IR 05000327/1986060
| ML20212J968 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 01/06/1987 |
| From: | Carroll R, Harmon P, Jenison K, David Loveless NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20212J922 | List: |
| References | |
| 50-327-86-60, 50-328-86-60, NUDOCS 8701280392 | |
| Download: ML20212J968 (21) | |
Text
p ub,D UNITEo STATES o
NUCLEAR REGULATORY COMMISSION
["
REGION 11 n
g j
101 MARIETTA STREET.N.W.
- I C
ATLANTA, GEORGI A 30323
\\;..... /
Report Nos.:
50-327/86-60 and 50-328/86-60 Licensee:
Tennessee Valley Authority 500A Chestnut Street Chattanooga, TN 37401 Docket Nos.:
50-327 and 50-328 License Nos.:
DPR-77 and DPR-79 Facility Name: Sequoyah Units 1 and 2 Inspection Conducted: October 6, 1986 - November 5, 1986 Inspectors:
l $ r$A n dh/
det
_
Date Signed i 6 l9i K./M. Jenison,penitr Resident Inspector 0 h Ak les f// k 7
<
P) E. Harmon, f}6siddnt' Inspector Date Signed
$ SMb JM ll6 l3 7 P. Loveless,uResident Inspector Date Signed Approved by
[N e YA
//M87 R. Carroll, Acting chaff, Section IA Date Signed
~
Division of Reactor Projects SUMMARY Scope:
This routine, announced inspection involved inspection onsite by the Resident Inspectors in the areas of: operational safety verification (including operations performance, system lineups, radiation protection, safeguards and housekeeping inspections); maintenance observations; review of previous inspection findings; followup of events; review of licensee identified items; review of IE Information Notices; and review of inspector followup items.
Results:
One violation was identified for failure to accomplish activities affecting quality in accordance with prescribed documented instructions, pro-cedures, or drawings of a type appropriate to the circumstances which led to the violation of TS 3.3.3.7.
This violation was a second example of Violation 327, 328/86-19-06 and therefore a separate Notice of Violation will not be issued -
paragrah 8.b.
Three unresolved items were identified:
paragraph 5.c (327,328/86-60-09)
Failure to control a Locked High Radiation Area paragraph 7.c (327,328/86-60-10)
Surveillance Instruction Review Findings paragraph 8.c (327,328/86-60-11)
Control of Safety-Related Pumps No deviations were identified.
8701280392 870116 PDR ADOCK 05000327 PDR G
i
.
.
REPORT DETAILS 1.
Licensee Employees Contacted
- H. L. Abercrombie, Site Director
- P. R. Wallace, Plant Manager
- L. M. Nobles, Operations and Engineering Superintendent
- B. M. Patterson, Maintenance Superintendent R. J. Prince, Radiological Control Superintendent
- M. R. Harding, Licensing Group Manager
- W. E. Andrews, Site Quality Manager
,
D. W. Wilson, Project Engineer R. W. Olson, Modifications Branch Manager J. M. Anthony, Operations Group Supervisor
- R. V. Pierce, Mechanical Maintenance Supervisor M. A. Scarzinski, Electrical Maintenance Supervisor
- H. D. Elkins, Instrument Maintenance Group Manager J. T. Crittenden, Public Safety Service Chief
- R. W. Fortenberry, Technical Support Supervisor
- G. B. Kirk, Compliance Supervisor
- D. C. Craven, Quality Assurance Staff Supervisor
- J. H. Sullivan, Regulatory Engineering Supervisor
- J. L. Hamilton, Quality Engineering Manager
- D. L. Cowart, Quality Engineering Supervisor
- H. R. Rogers, Plant Operations Review Staff
- R. C. Burchell, Compliance Engineer R. H. Buchholz, Sequoyah Site Representative Other licensee employees contacted included technicians, operators, shift engineers, security force memoers, engineers and maintenance personnel.
- Attended exit interview 2.
Exit Interview
.
The inspection scope and findings were summarized with the Plant Manager and members of his staff on November 12, 1986.
The violation and unresolved items described in this report's Summary paragraph were discussed.
No deviations were discussed.
The licensee acknowledged the inspection findings. The licensee did not identify as proprietary any of the material reviewed by the inspectors during this inspection.
During the reporting period, frequent discussions were held with the Site Director, Plant Manager and other managers concerning inspection finding.
,
3.
Licensee Action on Previous Inspection Findings (92702)
(Closed) Violation 327/85-35-01.
This violation involved a failure to implement and maintain procedures in accordance with TS 6.8.1.
Three examples were given (response time testing, plant modifications, and functional testing).
The inspector reviewed the licensee's response of January 9,1986 and the corrective actions described therein. This item is closed.
(0 pen) Violation 327/85-45-06.
This violation involved a failure to promptly identify and correct conditions adverse to quality. The inspector reviewed the licensee's responses of February 18 and June 27, 1986, and the corrective actions described therein. A supplemental response described in NRC letter Zech/ White dated September 29, 1986, was also reviewed.
The licensee's corrective actions will be addressed as part of an overall NRC response to several enforcement actions involving inspection reports 327,328/86-19, 86-37, and 86-49. This item is open.
(0 pen) Violation 327/85-45-09.
This violation involved a failure to properly implement required procedures in three instances (valve modifi-cation, air compressor preventive maintenance, and housekeeping).
The inspector reviewed the licensee's response of March 20, 1986, and the corrective actions described therein. The licensee's response to the issues of valve modification and air compressor maintenance is acceptable'.
The issue of housekeeping will be inspected prior to startup.
This item is open.
(Closed) Violation 327/85-46-05.
This violation involved a failure to properly implement required procedures in two instances (relay surveillance, and pen and ink changes to procedures).
The inspector reviewed the licensee's response of February 28, 1986, and the corrective actions described therein. This item is closed.
(Closed) Violation 327,328/84-10-01.
This violation involved failure to store safety related electric motors properly.
The licensee has taken corrective action by performing preventive maintenance on the motors at the time of discovery and by developing a preventive maintenance program for storage motors. The inspector did have questions as to the timeliness of implementation and the completeness of the program.
The inspector's questions will be followed as IFI 327,328/86-60-12.
The violation is closed.
(Closed) Violation 327,328/84-38-03. This violation involved the failure to determine and correct the OT2 swing back handswitch problem.
The hand-switches have been replaced and the site reorganization which was designed to coordinate operational problems between the plant and DNE is in effect.
The violation is closed.
(Closed) Violation 328/84-21-02. This violation involved a technical error in an ISI procedure for RHR pump testing. This item is closed in para-graph 8.c of this repor.
(Closed) URI 327,328/85-46-07 Auxiliary Feedwater (AFW) Pump - Vibration.
AFW _ pump vibration exceeded the acceptance criteria for the Unit 2, AFW System Cavitating Venturi post modification test. The licensee conducted an engineering avaluation to determine if the vibration level was acceptable. The licensee used a spectrum analyzer to measure the vibration levels. As a result of this evaluation, the licensee determined that the process piping was acceptable even under worst case cavitation, but cavitation modes should be avoided for normal operation. The licensee has added cautions to the system operating procedures to reduce operation of the system under cavitating conditions and plans to include the AFW system in the long term Pipe and Vibration surveillance program.
Related pump head and pump curve issues will be evaluated as part of URI 327,328/86-60-11 which is in paragraph 8e of this report.
(Closed) Violation 327/86-20-08.
This item involved licensee failure to verify the position of containment isolation valves per TS. This item is closed in paragraph 7.c of this report.
4.
Unresolved Items Unresolved items are matters about which more information is required to determine whether they are acceptable or may involve violations or deviations. Three unresolved items were identified during this inspection, and are identified in paragraphs 5.c, 7.c and 8.c.
5.
Operational Safety Verification (71707)
,
a.
Plant Tours The inspectors observed control room operations, reviewed applicable logs, conducted discussions with control room operators, observed shift turnovers, and confirmed operability of instrumentatic1.
The inspectors verified the operability of selected emergency systems, and verified compliance with Technical Specification (TS) Limiting Conditions for Operation ( LCO).
The inspectors verified that maintenance work orders had been submitted as required and that followup activities and prioritization of work was accomplished by.the licensee.
s i.
Tours of the diesel generator, auxiliary, control, and turbine buildings, and containment were conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, and excessive vibrations and plant housekeeping / cleanliness conditions A walkdown of the following system was accomplished:
Emergency Control Room Ventilation System - Units 1 and 2 No violations or deviations were identified.
- -
_.
-
_
.
_
.
_
.
.
-,
.. -.
.
-
.
b.
Safeguards Inspection In the course of the monthly activities, the inspectors included a review of the licensee's physical security program. The performance of various shifts of the security force was observed in the conduct of daily activities including protected and vital area access controls; searching of personnel and packages; escorting of visitors; and badge issuance and retrieval; patrols and compensatory posts.
In addition, the inspectors observed protected area lighting, protected and vital areas barrier integrity. The inspectors visited the central alarm station and interviewed security personnel regarding their respective duties.
Three issues involving safeguards information were identified.
The first issue involved the use of supplemental perimeter patrols under certain limited visibility situations.
The inspector observed two Public Safety Officers conversing rather than walking their zones, and found a large area of fence that was unpatrolled as a result. The second issue involved a vital area boundary.
Bars in the overhead of the Control Room had been removed by a modification and allowed access from the Control Building to the Control Room. Compensatory measures are now in effect and will remain in effect until permanent corrective action can be implemented. The third issue involved the process of key carding into a vital area. Regional security specialists were informed of the above issues and will follow up on upcoming inspections.
This item will be tracked as Inspector Followup Item (IFI) 327,328/86-60-01.
No violations or deviations were identified.
c.
Radiation Protection The inspectors observed Health Physics (HP) practices and verified implementation of radiation protection control. On a regular basis, radiation work permits (RWPs) were reviewed and specific work activities were monitorec to ensure the activities were being conducted in accordance with applicable RWPs.
Selected radiation protection instruments were verified operable and calibration frequencies were reviewed.
During the inspection period, the licensee received information concerning misplaced or lost incore instruments at the licenree's Brown's Ferry Nuclear Plant. This prompted an inventory of similar.but smaller incore thimbles at Sequoyah. These thimbles contain approxi-mately 40 milligrams of highly enriched U-235 in each assembly, and function as fission chambers when driven into the core.
62 incore thimble assemblies have been received at SQN. Of these, only 51 have been positively located. The other 11 are believed to be located in a high radiation storage area.
The 11 incore thimbles involved are highly radioactive as a result of having been inserted into high neutron flux regions of the core.
They were replaced along with approximately 14 feet of drive and power cable when the U-235 was expended.
>
-
-
..
= Several workers have indicated that they remember bagging the spent thimbles and _ tossing them over a concrete wall into the high. level storage area. The licensee is reluctant to enter this area where 35 bags and one 55 gallon drum would have to be opened to physically locate the missing 11 thimbles.
The licensee - contends that an inordinate number of man-rem would be expended. in pursuing this inventory of special-nuclear material (SNM). The licensee is exploring the possibility of using remote controlled robots to enter the area and open the bags to locate the thimbles, since radiation levels in excess of 15R are involved. The alternative is to declare the 11 thimbles lost and submit documentation to that effect.
A. special inspection was performed by R II personnel a.ssigned to the Nuclear Materials Safety and Safeguards Branch, and will be detailed in inspection report 327,328/86-63.
Final resolution of the _ missing incore thimbles will be tracked as IFI 327,328/86-60-02.
At approximately 3:00 p.m.
on November 4, 1986,. Health Physics personnel unlocked the door to the High Rad Waste Storage Area to allow work-on electrical junction boxes.in the area. The' technician left the area after opening the door.
This action was in violation of the licensee's procedure Radiological Control Instruction (RCI) 13 - Access Control of High Radiation Areas When Intensity is Greater Than or Equal to 1000 mrem / hour.Section IV.B of RCI-13 states that direct-surveil-lance shall. be controlled by a Public Safety Officer or, for short periods of time, may be controlled by a Health Physics representative.
At approximately 6:50 p.m.
EST the last electrician left the area leaving the door unlocked and unguarded. At 7:05 p.m. EST another HP technician was sent to check out the area and found the open door.
Technical Specification 6.12.2 states, "...each high radiation area in which the intensity of radiation is greater than 1000 mrem /hr... locked.
doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shift Engineer on duty and/or the Health Physicist."
Contrary to the above, on November 4,1986, the High Rad Waste Storage Area was left unlocked and-unattended from 6:50 p.m. to 7:05 p.m.
The
j inspector will conduct interviews and review associated documentation during the next inspection period. This item remains unresolved and
,
j will be followed as URI 327,328/86-60-09.
.
No violations or deviations were identified.
I.
L 6.
Engineered Safety Features Walkdown (71710)
This module was not performed during this inspection period.
Reference paragraph 18 of this report.
I
,
.
-
.
7.
Monthly Surveillance Observations (61726)
The inspectors observed / reviewed TS required surveillance testing and verified that testing was performed in accordance with adequate procedures; that test instrumentation was calibrated; that LCOs were met; that test results met acceptance criteria requirements and were reviewed by personnel other than the individual directing the test; that deficiencies were identified, as appropriate, and that any deficiencies identified during the testing were properly reviewed and resolved by management personnel; and that system restoration was adequate.
For complete tests, the inspector verified that testing frequencies were met and tests were performed by qualified individuals.
a.
The inspectors observed testing in progress for Surveillance Instruction (SI)-260 SIS / BIT /RHR Injection Flow Balance, Pump Performance and Check Valve Test on October 16, 1985.
During the performance of SI-260.2 BIT Cold Leg Injection Flow Balance, Pump Performance and Check Valve Test, the Charging pump flowrate exceeded 555 gpm. This event is in violation of procedures and is considered reportable by the licensee.
The inspector will review the LER and corrective actions when submitted.
b.
SI-98.1 Channel Calibration for Engineered Safety Feature Instrumenta-tion c.
The inspector reviewed the following open -items regarding the surveillance instruction reviews being conducted by the licensee:
327/LER 85-044 Safety Injection Fump Flow above TS limit 327/LER 86-007 Failure to Test Trip Function of Manual SI Hand Switch 327/LER 86-008 Monthly Channel Check for S/G Narrow Range Remote Shutdown Instruments Not Performed 327/LER 86-011 Setpoint for Containment Sump Level was in Error per TS 327/LER 86-013 Failure to Perform Adequate Testing of ESF Systems 327/LER 86-014 Inoperability of Auxiliary Building Gas Treatment System 327/LER 86-015 Failure to Functional Test RCP UV and UF Devices and Breakers 327/LER 86-017 Failure to Test Radiation Monitors During Core Alterations
>
_ -.
-
__
_-_
-
.
327/LER 86-018 Failure to Verify Indicated Power Availability Due to Inadequate Procedure 327/LER 86-020 Failure to Perform Quarterly Functional Test per TS 327/LER 86-023 Required Channel Calibration of Flow
'
Indicator Not Performed 327/LER 86-027 SR cannot be performed for EDG 327/LER 86-028 Some Test Instruments Do Not Meet All ASME Section XI Requirements 327/LER 86-035 Failure to Follow A SR Due to An Inadequate Procedure 327/LER 86-44 Inadequate Verification of ECCS Flow 327/LER 86-046 Verification of Snubber Drag Force Trending 327/LER 86-048 Verification of ECCS Flow 327/NRE 81-006 Fire Barrier Penetrations Not Inspected per Surveillance Requirements 327/VIO 86-20-08 Failure To Verify Position of Containment Isolation Valves IAW TS 4611A 327,328/IFI 86-28-07 TS Surveillance Requirements - Followup TVA Review 328/ PRO 86-109 Reactor Coolant Flow DP in SI-455 and SI-246 (LER has not been written yet)
With the exception of LERs86-007 and 86-027, these items were identified by the licensee during the surveillance instruction review.
This program includes a method of corrective action and preventative action.
The NRC will inspect this program and its products in inspections prior to the restart of Unit 2.
The above listed items are l
administrative 1y closed and the review for potential enforcement action-will be tracked as URI 327,328/86-60-10.
d.
The inspector reviewed the concern over sufficient water in the Cold Leg Accumulators.
In a September 9,1986 letter to TVA, Westinghouse informed the licensee that the high and low alarm setpoints used at Sequoyah are consistent with the TS. The letter went on to provide documentation that the values in the Sequoyah Precautions, Limitations and Setpoints Manual and TS are consistent with the Sequoyah LOCA Analysis. The inspector had no further questions.
-_
-
.
- - - -
-
.-
-..
. _ _ -
- ___ -
..-
-
.
8.
Monthly Maintenance Observations (62703)
a.
Station maintenance activities of safety-related systems and components were observed / reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides, industry codes and standards, and in conformance with TS.
The following items were considered during this review: LCOs were met while components or systems were removed from service; redundant components were operable; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable; procedures used were adequate to control the activity; troubleshooting activities were controlled and the repair record accurately reflected what actually took place; functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; QC hold points were established where required and were observed; fire prevention controls were implemented; outside contractor force activities were controlled in accordance with the approved Quality Assurance (QA) program; and housekeeping was actively pursued.
b.
The inspectors reviewed an incident that was initiated by a modifi-cation to the plant, during which the licensee operated Unit 1 in mode 3 and above on three occasions with only one Post Accident Monitoring (PAM) channel for wide range Reactor Coolant System (RCS)
pressure:
1) April 12,1984 through April 20, 2) the day of May the 5th, and 3) May 9 through July 26, 1984. Sequoyah TS 3.3.3.7 requires that the accident monitoring instrumentation channels shcwn in Table 3.3-10 shall be operable in modes 1, 2 and 3.
Table 3.3-10 states that two channels of RCS pressure are required. The following documents and drawings were reviewed:
Engineering Change Notice (ECN) L6055 ECN L5035 Workplan (WP) 10810 WP 11216 WP 11173 WP 10902 WP 11174 WP 10766 WP 10828 WP 10848 Facility Change Request (FCR) 2634 FCR 2120 Potentially Reportable Occurrence (PRO) 1-84-276 Final Safety Analysis Report (FSAR) section 7.6.2.1 Drawing 47W610-68-7 Revisions 3 through 7 Drawing 47W610-68-4 Revisions 19 and 20
r
-
.
Plant Modifications In December of 1983, an Engineering Change Notice (ECN) number L6055 was written to relocate the Pressure Transmitters (PT-68-66 and PT-68-69) outside the containment for Equipment Qualification (EQ) and accuracy concerns.
This ECN was completed in the field during the cycle 3 refueling outage (April 1984).
Before the implementation of the ECN, the control room had three control board indications from these instruments: a recorder and two indicators. The recorder (PR-68-66) and one indicator (PI-68-69) were scaled 0-3000 psig and functioned as the two required independent PAM channels.
The remaining indicator (PI-68-66) provided the operators with indication (0-600 psig) for cold shutdown operations and was supplied by the same transmitter as the recorder (PR-68-66). Table 1 shows the changes that occurred to those indications during the modification process.
After the ECN was complete the control board's appearance was the same.
Only the gauge placards had changed. (Modification changes thrcughout this process are listed in Table 1.) After ECN L6055 was completed the recorder (PR-68-69) and the indicator (PI-68-69) with the 0-3000 psig scaling were fed from the same transmitter (PT-68-69).
The third indicator (PI-68-66) was to be the second PAM channel by design.
However, PI-68-66 was still scaled 0-600 psig which does not meet the requirements for PAM channels. To the operators, this configuration still appeared as before, but in actuality there were no longer two independent PAM channels since the 0-3000 psig recorder and indicator were now being supplied by the same transmitter.
Unit 1 entered mode 3, requiring two PAM channels, while in this configuration and continued to operate in violation of the TS until July 26, 1984.
On July 25, an engineer discovered the error while performing accuracy calculations involving the Emergency Instructions (EI) and utilizing the instrument drawings approximatley three and a half months after the Unit returned to power.
A Temporary Alteration Change Form (TACF)
1-84-83-68 was issued and PI-68-66 was rescaled to 0-3000 psig.
This returned the second PAM channel to an operable status. This evolution occurred over a one and a half day period.
Finally, a permanent fix was effected by Facility Change Request (FCR)
number 2634 under the original ECN. The TACF was lifted as a result of this tCN.
This returned the original indicator (PI-68-66) to a 0-600 psig scale for cold RHR operations indication, and placed the wide (0-3000 psig) range indicator on transmitter 1-PT-68-66.
The change assured two independent PAM channels were indicating in the Control Room.
l
.
.
TABLE 1 Before ECN L6055 Indication Recorder Indicator #1 Indicator #2 Identification PR-68-66 PI-68-69 PI-68-66 Range (psig)
0-3000 0-3000 0-600 PM4 usage PAM II PAM I None Transmitter PT-68-66 PT-68-69 PT-68-66 After ECN L6055 (drawing revision 7) - Before TACF 1-84-83-68 Indication Recorder Indicator #1 Indicator #2 Identification
- PR-68-69 PI-68-69
.PI-68-66 Range (psig)
0-3000 0-3000 0-600 t PAM usage
- PAM I
- None
- PAM II!
Transmitter
- PT-68-69 PT-68-69 PT-68-66 After TACF 1-84-83-68 - Before FCR 2634 **
lt Indication Recorder Indicator #1 Indicator #2
.
Identification PR-68-69 PI-68-69 PI-68-66 Range (psig)
0-3000 0-3000
- 0-3000 PAM usage PAM I None PAM II Transmitter PT-68-69 PT-68-69 PT-68-66 After FCR 2634 Indication Recorder Indicator #1 Indicator #2 Identification PR-68-69
- PI-68-66A PI-68-66 Range (psig)
0-3000 0-3000
- PAM II
- None Transmitter PT-68-69
- PT-68-66 PT-68-66
- changes made
- This would have been the modified configuration had drawing 47W610-68-7 Revision 4 been implemented.
During a review of ECN L6055 the inspectors determined that the version of ECN L6055 which included revision 4 of the "as-designed" drawing 47W610-68-7 would not have resulted in a violation of TS LCO 3.0.4 and 3.3.3.7 (see note ** in Table 1 above). The version of ECN L6055 which included revision 4 of the "as-designed" drawing 47W610-68-7, would have removed a percent module (PM) PM-68-66B from pressure loop P-68-66.
This PM converted a 0-3000 psig process signal from the transmitter to a 0-600 psig output that drove indicator PI-68-66. In b
.
.
"
-
.
this version of the ECN, indicator PI-68-66, which was to be used as a PAM instrument, would have had the correct range of 0-3000 psig, and the TS would have been met.
After issuance of revision 4 to "as designed" drawing 47W610-68-7, the Department of Nuclear Engineering received feedback from the Operations Department that the design change would no longer provide 0-600 psig indication for RHR operations.
ECN L6055 was updated to add a PM (PM-68-66C) back into the loop. This was documented on Revision 5 and all subsequent revisions of 47W610-68-7. This change to ECN L6055 WP 11173 data sheet I was accomplished without consideration for the overall effect of the change and by an individual who apparently did not understand the design requirements of the plant.
In addition, drawing 47W610-68-7 Revision 5 was not accomplished in accordance with Engineering Procedure EN DES-EP 3.10, Design Verifica-tion Methods and Performance of Design Verifications. EP 3.10 requires that design drawings be carefully reviewed by experienced design engineers and that attachment I to EP 3.10 be used as directicn for checking design inputs and for design review of design input drawings, detailed construction drawings and procurement drawings.
EP 3.10 attachment 1 states chat:
"The design review must include, but is not limit'ed to satisfactorily addressing the following questions that are applicable. Applicability is determined by the checker and/or controlling documentation. These directions indicate that the designer and checker should independently consider the availability of inputs in these areas. However, they do not mean that criteria or other formal input will necessarily be required or available."
Three of the criteria identified in EP 3.10, attachment 1 are:
" Operational requirements under various conditions, such as plant startup, normal plant operation, plant shutdown, plant emergency operation, special or infrequent operation, and system abnormal or emergency operation."
" Instrumentation and control requirements including indicating instruments, controls, and alarms required for operation, testing, and maintenance.
Other requirements such as the type of instrument, installed spares, range of measurements, and location of indication should also be included."
" Redundancy, diversity, and separation requirements of structures, systems, and components."
The design effort resulting in the version of ECN L6055 which included revision 5 to "as-designed" drawing 47W610-68-7 did not include consideration of at least the above three directions resulting in a
,_
._,
.-
_
_.
_
._
..
__-___-
.
.
.
..
-
.
change to the operational pressure indicating range of PI-68-66.
Additionally, the version of ECN L6055 including revision 5 to
"as-designed" drawing 47W610-68-7 was reviewed by a checker
!
(independent review) and a design supervisor. The inspectors inter-
viewed the checker who was responsible for independent review. The review conducted by the checker, requiring coordination of many drawings, was partially conducted by telephone. The checker also stated that he was " walked through" the change that resulted in revision 5 to the "as-designed" drawing 47W610-68-7, by the design engineer who was responsible for the original design. Consequently, the checker failed to conduct an independent review.
Had these reviews been done properly the licensee should have deter-mined that, as drafted, the design change would have-resulted in a lack of adequate PAM instrumentation. Interviews with the design personnel involved indicated that they could not account for why the error was not discovered.
The described design review deficiencies are con-sidered to be an example of the previously cited violation 327, 328/
86-19-06.
TS 3.0.4 states, " Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the conditions for the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION requirements. TS 3.3.3.7 states that the accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE. TS Table 3.3-10 item 5 requires two PAM channels of reactor coolant pressure be operable.
Contrary to the above, on April 12, 1984, the licensee took Unit 1 into mode 3 with only 1 PAM instrument for RCS pressure operable. Likewise, the licensee took Unit 1 into mode 3 with only 1 PAM instrument for RCS pressure operable on May 5, 1984 and again on May 9.
Furthermore, the plant continued to operate for greater than 7 days 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (From 0905 on April 12 to 0655 on April 20, 1984, and again from 0541 on May 9 to 1600 on July 26,1984) which is also a violation of action statement a.
of T.S. 3.3.3.7 which states, "with the number of OPERABLE accident monitoring instrumentation channels, except for the RCS subcooling
'
margin monitor, less than the Required Number of Channels shown in Table 3.3-10, restore the inoperable channel (s) to OPERABLE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />". Being brought on by design deficiencies, this violation of TS is considered to be another example the previously cited violation 327, 328/86-19-12.
Because of the age of the issues discussed above and the corrective action in progress by the licensee for violation 327,328/86-19-06 no additional Notice of Violation will be issued for the identi fied violations.
,
_ - - _ _ - _. _ _ _ _ -. -. -. -
c.
The inspector reviewed the following open items involving problems with ECCS pump performance and pump curves:
327 LER 83-005-Failure of Turbine Driven Aux. Feedwater Pump to Reach Rated Speed 327/328 LER 83-114 Failure of SIS 2A and Both CCPS to Meet Head Curve 327 LER 85-043 1A-A and IB-B CCPS Failed to Meet TS 327/328 URI 85-46-07 Determine Acceptability of AFW Pump Discharge Piping Vibration 328 LER 83-160 Turbine-Driven Aux. Feedwater Pump did not Obtain Min. Flow 328 LER 84-019 Centrifugal Charging Pumps Injection Flow Failed to Meet TS 328 VIO 84-21-02 ISI Procedure for RHR Testing Was in Technical Error 327/328 IFI 86-28-09 WR B119787 The inspector determined that although indicated corrective actions are different, there is a generic concern for the licensee's method of controlling pump operability, testing and pump curves.
The above listed items are administratively closed. The generic issue involving control of safety related pumps is unresolved and will be tracked as URI 327,328/86-60-11.
No deviations were-identified.
9.
Licensee Event Report (LER) Followup (92700)
The following LERs were reviewed and closed. The inspector verified that:
reporting requirements had been met; causes had been identified; corrective actions appeared appropriate; generic applicability had been considered; the LER forms were complete; the licensee had reviewed the event; no unreviewed safety questions were involved; and no violat;nns of regulations or TS con-ditions had been identified.
LERs Unit 1 327/84-048 RCS Pressure Channel Fails to Satisfy Requirements 327/85-001 Surveillance Instruction Not Performed Within Time Limits 327/85-004 Surveillance Instruction Not Performed Within Time Limits 327/86-016 Exceeding Maximum Fuel Rod Uranium Weight LERs Unit 2 328/84-009 Maintenance Performed On Main Feedwater Pump 328/85-006 Inoperable Containment Spray Pump 328/83-097 Containment Sump Level Channel Iraperable
__
-
.
_
.
l
-
.
In LER 83-097 the licensee committed to evaluate the problems and provide a supplemental LER. On November 4, 1986, this information was supplied to the NRC.
The inspectors made an indepth review of the problems with the Containment Sump Level System (CSLS) during May and June of this year. This review was documented in Inspection Reports 86-31 and 86-37. The informa-tion contained in this supplemental LER was reviewed at that time.
LER 83-097 is closed and the continuing operability of the CSLS will be reviewed under NUREG 0737 item II.F.2.e.
10.
Event Followup (93702, 62703)
a.
Loss of Spent Fuel Pit Level On December 18, 1985, with both units in cold shutdown mode, the spent fuel pit water level dropped approximately four feet below its normal
'
level. The level drop in the spent fuel pit resulted from a leak into the transfer canal. A leaking seal on the Spent Fuel Pit Gate was identified as the cause of the loss of spent fuel pit water level. The following documents were reviewed:
Independent Safety Engineering Group (ISEG) report dated December 18, 1985 Design Change Request (DCR) 2138 Drawings 47W846-2, 47W491-5 Final Safety Analysis Report, Section 9.1 Two contributing factors existed during this event. The first was an inadequate design of the air supply piping, connections and applicable procedural controls. The second was a lack of reflash on the spent fuel pit high/ low level indicator alarm. The annunicator had been in a continuously alarmed condition and was vilowed to ringout because it was considered a nuisance alarm.
Consequently, when an actual low spent fuel pool level occurred no visible or audible indication of the alarm was evident.
This event was determined to be not safety related and not reportable by the licensee. The air supply to the spent fuel pool pit gate was also not safety related and the applicable drawings stated that the licensee was to field route the air supply.
The licensee evaluated this event to determine the consequences of having the transfer canal open to either of the two available reactor
>
cavities or annuli when the leak occurred. If any of the four volumes had been available, the spent fuel pit would have drained to the level of the lower gate lip. The spent fuel would have remained covered with only five inches of water in this case. The licensee determined that any releases would be less than any FSAR section 15.5 analyzed fuel handling accident and therefore there were no safety consequences.
-.
.
.
This event is an example of a non-safety related system having significant input into a FSAR section 15 analyzed condition.
The inspector will track this and other events to determine if non-safety related functions, including maintenance, surveillance, and modifica-tions, are activities that are adversely affecting the safe operation of the plant. This item will be followed as IFI 327,328/86-60-04.
b.
In inspection report 86-49 the inspectors discussed a stop work order on TVA Class A, B, and C/D pressure retaining piping components. The licensee is currently reviewing the status and documentation of selected Class A piping.
This activity is being followed by the inspectors and will be identified as IFI 327,328/86-60-03.
c.
The inspector reviewed an event which caused the start of the Emergency Diesel Generators EDGs on November 2, 1986. An ASE was returning the 18-B 6.9 KV Shutdown Board to normal lineup following breaker cleaning.
The correct operation was to close the alternate feeder breaker on the ID 6.9 KV Unit Board. The ASE mistakenly went to the IC Unit Board and tried to close the Normal feeder breaker to the 1B-B Shutdown Board.
(The ID and 1C Unit Boards are located physically in the same cabinet.)
This action had no effect because the Normal feeder breaker to the 18-B Shutdown Board was already closed. The ASE assumed that the position indication had not been. reset.
Therefore, he went to open on the breaker.
This caused an undervoltage condition on the 1B-B 6.9 KV Shutdown Board and the EDGs to start.
The switching discussed above is on ncn-safety related equipment and therefore, was not covered under a procedure.
The NRC resident inspector is currently reviewing the question of non-safety related work that affects safety related systems.
The above issue will be further reviewed after the issuance of the LER.
No deviations or violations were identified.
11.
IE Information Notices (92701)
The following IE Information Notices (IEN) were reviewed and closed. The inspector verified that:
corrective actions appeared appropriate; generic applicability had been considered; the licensee had reviewed the event and that appropriate plant personnel were knowledgeable; no unreviewed safety questions were involved; and that violations of regulations or Technical Specification conditions did not appear to occur.
IEN 86-89 Uncontrolled Rod Withdrawal IEN 86-93 Evaluation Of Motor-Operators Identifies Improper Torque Switch Settings The inspector reviewed the licensee's respon:e to IEN 86-93 concerning faulty torque switch settings in Rotork valve operators. Sequoyah curr'ently has eight Rotork valve operators installed in ERCW lines to the EDGs. These operators are on butterfly valves which are position seated and not torque
_ _
-
.
seated.
Therefore, the torque switches have been electrically bypassed.
The problems discussed in IEN 86-93 are not applicable to the valve operators at Sequoyah.
12.
Inspector Followup Items (92701)
Inspector Followup Items (IFI) are matters of concern to the inspector which are documented and tracked in inspection reports to allow further review and evaluation by the inspector. The following IFIs have been reviewed and evaluated by the inspector. The inspector has either resolved the concern identified, determined that the licensee has performed adequately in the area, and/or determined that actions taken by the' licensee have resolved the Concern.
328/85-43-04 (Closed) Category B Procedure Steps 327,328/84-38-04 (Closed) Verify ERCW flow to EDG during inadvertent Safety Injection 327,328/85-46-01 (Closed) Corporate Commitment Tracking System 327,328/85-47-06 (Closed) Spent Fuel Pool Leak 327,328/86-15-01 (Closed) Revision of Surveillance Instruction 1 327,328/86-28-07 (Closed) TS surveillance review followup 327,328/86-28-12 (Closed) Rod Weight Increase 13.
Fuse Replacement Program (62703) (62700) (36100)
During this inspection period, the licensee began replacing KAZ blown fuse indicator devices with MIS-5 indicator / fuses.
KAZ devices are listed by Bussman Corp as an indicator device, not a fuse.
It was intended to be placed in parallel with a high amperage fuse as a blown fuse indicator.
When the parallel fuse blows, current through the low-rated KAZ would immediately blow the wire in the body of the KAZ. This wire is also the retaining element of a spring-loaded indicator that pops out to actuate a blown fuse alarm.
This indicator, when extended, is also a visual indication of a blown KAZ device. TVA has used the KAZ in numerous circuits as a fuse, even though it is not certified as a fuse by Bussman. The KAZ is rated to have an infinite blow time at 6 amps or less, and essentially a zero time above 16 amps. Bussman stated early in 1972 that the KA7 could be used as a 6 amp fuse only if restrictive overload and short circuit limits could be met.
In a May 1972 letter to TVA, Bussman Corp. stated that the KAZ Fuse Indicator could be used as a fuse only if the minimum overload was 100 amps and the maximum short circuit was 15,000 amps (200,000 is the usual specification for short circuit ratings). Overload ratings are the basis of fuse designations. The KAZ did not have overload certification by Bussman.
Bussman stated this position in April and May 1972, and again in November 1981.
Bussman did, however, specify that if the 100 amp overload and the 15,000 amp short circuit limits were observed the KAZ was acceptable in the 125 vdc circuits in which TVA employed them.
,
-
.
In an April 1985 letter to TVA, Bussman provided specifications on KAZ resistance with a determined range of.210 to.290 ohms. Part of an SCR written by TVA (SQNEEB 8644) states that "all control power voltage drop analysis have previously assumed fuse resistance to be negligible; for the KAZ this is an invalid assumption. Thus, the voltage drop analysis on all circuits containing KAZs are rendered indeterminate; some safety related circuits may have excessive voltage drop, rendering them inoperable."
In a June 1986 letter to TVA, Bussman reiterated that the KAZ was not a fuse, but qualified that statement with the provision that if the previously stated limits were observed, the KAZ could function as a fuse.
In this letter, Bussman also stated that currents above 6 amps (the KAZ's nominal rating) and less than 15 amps, could cause the KAZ to catch fire and burn.
This potential hazard was pointed out in the licensee's SCR.
With all the attendant problems of the KAZ being used a a fuse, TVA decided to change out the KAZ's in applications where they were used as fuses with a Bussman fuse model MIS-5. The MIS-5 had the same distinctive shape and size as the KAZ, and would not require modifying the fuse holders and clips.
The MIS-5 fuse is no longer made by Bussman, and all available copies are held by Nutherm Inc., who also performs seismic shaker tests on the MIS-5 fuses they supply to TVA. Tne seismic testing apparently is causing a high rate of damage to the current-conducting silver wire, while leaving the parallel spring-retaining alloy wire intact.
This second wire holds a spring which is attached to a small metal tab which extends out from the fuse ferrule to indicate that the fuse has blown. While not capable of carrying more than roughly 0.5 amps, the alloy wire has better tensile strength than the silver wire.
When Sequoyah personnel began installing the MIS-5 fuses, several blew immediately when placed in energized circuits. Further investigation by TVA revealed that the fuses were being damaged during the seismic testing, but the damage was not readily identified by visual observation.
After discussing these problems with Bussman and Nutherm, TVA decided to stop the KAZ to MIS-5 replacement program, now approximately 80% complete.
Further evaluation by TVA concluded that MIS-5 fuses installed in unenergized or low current circuits could be defective with no observable indicators.
Further, these potentially defective fuses could render the equipment they served inoperable. On October 27, TVA made a determination to declare such equipment inoperable and to enter any applicable LCO or action statement until the MIS-5 fuses could be removed and verified intact.
Only one LCO was entered as a result of the action taken (RHR loop operability for RCS decay heat removal, 3.4.1.4).
The fuse in the control circuit for valve FCV-62-1, the RHR pump suction from the RWST, was assumed to be defective. As a result the valve was declared inoperable and the applicable LCO was entered. This LCO was exited later after the suspect fuses were pulled and replaced with certified fuse.-
.
18-A total of 55 suspect fuses were -identified as having been installed in circuits to equipment required for Mode 5, all oof which have now been replaced with certified fuses. All other unverified fuses are baing pulled and replaced _ on a lower priority. Of the 55 fuses pulled and checked by a resistance test, 15 failed. Further, 8 of the fuses which were to be used as replacements were also found defective during their certification. The fuse replacement program will be followed as IFI 50-327/328-86-60-05.
A report has been submitted by the licensee under the requirements of 10CFR21 and will be reviewed as a part of this IFI.
14. Microbiotic Induced Corrosion (62703) (62700)
As a result of discovering unidentified corrosion in both low carbon and stainless steel components at Watts Bar nuclear plant an inspection was conducted by TVA metallurgists at the Sequoyah Nuclear Plant.
The inspection team identified two stainless steel welds in the ERCW system that were visibly weeping and requested that approximately 27 additional welds be Ultrasonically Tested (UT). During the UT weld preparation process, which involved grinding the subject welds flush with the piping, approximately ten additional welds were identified as having through wall leaks.
A discussion was held with the two metallurgists and a licensing representative concerning Sequoyah's resolution of IEN 85-30.
TVA memo McCloud/ Miller stated that "MIC can be controlled -by maintaining flow through pipes and by chlorination.
Since our safety systems continuously run, even during outage and are chlorinated during the summer months, it is not likely that MIC will become a problem." Sequoyah's approach duplicated the above statement.
In addition, the inspector reviewed the following information:
Electric Power Research Institute (EPRI) Proceeding, May 1986 Watts Bar NCR W-471-P, Revision 0, dated August 8, 1986 IEN 85-30: Microbiologically Induced Corrosion of Containment Service Water System TVA Memo McCloud/ Miller response to IEN 85-30 PRO 2-86-104, Leak in Stainless Steel ERCW piping SER 73-84, October 1984 IEN 85-30 identified MIC at H. B. Robinson Unit 2 involving 6 inch stainless steel piping that provided raw service water. The corrosion at Sequoyah is similar to that discovered at H. B. Robinson. Both plants had corrosion in their safety related raw cooling water systems (raw service water vice essential raw cooling water). The corrosion occurred in the same size (both six inch schedule 10) and type (304 stainless) of pipe.
MIC was addressed in an INPO SER in 1984, and in EPRI documentation in 1985 and 1986. In addition, MIC was identified in stainless steel by at least two additional utilities (South Texas, Palo Verde) and in Copper Nickel (Wolf Creek).
--
-
._
.
.
.
- -.-
..
--
r
.
TVA is presently evaluating preventive and corrective measures to combat the identified MIC in the ERCW systems.
This issue will be tracked as IFI 327,328/86-60-06.
15. Generic Fittings (62703) (62700) (37700)
On April 19, 1984, incore instrument thimble D-12 of Sequoyah Unit I was forced out into the incore instrument room in containment by reactor coolant system (RCS) pressure.
Thimble D-12 was undergoing a drybrush cleaning process, which involved inserting a cleaning tool about 80 feet into the thimble. The cleaning process caused a failure of a high pressure boundary resulting in a 25 to 35 gpm reactor coolant leak.
It was determined that one of the major contributing factors which resulted in the RCS leak was the use of generic components in the seal table high pressure boundary fittings.
Fittings were crossed between manufacturers even though one major manufacturer prohibited the use of generic components as part of its pressure seal.
Th.s issue was discussed in Inspection Reports 327,328/
84-11, 84-12, 84-24, 85-27 and 86-15.
TVA replaced all high pressure boundary fittings in the Unit I seal table with deep bore fittings in the fall of 1985 and verified that no mixed fittings existed in the Unit 2 seal table. Appropriate maintenance and surveillance instruction changes have been accomplish.ed to ensure that generic components will not be used in conjunction with fittings on the seal.
The inspectors reviewed an interim TVA Employee Concerns Special Program report numbered C017304-SQN revision 1.
This report came to similar conclusions. However, the report stated that some incorrect installations have been found on new modifications. The modifications referred to in the report are not specifically identified and may be on instrument and drain lines. The inspector will review the completed version of this report to ensure that additional information or issues have not been identified IFI 327,328/86-60-07.
The interim version of this report was not retained by the inspector.
16. Qualification and Certification of Sequoyah Quality Control (QC)
Inspectors (36700)
As a result of several questions brought to the inspectors' attention through varied sources, a recent procurement team inspection examined the qualification of QC inspectors during the performance of the normal procurement inspection module.
Members of the procurement inspection team discussed preliminary findings with the resident inspectors and recommended a detailed examination in the areas of QC Inspector Qualification, Certification, On-the-Job-Training (0JT), and medical record r.
-
.20
-.
It was ' determined 'that-the licensee had identified similar issues through the current Employee' Concern Program and through the Watts Bar Special Employee' Concern Program. The inspector reviewed the licensee's actions, which 'were in progress, and examined the following completed and. interim
-
documents..Some of the documents were determined by the licensee to be
" administrative 1y confidential".
No documents were retained by the inspector.
SQ CAR 86-01-003 This Corrective Action Report (CAR) was dated January 31,-1986 and addressed many of the issues stated above.
Checklist 2-86-S-001 This checklist implemented investigation and'
reporting functions for. portions of CAR 86-01-003.
Checkli stl 21-86-S-026 This checklist implemented investigation and reporting functions for portions of CAR 86-C1-003 Nuclear Safety Review Staff (NSRS) Investigation Report I-85-373-NPS Interim-Element Report - Quality Control Inspector Qualifications Based on a review of the above documentation, it appeared that the licensee is ' accomplishing the appropriate corrective action to establish Quality Control Inspector qualification and certification requirements. This issue will be addressed by the NRC Technical Area Manager in a Safety Evaluation as part of the NRC evaluation of the Sequoyah related Watts Bar Special Employee Concern Program Element Reports.
This item will. be tracked as Inspector Followup Item 327,328/86-60-08.
,