramer. Station Chemist D. Walden, Fuel Handling Foreman i
F. Ost, Health Physics Engineer C. Silich, Technical Staff Engineer-ISI l
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Harl, Quality Ascurance Engineer
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Lukens, Quality Control Engineer i
- Denotes those present at management exit of November 30, 1981 I
2.
Summary of Operations Unit 1 Unit 1 operated at power levels up to 100% with no reactor trips or unscheduled shutdowns.
Unit 2 i
Unit 2 remained shutdown for a scheduled refueling outage until November 24, 1981.
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At this time the unit was placed in hot standby for zero power physics testing.
On November 26, 1981 the unit was shutdown due to auxiliary feed pump inopera-bility (See paragraph 17). The problem was corrected and the unit was returned
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to hot standby the same day. As of the end of the inspection period the unit had not yet resumed power operation.
3.
Licensee Action on Previous Inspection Findings (closed) Unresolved Item (50-295/81-09-2, 50-304/81-05-02) Diesel Generator Operability with Less Than All Cylinders Operable.
Correspondence between the licensee and the diesel vendor indicate that the diesel generator could operate at full load with one inoperable cylinder for at least seven days. However, to do so would require securing the fuel pump to that cylinder to prevent the buildup of unburned fuel.
Since this was not done during the original occurence i
the licensee has submitted LER 50-295/81-36 on the inoperable 1A diesel generator.
(closed) Unresolved Item (50-295/80-25-1, 50-304/80-27-1) Release Apparently Caused by Boric Acid Evaporator Gas Stripping of Raw Coolant. The licensee has taken several steps to preclude recurrance. The seals on the boric acid feed pumps have been replaced, and the internals of numerous diaphragm valves in the volume control and waste gas system have been replaced. Also, the minimum hold up tank level has been raised to 15%. No further releases of this type j
have occurred since.
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(closed) Item of noncompliance documented in Inspection Report 50-295/81-20; 50-304/81-16, Violation of 10 CFR 50.59.
The inspector has reviewed the l
licensee's corrective actions as stated in the response to the notice of violation.
4.
Gaseous Radioactive Releases During the inspection interval the following radioactive gaseous releases occurred:
During a twenty-hour period between October 28, October 29, 1981, a.
approximately 5700 cubic feet of waste gas inadvertently escaped to the environment from " nit 1.
The volume of gas released was estimated to contain approxim.tely 1.23 curies of nrimarily Xe-133 and was terminated upon discovery. Idacovery of the release was made difficult by the fact that no installed radiation monitoring equipment experienced abnormal trends which would signify an ongoing release.
No increased radioactive release trends were experienced because the gas released was of a low specific activity and released over an extended time period.
The inspector investigated the circumstances associated with the release.
A temporary bypass hose connection had been installed from the Unit I sump pump to the reactor coolant drain tank bypassing the normal sump.
The bypass permitted Unit I to continue to discharge liquid while the normal sump discharge path was out of service to install a tie in for the Three Mile Island sampling system. After the tie in was complete the normal sump discharge path was restored.
However, the hose coanection was left installed and the bypass valve left open. This permitted a gaseous leakage path to develop from the waste gas tank through the reactor coolant drain tank, to the auxiliary building equipment drain tank and released as filtered exhaust.
Installed radiation monitor on the ABEDI showed no abnormal trending. The release was detected when the pressure in one of six waste gas tanks unexpectedly dropped to atmospheric pressure from 45 psig.
The review indicated that the failure to close the bypass valve upon restoration of the normal sump discharge path was an oversight on the part of the engineer in charge.
However, upon restoration of normal sump pumping it was not immediately apparent that the bypass line resulted in a waste gas tank leak path. The evolution received a coordinated review to take into account operational and radioactive waste discharge considerations.
This event was reported to the NRC on October 17, 1981 on the Emergency Notification System and is designated unresolved item 295/81-26-04 and 304/81-23-04 until compliance with T.S.
3.12.2 is clarified by the NRC.
b.
On November 12, 1981 the waste gas system was full and a tank release was required. One of the reasons the system was full was related to the dilution operation on Unit 2 for startup af ter refueling.
The dilution was one which resulted in a decrease in holdup tank volume and an increase in gas decay tank pressure.
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The licensee reviewed the applicable technical specifications for release of gas decay tanks and concluded that release of the two lowest activity tanks was permitted, even though neither tank had 45 days decay because gas associated with refueling and purge and fill operations require no decay.
The NRC took the position that the tanks could not be released without 45 days decay because the waste gas from both units is a
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common system and not all the gas released was associated with a refueling or purge and fill operation.
This event is designated unresolved item 295/81-26-05; 304/81-23-05 until compliance with T.S. 3.12.2.A is clarified by NRC.
5.
ENS Reportability of Gaseous Radioactivity Releases The licensee requested clarification regarding gaseous radioactive releases I
that are not detectable on radiation monitoring equipment but can be de-i tected by studying trends in the gas volume inveatory control program and changes in waste gas tank pressures.
The inspector stated that the emergency notification system should only be used for releases that are detectable on installed radiation monitoring equipment. Other releases, the result of equipment failure such as valve leakage, should be reported by the normal reporting mechanisms.
No items of noncompliance were identified.
6.
Radiation Monitors Out,of Service In continuing review of the licensee's rndiation monitoring program, through review of licensee event reports and direct inspection activity, the following
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observations were made: Day to day there can be a large percentage of the l
station's various process monitors out of service for which daily grab sampling as permitted by technical specifications, are substituted. Of the personnel errors committed and documented in licensee event reports, a large percentage are the result of missed surveillance sampling.
It is frequently difficult to tell which radiation monitors are formally declared out of service since out of service cards often are not used.
The inspectors discussed the benefits and drawbacks of the following types of increased radiation monitoring surveillance:
a.
Establish a maximum period of time for which grab sampling can be substituted of process radiation monitoring. After this time an onsite review would be required to document the long term corrective action necessary to restore the process monitor to a reliable operable condition.
b.
Establish a rule whereby operating shift foreman review the status of radiation monitors that are out of service and caution tag these monitors so that they are visible during inspection.
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c.
Establish a computerized tracking system of all required effluent surveillance activities eminating from either routine sampling
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requirements or a limiting condition for operation.
No items of noncompliance were identified.
i 7.
Containment Spray Initiation Time Less Conservative Than That Assumes in
the FSAR i
While performing valve timing tests pursuant to technical specification
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3.3.4 the licensee observed that the containment spray isolation valves
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normally require about 50 seconds to fully open.
The FSAR assumes that contain-ment spray flow is achieved at 45 seconds after a LOCA. The actual containment spray time was calculated at 92.5 seconds based on safety injection signal j
time, diesel generator start time, containment spray pump start time, 60 second j
to open isolation valves, and time to fill the spray header.
Consequently, the l
licensee submitted a prompt notification to Region III as required by tech-l nical specifications.
Subsequent calculations by the licensee demonstrated that the delay in spray flow initiation would increase the post accident peak containment pressure
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from 39.6 psig (based on 45 seconds for spray initiation) to 40.64 psig.
This value is well below the 47 psig containment design pressure and no further action was taken.
This item is currently under NRC evaluation.
8.
Proportional Integrated Derivative Controller (PID)
The inspector determined the Item II.K.3.9 contained in the Three Mile Island Task Action Plati concerning control of the power operated relief valves (PID controller) was modified as recommended by Westinghouse. The modification was completed on April 30, 1979.
No items of noncompliance were identified.
9.
Masonry Walls IE Bulletin 80-11
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The inspector reviewed the licensee's response to IEB 80-11 and determined that an operability review involving the possible effects of a masonry wall toppling i
on safety related equipment was performed.
No items of noncompliance were identified.
10.
Fuel Damage Due to Unretracted Incore Thimble During Unit 2 Fuel Shuffle
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While attempting to complete the Unit 2 fuel shuffle by inserting a new fuel assembly into core position C-8, the refueling crane met resistance at about six feet above the lower core plate. The resistance was indicative of an obstruction in the core.
A television camera was inserted in the vacant
position and it was determined that the cause of the resistance was a bent-5-
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probe thimble. This was unexpected, since all incore probes by procedure are withdrawn prior to fuel shuffle. Adjacent fuel elements were examined for damage and one, a second generation element, was found to have a missing spring clip, torn grid strap, and a loose rod, and was found to be unusable.
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Westinghouse was contacted and due to core symetry considerations, a new I
core load pattern was required.
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A hydraulic gripper was used to stretch the incore thimble and cut it off as close to the lower core plate as possible. The remaining portion of the thimble j
was withdrawn from the incore seal table. A new thimble was installed at a
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later date.
Licensee investigation was performed to determine why the thimble was not retracted along with all other incores prior to refueling. The investigation
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revealed mitigating circumstances which are described below.
During the previous Unit 2 refueling outage (approximately March 6, 1980), the licensee was attempting to meet a January 1, 1981 deadline for installation of reactor vessel level indication (this deadline hac been revised). The vessel level modification required taps to be installed at the top of the reactor (control rod drive mechanisms) and at the bottom of the reactor at the seal table penetration for incore probe C-8.
At the time of June 1980, all parts
were nat available from Westinghouse to be able to tap into the incore position C-8 and still have use of that incore position. The parts not available were certain seal assemblics. A method was devised and implemented whereby reactor vessel level could be gained to avoid a unit shutdown of January 1, 1981 by inserting the incore thimlle C-8 and cutting and capping it above the seal table. This would make C-8 unusable for one fuel cycle.
l During the present refueling, the maintenance instruction used to pull the incores was uritten to identify the fact that C-8 was inserted in the core but was cut and capped above the seal table. Apparently, however, the C-8 incore probe was never retracted. Re-reading the maintenance instruction for uncoupling the incores does not specifically tell the individual how to retract C-8 and is not clear in hindsight.
Consequently, the cause of C-8 incore probe not being retracted appears to be a procedural inadequacy.
The fuel element that was attempted to be inserted into C-8 was an element that had experienced a torn grid strap during fuel movement the week of October 6, 1981. This element was repaired by Westinghouse in North Carolina and returned l
to the site. At the time it was determined that this fuel element had a damaged grid strap, the presence of the unretracted incore probe was unknown.
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It is possible that the damage to the element was somehow related to this unknown obstruction.
No items of noncompliance were identified.
11.
Crid Strap Foreign Object In IC Steam Generator On October 3, 1981, the resident ir.spector's office was notified that a foreign object appearing to be a piece of Irid strap from a fuel element was discovered on the scaffolding erected underneath the 2C steam generator, adjacent to a 55 gallon drum. The foreign object was measured to be about 30R and was-6-
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discovered durin an extensive area survey performed af ter dosimetry badges s
for workers in area of 1C steam generator were higher than expected.
Upon discovery, the general area was cleared and all dosimetry for workers was evaluated. No overexposures were identified. The concern raised was that a worker might have inadvertantly stepped on this object. An analysis was performed to determine the maximum possible extremity dose that might have been received.
The analysis assumed that the closest worker was three to six inches from the object allouing for clearance from the 55 gallon drum.
In the worst case, no worker would have received a dose in excess of the maximum extremity dose of 18.7R specified in 10CR20.
The licensee investigated the circumstances to determine why the original survey did not detect this object.
It was determined that the object's radiation field would probably have been masked by the generally high radiation field resulting from the IC steam generator tube sheet which was exposed for per-formance of eddy current.
Review of plant records indicated that the IC steam generator had not been opened for at least five years. The object most likely fell out of the generator lower plenum when the manway was removed, and it was impossible to determine the fuel assembly that yielded this object. There is a high probability that the damaged fuel assembly has already been discharged.
No items of noncompliance were identified.
12.
Foreign Object Discovered During Sludge Lancing When the 2B steam generator was opened for sludge lancing, a seven inch piece of spring clip from a previous sludge lancing operation was discovered. The piece had been in the steam generator for at least one cycle. The piece was removed.
A borescope was obtained to search for additional pieces that may have been in the steam generator. None were found.
Sludge lancing was performed and the generator inspected with a borescope prior to closure of the manway.
No items of noncompliance were identified.
13.
Westinghouse DB-50 Circuit Breakers The resident inspector's office was informed by the licensee that cracks had been discovered in base plates holding contacts for Westinghouse type DB-50 circuit breakers. The cracks were discovered after disassembling one breaker for routine five year inspection and extend about halfway through the base plates. These breakers are used in the control rod drives, reactor trip breakers and the output breakers for the rod drive MG sets. There are six breakers per unit. Three of the base plates were sent to Westinghouse for further analysis.
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An onsite review was conducted to determine the safety significance of the cracking and any remedial action that might be required.
It was determined
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that the only conceivable situation that would have safety significance would be the development of a short duration phase to ground fault which would weld the contacts closed in both the two reactor trip breakers and the two control rod MG set breakers. However, it was determined that the proba-bility of a fault occurring with an intensity large enough to cause instan-taneous welding of all contacts was remote. Furthermore, should such a fault develop, the instantaneous overcurrent trip of the rod drive MG sets would trip, resulting in a reactor trip.
With the conclusions arrived at above, the licensee determined that the base plate cracking did not represent a threat to the safe operation of either unit.
The cracked base plates that were found during the breaker inspection were replaced. Four breakers remain uninspected awaiting final analysis from Westinghouse.
This item is designated Open Inspection Item 295/81-26-01, 304/81-23-01.
No items of noncompliance were identified.
14.
Unit 2 Primary Coolant Leak on Vessel _ Level _ Indicator At approximately 3:30 A.M. November 11, 1981 operators discovered a primary coolant leak on 3/8 inch tube connecting the reactor vessel level instrument to top of the vessel. Plant pressure had just been raised to normal operating pressure. Attempts to shut the manual isolation valve and stop the leak were impeded by the high temperature steam and water issuing from the leak. A site unusual event was declared and plant cooldown and depressurization was commenced. At a plant pressure of approximately 400 psig the manual isolation valve was closed and the Icak stopped.
The leak was found to be caused by a cracked swageloc fitting. The fitting was repaired and the instrument returned to service.
Since the upper head area had been saturated with steam and condensate, the licensee checked the rod position indicator and control rod drive (CRD)
resistance readings. One CRD coil resistance could not be brought into specification. This required replacement of the CRD.
No items of noncompliance were identified.
15.
Steam Generator Low Low Trip The inspector reviewed the licensee's recent setpoint change which restored the steam generator low low level trip to 10% wide range. The previous trip point had been raised to 15% wide range level in response to a generic Westinghouse concern. The concern involved a nonconservative bias that might result in the level measuring instrument as a result of a high energy line break.
The licensee reviewed the Westinghouse generic concern and performed a Zion specific safety analysis.
It was determined that the Westinghouse analysis-8-
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was overly conservative and resulted in increased reactor trips during startup and power operation.
No items of noncompliance were identified.
16.
Unit 1 Primary to Secondary Leakage The licensee has continued to monitor the Unit 1 primary to secondary leakage. The earlier leak test results are documented in inspection reports 50-295/81-09, -14 and -20, and 50-304/81-05, -10 and -16.
The leak test results for October and November 1981 are as follows:
Date 1A S/G 1B S/G IC S/G ID S/G October 2 205.9 gpd October 5 246.7 gpd October 9 13.4 gpd 255.5 gpd 31.3 gpd 20.3 gpd October 14 none det.
238.7 gpd 20.2 gpd 18.4 gpd October 23 11.9 gpd 220.0 gpd 29.5 gpd none det.
October 30 235.9 gpd 23.6 gpd November 9 230.9 gpd 9.62 gpd November 16 283.8 gpd 17.6 gpd November 23 264.0 gpd 26.0 gpd The technical specification limit is 500 gpd leakage from any generator.
The licensee is continuing to monitor leak rates.
No items of noncompliance were identified.
17.
Inoperable Auxiliary Feed Pumps On November 26, 1981 operators discovered that the 2B and 2C auxiliary feed pumps could not be started.
Investigation showed that the pumps were tripping out on low suction pressure and that by throttling the discharge valves the pumps could be started.
Once running the discharge valves could be re-opened and the pumps would continue to operate. Since the unit was in Mode 2 requiring two of the three auxiliary feed pumps operable, a unit shutdown was performed.
The licensee ascertained that the suction pressure switches had not been set properly.
The pressure switches were adjusted, proper pump operation was verified and the unit was returned to hot standby.
The licensee also determined that the same maladjustment existed on the IC auxiliary feed pump.
Unit I was operating at full power.
The IC auxiliary feedpump suction pressure switch was readjusted and the pump operationally tested satisfactorily.
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J The licensee is currently investigating the cause of the apparent improper setting of the suction pressure switches. The regulatory aspects of this event
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are currently under NRC review pending receipt of additional information. This item is designated Open Item 50-295/81-26-03, 50-304/81-23-03.
18.
Preparation for Refueling The inspector verified that the licensec's 10 CFR 50.59 safety evaluation of the reload core showed that prior NRR review is not required. The in-spector also reviewed the licensee's program for overall outage control.
No items of noncompliance were identified.
19.
Refueling Activities The inspector observed two shif ts of the fuel handling operations (removal inspection and insertion) and verified the activities were performed in ac-cordance with the technical specifications and approved procedures; verified that containment integrity was maintained as required by technical specifi-cations; verified that good housekeeping was maintained on the refueling area; and, verified that staffing during refueling was in accordance with technical specifications and approved procedures.
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No items of noncompliance were identified.
20.
Operational Safety Verification The inspector observed control room operations, reviewed applicable logs and conducted discussions with control room operators duriag the months of 0,tober and November. The inspector verified the operability of selected emergency systems, reviewed tagout records and verified proper return to service of affected components. Tours of Unit 2 containment, the auxiliary building and turbine building were conducted 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 conditions and verified implementation of radiation protection controls. During the month of November, the inspector walked down the accessible portions of the auxiliary feed system to verify operability. The inspector also witnessed portions of the radio-active waste system controls associated with radwaste shipments and barreling.
These reviews and observations were conducted to verify that facility operations were in conformar.ce with the requirements established under technical speci-fications 10 CFR, and administrative procedures.
No items of noncompliance were identified.
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21.
Monthly Surveillance Observation The inspector observed technical specifications required surveillance testing on the safeguards actuation system 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 appro-priate management personnel.
The inspector also witnessed portions of pre-operational testing on the automatic gas analyser.
No items of noncompliance were identified.
22.
Licensee Event Reports Followup Through direct observations, disucussions 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 accomplished, and corrective action to prevent recurrence had been accomplished in accordance with technical specifications:
Unit i LER NO.
DESCRIPTION 81-36 1A Diesel Generator Inoperable 81-40 OIST Drifting Indication 81-41 Loss of Boric Acid Heat Trace Channel 81-42 Post Accident Sample System Surveillance Missed 81-43 Failure of Containment Isolation Valve to Close 81-44 Loss of Failed Fuel Monitor Unit 2 81-17 Mispositioned 120 VAC Power Supply Breaker 81-18 Blower for Containment Particulate and Gaseous Monitor Tripped 81-19 S/G Level Transmitter Reading High 81-21 Inoperable Aircraft Fire Protection-11-
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81-22 Failure of Condenser Of fgas Monitor 81-23 Missed Shiftly Boron Sample i
81-24 Blower for Containment Particulate and Caseous Monitor Tripped Regarding LER 304/81-17, the licensee has committed to install an undervoltage alarm.
The event report review is considered closed, however, the installation
of the undervoltage alarm is classified as Open Item 295/81-25-02 and
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304/81-23-02.
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Meetings _, Offsite Functions The inspectors attended the following meetings and offsite functions during t
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the inspection period:
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J. R. Waters i
I October 19-30, 1981 BWR Technology Course NRC Training Facility, Chattanooga, Tennessee 24.
Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, items of noncompliance or deviations.
Four unresolved items (paragraphs 4, 13, 17 and 22) were I
disclosed during this inspection.
25.
Exit Interview The inspector met with licensee representatives (denoted in paragraph 1)
throughout the month and at the conclusion of the inspection on November 30, 1981 and summarized the scope and findings of the inspection activities.
The licensee acknowledged the inspector's comments.
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