IR 05000295/1986022
| ML20212C220 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 12/19/1986 |
| From: | Burgess B, Eng P, Wohld P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20212C142 | List: |
| References | |
| TASK-2.K.3.10, TASK-2.K.3.12, TASK-TM 50-295-86-22, 50-304-86-20, IEB-86-002, IEB-86-2, IEIN-86-002, IEIN-86-007, IEIN-86-014, IEIN-86-053, IEIN-86-058, IEIN-86-072, IEIN-86-14, IEIN-86-2, IEIN-86-53, IEIN-86-58, IEIN-86-7, IEIN-86-72, NUDOCS 8612290421 | |
| Download: ML20212C220 (27) | |
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U.S. NUCLEAR REGULATORY COMMISSION
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REGION III
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Reports No. 50-295/86022(DRP); 50-304/86020(DRP)
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Docket Nos. 50-295; 50-304 Licenses No. DPR-39; DPR-48 Licensee:
Commonwealth Edison Company P. O.-Box 767 Chicago, IL 60690 Facility Name:
Zion Nuclear Power Station, Units 1 and 2 Inspection At:
Zion, Illinois Inspection Conducted:
September 26 through November 17, 1986 Inspectors:
Mark M. Holzmer Lyle E. Ka r
Pete[R Woh
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/a//fdPfo Reactor Projects Section 2A Datd Inspection Summary Inspection on September 26 through November 17, 1986 (Reports No. 50-295/86022 (DRP); No. 50-304/86020 (DRP))
Areas Inspected: Routine, unannounced resident inspection of licensee action on previous inspection findings; loss of recirculation flow to Unit 1 boron injection tank; incorrect negative flux rate trip setpoints; Unit 1 main steam check valves; September 20, 1986, reactor trip due to valve alignment error; degraded grid voltage; Part 21 report on Anderson-Greenwood manifolds; response to flooding in northeast Illinois; fuel building ventilation filters bypassed; failure of 1B diesel generator; operational safety and engineered safety feature (ESF) system walkdown; surveillance; maintenance; licensee event reports (LERs); training; IE Bulletin followup; followup of Region III requests, follewup of TMI action plan items; site visit by ACRS; site visit by Kenosha County Board members, and IE Notice followup.
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Results: Of the 21 areas inspected, no violations or deviations were
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identified in 19 areas, and two violations were identified in the remaining two areas (failure to maintain continuous recirculation flow through the Unit 1 boron injection tank - Paragraph 4; failure to assure proper negative i
flux rate trip setpoints incorporated into Technical Specifications and
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calibration procedures - Paragraph 5).
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DETAILS 1.
Persons Contacted
- G. Plim1, Station Manager
- E. Fuerst, Superintendent, Production
- T. Rieck, Superintendent, Services
- W. Kurth, Assistant Station Superintendent, Operations
- R. Johnson, Assistant Station Superintendent, Maintenance
- J. Gilmore, Assistant Station Superintendent, Planning
- R. Budowle, Assistant Station Superintendent, Technical Services L. Pruett, Unit 1 Operating Engineer N. Valos, Unit 2 Operating Engineer M. Carnahan, Training Supervisor
- R Cascarano, Technical Staff Supervisor C. Schultz, Regulatory Assurance Administrator V. Williams, Station Health Physicist
- J. Ballard, Quality Control Supervisor
- W. Stone, Quality Assurance Manager, Zion Station
- L. Holden, Engineer, Regulatory Assurance
- C. Albert, Engineer, Station Construction
- L. Simon, Site Superintendent, Station Construction
- R. Niederer, Nuclear Group Leader, Technical Staff
- Indicates persons present at exit interview.
2.
Licensee Actions On Previous Inspection Findings (0 pen) Open Item (295/84023-02):
1A Boric Acid Transfer (BAT) Pump mator pedestal repair.
This was considered an Open Item pending NRC review to determine whether a seismic analysis should have been performed and whether the changes to the pump's configuration could have violated the licensee's seismic analysis.
The BAT pumps are classified as seismic class I, non safety-related equipment. The pump's seismic calculations were furnished by Goulds Pump Inc.
The original component was seismically qualified to ensure that the pump did not become a missile hazard during a Design Basis Earthquake (DBE).
On November 7, 1984, one of the four pump motor supporting pedestals on the 1A BAT pump was discovered to be broken off.
A temporary repair, performed on November 22, 1984, utilized a 3/8 inch thick piece of flat bar, wedged in a motor vent and bolted to the motor base plate.
Subsequently, the motor was replaced on April 16, 1985.
No seismic analysis of the BAT pump was performed prior to or after the temporary attachment was installed.
The NRC resident inspector requested clarification of the acceptability of the 1A BAT pump repair and a determination of whether an analysis should have been performed.
A memorandum from the Station Nuclear Engineering Department (SNED) stated that a seismic analysis was not necessary because the clamp installation was a repair, not a modification to the original design.
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During an October 31, 1986, telephone conversation with the resident inspector, the NRR project manager indicated that a seismic analysis should be performed prior to the repair of a component if the component was part of the original seismic analysis of the plant.
This item will remain open pending written confirmation of the conversation held with NRR.
(0 pen) Violation (295/86011-01; 304/86010-03):
Failure to follow Zion Administrative Procedure, No. ZAP 3-51-1.
This violation consisted of two examples of violation of TS 6.2A and pending the review of the second example in the violation (295/86011-01); failure to follow S01-31 Caution Statement, this item will remain open.
Corrective action included making licensed personnel and equipment operators aware of the event in their routine retraining and by placing a report of the event in the required reading package.
In addition, the Mechanical Maintenance (MM) department has been assigned responsibility for filter changeouts to ensure that this work follows the normal Work Request (WR) procedure.
Prior to this event, the filter replacements were assigned to the stationman group and the followup WR processing was completed by the MM department. The stationmen are no longer coordinating filter replacement work.
(0 pen) Open Item (295/86011-02):
Failure of the IB Main Steam (MS) Check Valve. This item will remain opea pending NRC review of the results of the licensee's inspection of the IA, IC, and ID MS check valves during the current Unit 1 outage and the results of the inspection of the MS check valves during the February 1987 Unit 2 outage. The review conducted by Region III inspectors from the Operational Programs Section, Division of Reactor Safety (DRS) regarding Unit I will be discussed in detail in Paragraph 6 of this report.
(0 pen) Violation (295/86013-01;304/86012-01): Units 1 and 2 Containment Spray Additive Tank Levels below Technical Specification limits.
Corrective action to prevent recurrence included everal revisions to the existing calibration procedures, LIS-CS44 and LIS-CS45, to compensate for density changes and other nonconservative assumptions. Procedural changes will be completed by November 7, 1986.
In addition, an investigation is in progress to find an alternative level indication system which does not rely on density for calibration.
If an alternative l
level indicating system is determined to be appropriate, installation will be completed for each unit by the end of the second refueling outage after March 31, 1987. This item will be considered open pending NRC review of the licensee's decision regarding an improved level indication system.
(Closed) Violation (295/85012-02):
Inappropriate maintenance procedure for relief valve nozzle rings.
Corrective actions included revisions to Maintenance Procedure P/M 000-1N which involved the writing of additional procedures to address the specific types of safety / relief valves, for
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example, steam service, gas service, water service, standard and special.
In addition, in each procedure a statement expressing the zero reference setting required to properly set all relief valves was added for clarity.
This item is considered closed.
(Closed) Violation (295/85012-03A):
Failure of Quality Control (QC)
Inspector to implement corrective action to ensure nozzle ring was reset to its original position.
Corrective action included training for the QC Department on procedure revisions made for P/M 003-1N.
The inspector reviewed the training and found it adequately addressed the problem.
This item is considered closed.
(Closed) Open Item (304/85013-02):
Unit 2 Reactor Trip during ground isolation activities.
This item remained open pending the review of Procedure ZED-3, (ground checking procedure).
The procedure was revised to include instructions to have personnel contact the Operational Analysis Department (0AD) when ground checking involves OAD equipment.
The resident inspector reviewed the procedure revision and found it acceptable.
This item is considered closed.
(Closed) Violation (295/85020-01):
Failure of technical staff surveillance (TSS) 15.6.96.23-1, Appendix C to contain instructions to return manual isolation valve, ICC9499, to normal operating status.
Corrective actions included revisions to TSSs which require valve line-ups to be consistent with the latest revision of the System Operating Instructions (50Is).
Also the Technical Staff Engineer performing the test must verify that the valve restoration line-up coincides with the latest revision of the S0Is before proceeding with the test.
Finally, Zion Administrative Procedures (ZAPS) which control procedure writing and content were changed to require the use of the latest valve line-up revision.
This item is considered closed.
(Closed) Violation (295/85031-01; 304/85032-02):
Failure to meet reporting requirements of 10 CFR Part 50.72.
Corrective actions included discussion sessions, conducted by the Assistant Superintendent of Operations with all Shift Supervisors (SS), emphasizing a conservative approach when reporting actuation of any Engineered Safety Feature (ESF) or Engineered Safety System (ESS) component. A followup memorandum was also sent to all SS and Shift Overview Superintendents to restate their approach to this reporting requirement.
An on-site review was prepared to define the ESF components and a listing of those components was included for the use of SS when evaluating events to determine reportability.
This item is considered closed.
(Closed) Open Item (304/85043-03):
Inadvertent trip of Unit 2 purge.
This was considered an Open Item pending revision to Radiation Procedure No. 1350-8, "Out of Service Surveillance for Radiation Monitors".
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"special instructions" portion to the procedure and surveillance form was added to give specific written guidance to technicians.
The procedure was reviewed by the resident inspector and found acceptable.
This item is considered closed.
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(Closed) Open Item (304/86002-01):
Miswiring of reactor trip switchgear during modification.
This was considered an Open Item pending further investigation by the inspectors into the root cause and a review of the licensee's analysis of the event in a supplemental report.
Review of the supplemental report and discussions with Technical Staff personnel revealed that corrective action was above average.
l The root cause of the event was determined to be the conduit routing on two physical electrical prints which did not show sufficient detail for proper conduit routing and which had not been discovered during the review process.
The cables were pulled through the conduits per the routing drawings and labeled accordingly.
Post modification testing identified the problem before the system was declared operable.
The system was never turned over to the operation's department, and there were no adverse effects on the plant.
Corrective actions were initiated immediately after the event occurred.
Licensee personnel walked down the physical installation and reviewed the installation drawings for the shunt trip modification.
Improper cable routing was identified as the cause and a Field Change Request (FCR)
to detail rerouting of the cables into the appropriate reactor trip switchgear compartment was completed approximately six hours after the event.
Post modification testing was completed the following day, January 23, 1986, at 3:30 a.m.
To prevent recurrence, a training session was held on June 3, 1986 for all Technical Staff Engineers.
The meeting illustrated this event and stressed the importance of reviewing the routing prints as well as other associated prints before final approval.
In addition, the Operational Analysis Department (0AD) will perform construction testing for all modification work performed by the Electrical Maintenance (EM) department after installation.
This will include a continuity test after installation.
Prior to this event, 0AD only used Electrical Construction Test Procedures (ECTPs) on an informal basis (as guidance).
Since the event, ECTPs will be required along with a station modification checklist for each job which involves the EM department.
A visual checklist, which will be implemented by the february 1987 Unit 2 refueling outage, will also be utilized.
The purpose of the checklist is to provide guidance to the various workgroups when performing a visual inspection on the installation of a modification.
This item is considered closed.
(Closed) Unresolved Item (295/86005-03; 304/86005-02):
Operation with incorrect negative flux rate trip (NFRT) setpoints.
This item was unresolved pending resolution of issues raised at a meeting on April 29, 1986 between the licensee and Region III.
A detailed discussion of this item can be found in Paragraph 5 of this report.
(Closed) Violation (304/86005-06):
Failure to take a reactor coolant iodine sample in the time required by Technical Specifications (TS).
Corrective actions included counselling of the Radiation-Chemistry foreman involved on the correct interpretation of the TS requirement.
This event was also discussed at the Radiation Chemistry department
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weekly meeting.
A new procedure has been developed, Zion Chemistry Procedure No. (ZCP) 301-6, " Power Change Iodine Surveillance," to document and assure compliance to TS 4.3.6.
Its accompanying form was also added to ZCP 401.
The procedure was approved on September 22, 1986 and distributed on October 15, 1986.
The inspector reviewed the procedure and its accompanying form and found them to be adequate.
This item is considered closed.
(Closed) Open Item (295/86005-08; 304/86005-07):
Failure to identify exempt valve in inservice testing (IST) program per ASME code.
This was considered an Open Item pending review of an upgraded IST valve program.
By letter dated June 17, 1986, the station submitted the second ten year program to NRR for review.
Temporary relief has been granted until either June 31, 1987 or until the detailed review is completed.
The inspector reviewed the upgraded IST program package.
This item is considered closed.
(Closed) Open Item (295/86005-11):
Obstruction of fire damper in cable penetration vault, and removed Aircraft Fire Damper on cribhouse ventilation fan.
These events were considered an Open Item pending the implementation of additional training on fire barriers, dampert, and fire doors in the licensee's Nuclear General Employee Training (NGET) course.
This Open Item will be closed; however, these events will be incorporated into the existing Unresolved Item which addresses several similar events involving degraded fire barriers.
These LERs will be reviewed collectively by NRC Region III.
See Unresolved Item (No. 295/86019-05),
Paragraph 12 in Inspection Reports No. 50-295/86019(DRP);
No. 50-304/86018(DRP) for details.
This item is considered closed.
(Closed) Violation (295/86011-03; 304/86010-02):
Missed surveillance of control room charcoal filter following replacement.
Corrective action included reinstruction of the Shift Control Room Engineer (SCRE) and the Operating Engineer on the testing requirements of the Technical Specification (TS) related to filter testing.
Licensed personnel and equipment operators were also made aware of the event in their required reading package.
This item is considered closed.
No violations or deviations were identified.
3.
Summary of Operations Unit 1 The unit remained shutdown for the entire reporting period for a refueling and maintenance outage.
The unit is scheduled to be on-line February 1987.
Unit 2 l
The unit remained at full power for the entire reporting period.
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4.
Failure to Maintain Continuous Recirculation of Boron Injection Tank on Unit 1 a.
Event Chronology On July 9, 1986 the 1A Boric Acid (BA) filter was tagged out to replace the sight glass at the bottom of the filter due to leakage
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problems.
Boron injection tank (BIT) recirculation is normally from 1A boric acid transfer (BAT) pump, through the BA filter, to the BIT and back to the OA BA tank.
With the BA filter out-of-service (005), BIT recirculation was from the 1A BAT pump through Valves No. IVC 8460, No. IVC 8469, No. IVC 8470, and No. IVC 8468, respectively, (these were the valves that provided a pathway bypassing the filter),
to the BIT. At 10:25 a.m. on July 23, 1986, the 1A BAT pump was tagged 00S to replace a diaphragm on the 1A BAT pump discharge pressure gauge isolation Valve No. IVC 8462.
Placing the 1A BAT pump 005, combined with the BA filter isolation 005, isolated the BIT recirculation flow path.
At 2:25 p.m. that same day, an audible alarm with no visual indication was observed by the control room operator.
The operator tested his annunciators after receiving the alarm and found that the BIT low recirculating flow alarm lamps had burned out.
Lamp replacement identified the BIT low flow condition.
After discovery, the proper valve lineup was restored and BIT recirculation re-established at 3:15 p.m..
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Root Cause (1) Cognitive error by licensed operators in failing to recognize that BIT tank recirculation was being provided by an unusual l
lineup to accommodate the 1 BA filter being 00S.
l (2) Potential training inadequacies for equipment operators and control room operators who could have identified the potential for BIT isolation prior to issuance of the 00S.
(3) The combination of burned out lamps for the BIT low recirculation flow alarm and other alarms resulting from maintenance work being performed on the unit prevented the operator from discovering that recirculation was isolated until 2:25 p.m. that afternoon.
(Shif t lamp checks had been changed to daily checks due to the fuses burning out during the lamp checks).
l (4) The cancellation of Standing Order No. 0165 which required management verification when doing any BA valve lineups or BA transfers.
This removed an opportunity to discover the error, c.
Safety Significance The BIT provides sufficient negative reactivity to the reactor by injection of 11.5 to 13 weight percent boric acid to prevent
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reactor restart after a steamline break.
For this analysis, the l
full contents of the tank (900 gallons) are assumed to be available
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for injection into the Reactor Coolant System (RCS). The BIT is to be maintained in a 100% full condition whenever the plant is
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critical. A recirculation path is set up to recirculate the BIT l
contents to and from the BAT. Since there is no level indication on l
the BIT, the recirculation flow path ensures that the BIT is always l
100% full.
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No indications of tank level changes were observed on the strip
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chart for 0A and OB BAT during the period from July 20 to July 23,
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1986. In addition, a review of work requests and log books showed no abnonnal leakage during this time frame indicating that the BIT remained full; consequently, there was no technical safety
significance associated with the loss of BIT recirculation.
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Recirculation of the BIT was re-established approximately 50 minutes after discovery. The estimated time between the initial low flow condition (time when BAT pump was tagged 00S) and restoration was five hours, ten minutes.
d.
Corrective Actions Taken By the Licensee (1)
Individuals involved were instructed as to the significance of reviewing existing system lineups prior to performing any additional 00S lineups.
(2) Procedure change requests to ths System Operating Instructions (S0Is)and00SProcedure, ZAP 14-51-2havebeensubmittedto ensure that independent verification of BA System valve lineups by shift management will be performed in the field.
e.
Enforcement Technical Specification (TS) 3.8.1.E.2 states in part, "The boric acid solution shall be in continuous recirculation through the tank
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except during testing...". TS 3.0.3 requires that action be initiated within one hour to place the unit in Hot Shutdown within the following four hours if a Limiting Condition for Operation and/or associated Action requirements cannot be satisfied. On July 23, 1986, for a period of about five hours, the boric acid solution was not in continuous recirculation through the Unit 1 boron injection tank due to an improper valve alignment and no action was taken to place the unit in Hot Shutdown.
This event was reviewed during an enforcement board held in Region III and classified as a Severity Level IV violation (295/86022-01).
One violation and no deviations were identified.
5.
0)eration Of Units 1 And 2 With Incorrect Negative Flux Rate Reactor Trip (iFRT) Setpoints A meeting between the licensee and Region III personnel was held on April 29, 1986 to discuss event chronology and corrective actions which had already been initiated by the licensee.
Following the meeting, an action plan was developed by Region III in which Region III would:
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Determine whether an Enforcement Board should be convened immediately
or be delayed until assessment of safety significance was available, Determine whether this issue applied to Prairie Island, which appeared
to be the only other plant in Region III which may have been subject m>
to this problem, E i
Determine whether an Information Notice should be written regarding P
inadequate review and validation of safety analyses provided to
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license holders by NSSS suppliers,
Request, in the cover letter to Inspection Reports (IR) No. 295/86005;
No. 304/86005, that the licensee respond to the issue indicating their actions to prevent recurrence and assessing whether this issue applied to other Ceco plants, Conduct an inspection at Zion in order to obtain sufficient
information to request the Office of Nuclear Reactor Regulation m
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(NRR) to provide an assessment of the safety significance of this
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issue, and
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Determine whether a team inspection of the licensee's 10 CFR
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Part 50.59 review methods and practices is necessary.
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It was subsequently determined that an Enforcement Board should not be cr.nvened until after an assessment of safety significance by NRR.
A review of this issue at the Prairie Island plant revealed that their NFRT setpoints had been reset per Westinghouse guidance for several years.
To this date, no Information Notice has been deemed necessary since this does not appear to be a generic concern.
In their June 27, 1986 response to IR No. 295/86005; No. 304/86005, the licensee stated that a descriptive listing of all pertinent FSAR parameters, t,(
assumptions, and analytic methodologies would be prepared.
This list was y
expected to be finalized by early 1987.
Such a list was already available to the Ceco boiling water reactors (BWR), and, if the results were considered acceptable at Zion, then Byron and Braidwood would be considered for a similar program.
The licensee also stated that the Vestinghouse setpoint study which originally identified this item would be completed.
Followup of the licensee's corrective actions is considered an Open Item pending NRC review (295/86022-02; 304/86020-01).
In a memorandum dated May 6, 1986, Region III requested NRR to review this issue to determine the safety significance of operation with improper NFRT setpoints.
NRR responded on August 18, 1986 indicating that operation in this condition was not a significant safety problem.
In a meeting on September 9, 1986, Region III representatives met and oetermined that a violation did occur but that this issue should not result in escalated enforcement.
s 10 CFR Part 50, Appendix B, Criterion III requires that measures shall be established to assure that applicable regulatory requirements and the
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design basis for systems to which Appendix B applies are correctly trans-lated into specifications and procedures.
Contrary to this requirement, the licensee failed to establish measures to assure that the correct basis for the setpoints of the NFRT reactor trip were translated into the Zion Technical Specifications (TS) and into calibration Procedures No. N41E, No. N42E, No. N43E, and No. N44E, " Power Range Nuclear l
Electronics," in that the TS and the Calibration procedures specified l
NFRT setpoints of -15 percent power change with a 5 second time constant instead of -5 percent power change with a 2 second time constant as used in the Westinghouse design basis.
This is considered a violation (295/86022-03; 304/86020-02).
l One violation and no deviations were identified.
One Open Item was identified.
6.
Inspection Of Unit 1 Main Steam (MS) Check Valves Following The Failure Of The 1B Check Valve
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On May 8, 1986, Unit 1 was shut down from 55% power due to the failure of the 18 MS check valve.
This event was documented in Inspection Reports No. 295/86011; No. 304/86010.
During the current Unit 1 refueling and maintenance outage, the licensee opened the 1A, 10, and 1D MS check valves for inspection and, if necessary, repair.
The licensee's inspections were reviewed by NRC inspectors from the Region III office, I
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Operational Programs Section, Division of Reactor Safety (DRS).
The DRS review indicated that the apparent mode of failure of the 1B MS check valve was the failure to install a locking plate which should have prevented backing out of two bolts which joined the two halves of the split locking device.
This may not have been the cause of failure, since the locking plate was installed in the 1A MS check valve, which was found to be in good condition although some separation of its split locking device was observed.
The ID valve was found to be significantly degraded, and the 1C valve was degraded, but not to the extent of the 10 valve.
A review of the system configuration indicated that a main steamline break upstream of the main steam isolation valves (MSIV'S) coupled with a failure of the corresponding MS check valve and a failure of one MSIV on a different loop to close could result in blowdown of two steam generators to the containment.
This is due to the fact that the Y type globe valves used at Zion are not designed to operate under reverse ficw conditions with a differential pressure greater than 35 psig.
For this to occur, a main steamline break would have to occur between the steam generator and the failed main steamline check valve coupled with a failure of one of the remaining three MSIVs.
Flow could then occur from an intact line through the faulted MSIV and then in reverse down the line with the line break.
The licensee stated that a preliminary analysis indicated that the containment was capable of withstanding the pressure spike and the primary system could withstand the rapid cooldown resulting from a
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blowdown of two steam generators, assuming that all trains of ECCS and containment cooling were available (this assumption is valid using single failure criteria, e.g., active failure - MSIV, passive failure -
steamline break).
However, the licensee is taking no credit for this analysis, and is therefore required to maintain these check valves operable.
The check valves on Unit 1 will be repaired prior to restart from the current refueling outage.
Since the licensee is operating within the design basis of their Updated Safety Analysis Report (USAR),
NRR has chosen not to review the preliminary analysis.
Since the four check valves installed in Unit 2 were manufactured and delivered by the same company that produced those in Unit 1, a potential exists that Unit 2 valves will experience a similar failure.
All eight of these valves, plus one spare which was used to replace the IB valve in May 1986, were manufactured in the mid-1970s to the same specifications.
They have not been subject to visual inspection since installation.
The licensee stated that although all of the parts of the check valves were not installed as specified on the vendor drawings, they had confidence that the check valves would be operable for one closure on Unit 2.
Inspection of the valves in the "as found" condition on Unit 1 supported this conclusion.
In addition, the licensee has committed to a comprehensive testing program in their September 29, 1986, letter from Mr. M. S. Turbak to Mr. J. G. Keppler.
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These surveillance tests would be conducted on Unit 1, which was operating at the time, to ensure that the check valves were installed and were capable of performing their function.
These tests include:
measuring back leakage through the valve to ensure that it could withstand back flow with the unit shutdown; measuring the differential pressure across the check valve during operation to provide evidence that the valve disc had not failed such that the disc had separated from the tail linkage thereby restricting steam flow; and the use of radiography to verify that the valve disc was, in fact, attached to the tail linkage.
Test results indicated that the valves could prevent back flow in the steam line, and would were not significantly affect steam flow, and that the valve discs were attached to the tail linkages.
The licensee also committed to shut down Unit 2 on any indication of valve failure, either from radiography or flow reduction.
The licensee stated that they will continue to perform a main steam check valve leak test after each turbine trip to ensure that the valve will fulfill its function.
tiRC will review the results or the Unit 2 MS check valve inspections.
No violations or deviations were identified.
7.
September 20, 1986, Unit 2 Reactor Trip Due To Valve Alignment Error Following Instrument Maintenance
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On September 20, 1986, Unit 2 tripped from approximately 20% reactor power shortly af ter the main generator was tied to the grid.
The trip occurred when an instrument mechanic (IM) inadvertently left a turbine
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impulse pressure transmitter, 2PT MS-24, valved out of service following maintenance.
The pressure transmitter should have provided a feedback signal to the turbine electro-hydraulic control (EHC) system. With the pressure transmitter valved out of service, the turbine load increased until the reactor tripped on low steam generator level coincident with a mismatch between steam flow and feed water flow.
An inspector from the Region III office was dispatched to the site to follow the licensee's investigation.
The results of his inspection were documented in Inspection Reports No. 295/86023; No. 304/86022.
No violations or deviations were identified.
8.
September 21, 1986, Degraded Grid Voltage On September 21, 1986, the voltage on bus 247 decreased to below 3846 volts (degraded grid condition).
After five minutes in this condition, the bus 247 undervoltage relay energized, the loads on bus 247 were stripped and the "0" diesel generator (DG) start sequence was initiated as designed.
Before reaching rated speed and voltage, the "0" DG tripped, and the bus 247 loads remained de-energized until they were manually re-energized by operations personnel.
An inspector from the Region III office was dispatched to the site to follow the licensee's investigation.
The results of his inspection were documented in Inspection Reports No. 295/86023; No. 304/86022.
No violations or deviations were identified.
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9.
September 24, 1986, 10 CFR Part 21 Report Regarding Anderson-Greenwood Five Valve Manifolds On September 24, 1986 leakage was found to be coming from the five valve manifold for 2LT-461, the Unit 2 pressurizer level channel.
The five valve manifolds are manufactured by Anderson Greenwood, Model No. DPMHS-4B2-P, assembly No. 02-8248-522.
It was found that the cause of the leakage was that the valves were loose at the point at which they were threaded into the manifold.
l On September 26, 1986, the licensee reported the problem to the NRC pursuant to 10 CFR Part 21.
A written report was issued by the licensee on September 30, 1986, under Licensee Event Report (LER) No. 304/86-017.
The apparent cause of the leakage was believed to be inadequate torquing of the valves at the factory, since there was no record of the valves having been adjusted during installation.
IE hat evaluated the problem and has determined that there is no applicability to any other plant and in addition the licensee continues to investigate the problem in regards to their other sites.
The licensee's corrective actions are included in LER No. 304/86-017.
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I No violations or deviations were identified.
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10.
September 30, 1986, Response To The Effects Of Flooding In Northeast
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Illinois As a' result of significant rainfall in Northeastern Illinois during_the
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L week of September 12, 1986, flooding occurred within a 10-mile radius
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around Zion Nuclear Generating Station.
As a result, the State of Illinois expressed concern regarding the availability of emergency 1-
evacuation routes for Zion.
Several conversations between Region III,
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the State of Illinois and Commonwealth Edison Company took place regarding e
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this-concern.
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The flooding had no effect on the operation of Zion Station.
Unit I was shut down for refueling, and Unit 2 operated at 99 percent power.
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Based on information from Commonwealth Edison,_ Region III contacts
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I with the Lake County Emergency Services and Disaster Agency (ESDA),
FEMA-Region V, the U. S. Army Corps of Engineers, and the direct
observations by NRC resident inspectors, it was determir.ed that most roads were open and sufficient evacuation routes were available.
Commonwealth Edison Company initiated several conservative actions, including 24-hour monitoring of the situation from their Emergency Gperations Facility, located near the Zion Nuclear Generating Station,
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and the placement of an additional Nuclear Station Operator (NS0) in the Control Room to assist in the event of a plant transient.
Unit 2 was maintained at steady state conditions to minimize the potential for
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transients or a reactor trip.
In addition, the licensee postponed all unnecessary instrument surveillances and testing, and limited access to
' equipment, the inadvertent disturbance of which could cause a turbine or
These actions were relaxed during the week of September 17,
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1986 as the water levels subsided opening the previously flooded
evacuation routes.
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No violations or deviations were identified.
i 11.
October 16, 1986, Fuel Building Ventilation $1 stem Filters Bypassed On October 16, 1986, while performing Procedure No. PT-19, " Fuel Building i
Ventilation Test," the access door to the fuel building filter plenum was
found open (normal position closed) causing the ventilation flow to
bypass the pre-filter and the HEPA filter.
This issue was referred to a regional based health physics inspector who will be following this event
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in his next inspection which is scheduled for November - December 1986.
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No violations or deviations were identified.
12.
October 24, 1986 Failure Of The 1B Diesel Generator During Post-Maintenance Testing On October 24, 1986, while conducting post-maintenance testing of the IB diesel generator (DG), the articulating connecting rod for cylinder 4-left
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was thrown through the side of the engine along with one of two crankshaft counterweights for cylinders 4-left and 4-right.
At the time, Unit 1 was in cold shutdown (mode 6) and the DG was not required to be operable per the Technical Specifications (TS).
No fire or personnel injuries resulted.
The licensee had performed an overhaul of the IB DG during the current Unit 1 refueling cutage, and was running the engine to verify proper operation prior to post-modification testing for the associated safeguards bus.
An operator and an engineer were in the room when they heard loud noises coming from the engine.
The engineer directed the operator to shut down the engine, and the operator pressed the emergency stop pushbutton.
Both individuals immediately left the room as the coa 3tdown began. While they were exiting the DG room door, the articulating rod and counterweight were thrown through the engine, and the room filled with steam and oil spray from the jacket water and lube oil systems.
The licensee quarantined the room until an investigation into the root cause of the failure could be initiated.
On October 25, 1986, in a telephone conversation with Region III, the licensee agreed to certain restrictions on the cleanup and investigation of the cause of the failure, including actions to permit adequate inspection opportunity by Region III inspectors before disassembly or repair of the engine.
This was confirmed in writing in a Confirmatory Action Letter (CAL) issued to the licensee on October 27, 1986.
On October 27, 1986, three inspectors from Region III were sent to the site to observe the licensee's investigation.
This event will be covered in detail in their Inspection Reports (No. 295/86026; No. 304/86026).
No violations or deviations were identified.
13. Operational Safety Verification and Engineered Safety Features System Walkdown The inspectors observed control room operations, reviewed applicable logs and conductea discussions with control room operators from September 26 through November 17, 1986.
During these discussions and observations, the inspectors ascertainea that the operators were alert, fully co0nizant of plant conditions, attentive to changes in those conditions, and took prompt action when appropriate.
The iaspectors verified the operability of selected emergency systems, reviewed tagout records and verified proper return to service of affected components.
Tours of the auxiliary and turbine buildings 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.
l The inspectors, by observation and direct interview, verified that selected j
physical security activities were being implemented in accordance with the station security plan.
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The inspectors observed plant housekeeping / cleanliness conditions and verified implementation of radiation protection controls.
From September 26, 1986 to November 17, 1505, the inspectors walked down the accessible portions of the 2A and 2B diesel generators, and the auxiliary feed water systems to verify operability.
These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established under Technical Specifications, 10 CFR and administrative procedures.
The inspectors noted that good progress has been made in painting the turbine building.
No violations or deviations were identified.
14.
Monthly Surveillance Observation The inspector reviewed Technical Specification required surveillance testing in the control room 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 personnel other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel.
The inspector reviewed portions of the following test activities:
PT-0, Shift Surveillance Checklists
No violations or deviations were identified.
15.
Monthly Maintenance Observation Station maintenance activities on safety related systems and components listed below were observed or reviewed to ascertain whether they were conducted in accordance with approved procedures, regulatory guides industry codes or standards and in conformance with Technical Specifications.
The following items were considered and verified during this review: that the limiting conditions for operation were met while components or systems were removed from service, that approvals were obtained prior to initiating the work, that activities were accomplished using approved procedures and were inspected as applicable, that functional testing and/or calibrations were performed prior to returning components or systems to service, that quality control records were maintained, that activities were accomplished by qualified personnel, that parts and materials used were properly certified, that radiological controls were implemented and that fire prevention controls were implemented.
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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.
The following maintenance activities were observed or reviewed:
1B Diesel Generator repair
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Observation of a demonstration on inserting incore thimbles
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Observation of Unit 1 and Unit 2 containment dome repair The following procedures were reviewed:
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RC00-12, " Retracting and Inserting Incore Instrumentation Thimbles"
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Wm. A. Randolph, Inc. Procedure No. QCP-10, " Grouting Procedure,"
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Wm. A. Randolph, Inc. Procedure No. QCP-28, " Grouting Test."
An inspection of completed grouted patches on Unit 2 and the saw cut edges of " bad concrete" on Unit 1 were examined by the resident inspector.
In conjunction with the inspection of the grouted containment patches, a review of Procedure QCP-10 raised a concern as to whether the appropriate revision to the procedure was being used after is was learned by the resident inspector that the contractor had been utilizing Re/ision 3.
Discussions between the inspector and Station Contractor (SC) and Quality Assurance (QA) personnel regarding the use of Revision 3 disclosed that final approval and acceptance of a contractor procedure is contingent upon a technical review performed by Sargent & Lundy (S&L), the licensee's Architectural Engineer (AE). Subsequent to that review, S&L classifies the contractor procedure. Status I classification means that the procedure is approved and accepted whereas a Status 2 classification indicates the procedure is to be revised as noted by S&L.
Procedures classified either a Status 1 or Status 2 can be used on a job; however, a contractor can proceed with a Status 2 procedure only after incorporating the revisions noted by S&L. The procedure in question, No. QCP-10, Revision 3, had been classified as Status 2, and the appropriate revisions to the procedure had been incorporated prior to use on the subject job.
The QA Department controls the issuance of contractor procedures and receives a monthly listing of the current status of all revisions to contractor procedures.
Classification of procedures by the AE and their subsequent use by a contractor will be considered unresolved pending NRC review.
(304/86020-03).
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A second concern regarding the grouting of containment patches resulted from a review of the work request package for Unit 2,(2-50326),where it was noted that the Quality Control (QC) hold point for observing the grouting of the containment patches was not signed at the time it was observed and not until approximately two weeks after the job was completed.
The QC inspector stated that he had observed the grouting of the Unit 2 containment patches and that he felt the hold point included the entire dome grout repair work, not any one particular step.
This will be considered an Unresolved Item pending additional NRC review of the work package and to determine the purpose of the hold point in question (304/86020-04).
No violations or deviations were identified.
Two Unresolved Items were identified.
16.
Licensee Event Reports (LER) Followup Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with Technical Specifications.
The LERs listed below are considered closed:
UNIT 1 LER N0.
DESCRIPTION 86010-01 Removed Aircraft Crash Fire Damper on Cribhouse 86019-01 Incomplete Review of Fuel Handling Procedure Change 86028-00 Inadvertent Isolation of Boron Injection Tank Recirculation 86030-00 Degraded Fire Barriers 86031-00 Failure of Temporary Procedure to Receive 14 day On-site Review 86032-00 Engineered Safety Feature (ESF).
Actuation while Performing Test 86034-00 ESF Actuation due to Accidentally Bumping Contact while Troubleshooting 86036-00 Safety Related Mechanical Snubber Inoperable for More Than 72 Hours
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UNIT 2 LER NO.
DESCRIPTION 86006-01 Mis-wiring of Reactor Trip Switchgear during Modification 86015-00 Inadvertent ESF Actuation during Safeguards Testing 86017-00 Anderson-Greenwood Five Valve Manifold Leakage on 2LT-461 (10 CFR Part 21 Report)
86018-00 Secondary Undervoltage on Essential Service Bus 86019-00 Reactor Trip Due to Personnel Error Installing Non-Safety Related Transmitters Regarding LERs No. 295/86010-01 and No. 295/86030-00, " Removed Aircraft Crash Fire Damper on Cribhouse Ventilation Fan" and " Degraded Fire Barriers" respectively, these events will receive additional review under existing Unresolved Item, No. 295/86019-05.
See Inspection Reports No. 295/86019; No. 304/86018 for further details.
LER No. 295/86028-00, " Inadvertent Isolation of Boron Injection Tank Recirculation", is considered a Violation.
See Paragraph 5 of this report and Inspection Reports No. 295/86019; No. 304/86018 for further details on the event.
Regarding LER No. 295/86036-00, " Safety Related Mechanical Snubber Inoperable for More Than 72 Hours", the licensee's proposed corrective action included a revision to procedure TSS 15.6.48, " Hydraulic and Mechanical Snubbers Surveillance".
The revision will add a list of all snubbers which are required to be available during plant Modes No. 5 and No. 6 to the current procedure.
This will remain an Open Item pending NRC review of the revised procedure (295/86022-04; 304/86020-05).
LER No. 304/86006-01, " Mis-wiring of Reactor Trip Switchgear during Modification" is closed.
Corrective action was reviewed and found to be above average.
See Paragraph 2 of this report for details on the event.
Regarding LER No. 304/86018-00, " Secondary Undervoltage on Essential Service Bus" and No. 304/86019-00, " Reactor Trip Due to Persornel Error Installing Non-Safety Related Transmitters", these events were covered in detail in Inspection Reports No. 295/86023; No. 304/86022.
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No violations and no deviations were identified.
One Open Item was identified.
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Training
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During the inspection period, the inspectors reviewed events and unusual occurrences which may.have resulted, in part, from training deficiencies.
Selected events were evaluated to determine whether the classroom, simulator, or on-the-job training received before the event was sufficient to have either prevented the occurrence or to have mitigated its effects by recognition and proper operator action.
Some qualifications were also evaluated.
In addition, the inspectors determined whether lessons learned from the events were incorporated into the training program.
Events reviewed included the events discussed in this report.
In addition, LERs were routinely evaluated for training impact.
One department head meeting was attended by the resident inspectors, as
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well as a meeting between station management and supervisory personnel and the CECO Manager of Quality Assurance.
No violations or deviations were identified.
18.
Followup of IE Bulletins In a memorandum from Charles E. Norelius dated August 22, 1986, all resident inspectors were requested to review the licensee's response to IE Bulletin No. 86-02, " Static "0" Ring (Model No. 102 or No. 103)
Differential Pressure Switches" to include (a) whether the licensee's response provided the information required to be reported (b) whether the licensee's response was provided by the date specified in the Bulletin and (c) whether the licensee's corrective actions in response to the Bulletin were, app'ropriate.
In accordance with the above request the resident inspector reviewed the licensee's response to the bulletin.
Static "0" Ring (SOR) Model No. 102 or No. 103 differential pressure switches are not installed in a safety related electrical equipment capacity at the station.
Bulletin No. 86-02 requested that the scope of licensee's reviews include systems important to safety as specified in 10 CFR Part 50.49b. Although such a review was performed, the licensee's response dated July 25, 1986, stated that no SOR Model No. 102 or No. 103 differential pressure switches were installed in safety related applications.
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No violations or deviations were identified.
19.
Followup Of Region III Requests a.
10 CFR Part 21 Report on SOR Inc. Pressure Switches In a memorandum from W. G. Guldemond dated August 27, 1986, resident inspectors were directed to review the licensee's response to a
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. 10 CFR Part 21 report by Static "0" Ring (SOR),LInc.
The report
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concerned potential changes in allowable operating repeatability
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. ranges of' gauge pressure switches.with design designators beginning with number 9, 8, or 1.
The SOR Model No. 9TA-B45-NX-CIA-JJTTX7 was
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listed in the report as having been supplied to Zion Station. On October.8, 1986, the Station Nuclear Engineering Department (SNED)
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issued.a close-out report relative to the.50R Inc., repeatability problems.- ;It'was determined by the SNED that Zion Station utilizes no SOR. class 1E qualified Series 9, 8, and 1 pressure switches in its EQ Program.
This SOR switch series is used for the main steam isolation valve
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closing accumulator high/ low nitrogen pressure alarms.
The switches (PS-MS118, MS119, MS120, and MS121) are used exclusively for alarm o
P purposes 'and are not environmentally qualified or Technical Specification.related.
A separate transmitter provides pressure
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values available'on computer and a. local gauge provides nitrogen pressure indication as a backup.
The auxiliary operc. tor checks one of these indications every 8-hours on his rounds.
The licensee intends to replace these switches with a model which is not subject to the problem mentioned above.
The installation of the new switches is currently scheduled for the 1986 Unit 1 and 1987 Unit 2 outages.
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'b.
10 CFR Part 21 Report On Valcor Valve Spring Failures (IE Information Notice 86-72)
3.
In a memorandum from Charles E. Norelius dated October 28, 1986, resident inspectors were directed to review the licensee's response to a 10 CFR Part 21 report by Valcor Engineering Corporation regarding the failure of 17-7PH stainless steel springs in the valve internals which were in contact with primary coolant in pressurized water reactors.
Spring failures have occurred and have been attributed
'to hydrogen embrittlement. This matter was also the subject of IE Information Notice 86-72.
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The licensee uses the subject valves in the reacter vessel head vent
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and reactor vessel level systeins.
These valves are in contact with primary coolant, but are normally closed during operation and there is no process flow through the valves.
The licensee plans to have Valcor representative onsite during the current Unit 1 outage in l
order to replace the springs with ones which are not subject to hydrogen embrittlement.
Unit 2 valves will be examined and the subject springs will be replaced during the next refueling outage i
(scheduled for Spring 1987).
This is considered an Open Item
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pending NRC review of the results of the licensee's actions (295/86022-05; 304/86020-06).
c.
IE Information Notice (IN) 86-53, Improper Installation of Heat Shrinkable Tubing j
In a memorandum from Charles E. Norelius dated October 28, 1986, resident inspectors were directed to determine the extent of i
deficient splices involving heat shrinkable tubing at their i
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respective sites in response to IE IN No. 86-53.
IN No. 86-53 reported several types of splice deficiencies due to installation errors, which included insufficient tubing length, severe bending of the splice, splicing over braided jackets, ana bare conductors exposed. The splices in question were required to be environmentally qualified (EQ) in accordance with 10 CFR Part 50.49.
A list of the results of EQ splice inspections at Zion follows:
Number of 5)1 ices:
Category Unit 1 Jnit 2 Found qualified at the time of
92
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the inspection (installed properly per manufacturer's procedures)
Splices not installed properly, but-
38 subsequently qualified by special testing of the installed configuration Splices not installed properly, but
34 replaced per manufacturer's procedures prior to special configuration testing Total splices inspected
164 Total splices on EQ list 288 270 The licensee also identified EQ instrumentation wires landed on terminal blocks in areas subject to harsh environment.
In some cases, the landed leads were in addition to splices identified on the EQ splice list, and in other instances, the landed leads were found in place of EQ splices.
To date approximately 9 instances of this condition have been identified for Units 1 and 2.
In this configuration, leakage currents may be excessive and may result in erroneous information to operators if the terminal blocks were to be subject to harsh environment.
The licensee has also identified 7 examples of non-EQ heat shrink tubing in locations where EQ splices should have been installed.
One such splice was opened for inspection, ar.d it was found that the non-EQ tubing was installed over a pair of splices which were either acceptably installed or installed in a configuration which was qualified by subsequent special testing. All such splices were found on Unit 1 which is currently in a refueling outage.
The licensee will open and re-splice all 7 non-EQ heat shrink splices.
The licensee's inspection program for EQ splices will be complete before the Unit 1 startup in February 1987.
At that time all EQ splices will be either acceptably installed or installed in a configuration which was qualified by subsequent special testing.
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Inspection Of Seismic Monitoring Instrumentation In a memorandum from W. G. Guldamond dated August 29, 1986, resident inspectors'were directed to obtain'information regarding seismic i
monitoring instrumentation at their respective sites. The
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information for Zion is as follows:
System description: The seismic monitoring' instrumentation was
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not described in.the FSAR. Copies of selected pages of the applicable vendor manual were provided to Region III.
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Technical Specifications:. The seismic monitoring instrumentation is not covered in the Technical Specifications
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(TS). A TS amendment is planned to include this system at some future date.
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Non-TS surveillances performed: The licensee does not perform any surveillances on,the seismic monitoring system, except for the passive seismic monitors which are' checked.each Unit 2-
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refueling outage per Procedure No. 0X-ME 256 and No. 257.
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Preventive maintenance (PM) performed:.The licensee does not perform any PMs on the seismic monitoring system.
Failure data for the,last 24 months: Five work requests (WR)
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were identified for the seismic monitoring system for 1986.
Data before 1986 was unavailable at the time of the inspection.
The WRs were typically initiated as a result of spurious actuations or alarms for the system.
Resident inspectors were directed to perform annual inspections of the seismic monitoring system. This will be incorporated into the master inspection plan for Zion, e.
National Survey On Steam Driven Auxiliary Feedwater (AFW) Pumps In a memorandum from Edward G. Greenman dated October 24, 1986, resident inspectors were directed to obtain information regarding steam driven auxiliary feedwater (AFW) pumps due to an event that
. The resident inspectors occurred at the Indian Point-2 plant.
determined that:
The steam driven AFW pumps for Zion Units 1 and 2 are fully
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automatic
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The steam driven AFW pump is powered by a Terry turbine, and There is no relief valve on the steam supply line between the
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stop and governor valves to the AFW pump turbine as was the case at Indian Point 2.
A 10 psig relief is located on the Terry turbine discharge, and a steam line drain trap relief exists upstream of the turbine inlet valve, but neither of these reliefs has the potential to impact operation of the turbine.
_ _ _ _ _,. _. _,..... -., _. -.... _ _ _
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o This completes the requested actions.
f.
Use Of ITT Barton Model 288A Differential Pressure Devices In a memorandum from R. D. Walker dated August 26, 1986, Region III was inforned of a potential generic problem with Barton Model No. 288A differential pressure switches.
These switches had experienced repeated drift problems at Sequoyah.
Resident inspectors were requested to determine the extent to which these switches were in use at their respective sites.
The following is a list of uses of these switches at Zion:
System Parameter Function Component Cooling Reactor Coolant pump Indication, Alarm (RCP) motor upper-radial bearing flow RCP thermal barrier flow Indication, Alarm, Control l
Condensate Condensate pump Indication, Alarm suction differential pressure (DP)
Condensate pump seal Indication, Control water injection filter DP Auxiliary Auxiliary Feedwater Indication, Alarm Feedwater pump suction Reactor Coolant RTD bypass flow Indication, Alarm Loop stop valve Indication, Control low flow Volume Batching tank level Indication, Alarm Seal water injection Indication, Alarm filter DP Waste Disposal Carbon column tank DP Indication, Alarm This completes the requested action, g.
Clarification of Reporting Requirements of 10 CFR Part 50.72 and 50.73 In a memorandum from Charles E. Norelius, resident inspectors were directed to discuss with their respective licensees a clarification of reportability pursuant to 10 CFR Part 50.72 and 10 CFR Part 50.73.
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O On September 5, 1986, the Quad Cities station discovered that the discharge valve on the reactor core isolation cooling system (RCIC)
pump would not open remotely.
This failure, by itself, would have l
prevented RCIC from performing its design function.
Historically, I
the licensee has reported RCIC system failures pursuant to
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10 CFR 50.72 and 50.73; however, following the September 5 event, they questioned whether the event was reportable since RCIC is not a system for which their accident analysis takes credit and is not l
needed for safe shutdown.
On September 5, this question was posed to Mr. J. Rosenthal, Chief, Events Analysis Branch, IE and Mr. E. Weiss of his staff.
Their response was that RCIC inoperability alone, for whatever reason, was not reportable pursuant to either 10 CFR 50.72 or 50.73 provided that the licensee has not taken credit for RCIC as either an Engineered Safety Feature (ESF) or in their safe shutdown analysis.
This clarification was provided to the licensee on October 30, 1986.
,is completes the required action.
No violations or deviations were identified.
One Open Item was identified.
20.
Followu) of Licensee Actions In Response To Three Mile Island (TMI)
Action )lan Items The following TMI Action Plan items were inspected to verify that the licensee's response was provided in the time required, that the licensae's response was technically adequate, and that any necessary changes were implemented correctly.
a.
II. K. B. 10 - Proposed Anticipatory Trip Modifications This item was applicable to owners of Westinghouse NSSS design plants which had proposed modifications to the anticipatory reactor trip logic.
In a letter dated December 15, 1980, the licensee stated that the anticipatory trip is active above 10% power as required by Technical Specification (TS) 3.1, that there were no plans to revise the set point, and that this item was, therefore, not applicable to Zion Station.
b.
II. K. B. 12 - Confirm Existence of Anticipatory Reactor Trip Upon Turbine Trip In a letter dated December 15, 1980, the licensee confirmed the anticipatory reactor trip upon turbine trip as described in Section 7.2.2 of the Zion FSAR and Section 3.1 of the TS.
The licensee also stated that no modifications were proposed or required.
In a letter dated September 21, 1981, from the Division of Licensing, NRR, the NRC stated that the licensee's responses to these items were considered acceptable and that Items No. II.K.3.10 and No. II.K.3.12 were considered resolved.
The inspectors have verified the setpoints of these anticipatory trips on several occasions.
These items are closed in this report for record purposes only.
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o No violations or deviations were identified.
21.
Site Visit By The Advisory Committee On Reactor Safeguards (ACRS) And Their Counterparts From Japan And France On October 23, 1986, the licensee conducted a site tour for members of the NRC Advisory Committee for Reactor Safeguards (ACRS), their spouses, ACRS staff members, and members of their counterpart organizations from France and Japan.
The tour was arranged as part of a meeting between these organizations which was held in Racine, Wisconsin. About 30 persons participated in the tour.
The Senior Resident Inspector attended the introductory briefings conducted by the plant management and answered questions posed by the group about the NRC in general and the resident inspector program.
No violations or deviations were identified.
22.
Site Visit By Members of the Kenosha County Board On November 1, 1986, the licensee conducted a site tour for members of the Kenosha County Board.
The tour included the control room and the turbine building.
In addition, a meeting was held to present information about Zion Station. The Senior Resident Inspector attended the meeting and answered questions relating to the NRC in general and the role of resident inspectors.
No violations or deviations were identified.
23.
Followup of of Information Notices The following IE Inforination Notices (IN) were examined to determine whether the licensee has received the IN, whether site and corporate distribution were appropriate and whether the licensee's response was adequate:
IN TITLE 86-02 Failure of Valve Operator Motor During EQ Testing 86-07 Lack of Detailed Instruction During Maintenance and Testing of Woodward Governors 86-14 Auxiliary Feedwater Pump Turbine Control Problem 86-58 Dropped Control Rod No violations or deviations were identified.
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Open Items Open Items are matters which have been discussed with the. licensee which will be reviewed further by the. inspector and which involve some action on the.part of the NRC-or licensee:or both.
Three Open Items disclosed during this inspection are discussed in Paragraphs 5, 16, and 19.
25.. Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acccptable items, deviations or items of noncompliance.
Two Unresolved Items disclosed during this inspection are discussed in Paragraph 15.
26.
Exit Interview The inspectors met with licensee representatives (denoted in Paragraph 1)
throughout the inspection period and at the conclusion of the inspection on November 17, 1986 to summarize the scope and findings of the inspection activities.
The licensee acknowledged the inspectors' comments.
The inspector also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspector during the inspection.
The licensee did not identify any such documents or processes as proprietary.
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