ML20148S179
| ML20148S179 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 12/29/1987 |
| From: | Jenison K, Mccoy F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20148S125 | List: |
| References | |
| 50-327-87-71, 50-328-87-71, GL-85-28, NUDOCS 8802020355 | |
| Download: ML20148S179 (28) | |
See also: IR 05000327/1987071
Text
URITED STATES
[gR Riog*o NUCLEAR REGULATORY COMMISSION l
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< 101 MARIETTA STR EET, N.W.
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E ATLANTA, GEORGI A 30323
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Report Nos.. 50-327/87-71, 50-328/87-71
Licensee: Tennessee Valley Authority
500A Chestnut Street
Chattanooga, TN 37401
Occket Nos. 50-327 and 50-328 License Nos. OPR-77 and DPR-79
Facility Name: Sequoyah Units 1 and 2
Inspection Conducted: November 6, 1987 thru December 4, 1987
Lead Inspector:_{ AfuAs let L2/79/87
f.~M.JenisbgSdniorResidentInspector Date Signed
Accompanying Inspectors: P. E. Harmon, Resident Inspector
D. P. Loveless, Resident Inspector
W. K. Poertner, Resident Inspector
W. C. Bearden, Resident Inspector
M. W. Branch, Sequoyah Restart Coordinator
Approved by:2 b
McCoy, Cnief, Prod %t'tslection 1
_/M2f/gv
F. ' Date S'ign(d
Division of TVA 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 protectien, safeguards and
housekeeping inspections); maintenance observations; review of previous j
inspection findings; followup of events; review of licensee identified items,
and review of inspector followup items. l
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Results: Four violations were identified. 1
paragraph 8 - (327,328/37-71-01), Failure to implement adequate
design control.
paragraph 8 - (327,328/87-71-02), Failure to affect adequate
corrective action.
paragraph 13 - (327,328/87-71-03), Inadequate surveillance
instruction.
paragraph 13- (327,328/8)-71-04), Failure to report a Technical I
Specificatie, vic'ation.
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REPORT DETAILS
1. Licensee Employees Conticted
H. L. Abercrombie, Site Director
- J. T. La Point, Deputy Site Director
- S. J. Smith, Sequoyah Plant Manager
- B. M. Willis, Operations and Engineering Superintendent
B. M. Patterson, Maintenance Superintendent
R. J. Prince, Radiological Control Superintendent
- M. R. liarding, Licensing Group Manage'
L. E. Martin, Site Quality Manager
D. W. Wilson, Project Engineer
R. W. Olson, Mcdifications 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
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. H. Buchholz, Sequoyah Site Representative
- M. A. Cooper. Assistant to Licensing Manager
Other licensee employees contacted included technicians, operators, shif t
engineers, security force members, engineers and maintenance personnel.
- Attended exit interview
2. Exit Interview
The inspection scope and findings were summarized with the plant manager
e and members of his staff on December 4, 1986. The four violations :
described in this report's summary paragraph were discussed. No I
deviations were discussed. The licensee acknowledged the inspection l
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findings, and did not identify as proprietary any of the material reviewed
by the inspectors during this insoection. During the reporting period,
freauent discussions were held with the Site Director, Plant Manager and
other managers concerning inspection findings.
3. Licensee Action on Previous Inspection Findings (92702)
(Closed) Violation 327,328/87-36-01; Inoperable Chlorine Detection System. l
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This violation was written to address the inadequate functional testing of
the chlorine detection system which was not adequately tested to ensure
that the system would 4solate the control room ventilation in the event of
a chlorine spill. The licensee had previously tested this function by
observation of an a'. arm actuation rather than the actuation of the
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contacts which initiated control room emergency ventilation (CREV). Since
the alarm contacts were in parallel with the CREV contacts, the licensee
assumed that operation of the alarm proved the corresponding actuation of
the CREV. Surveillance instruction (SI)-168, Calibration of Control Room
Air Intake Chlorine Detection System, and SI-240, Functional Testing cf
Control Room Air Intake Chlorine Detection System, have been revisea to
require that the CREV actuation contacts are functionally tested as part
of the surveillances. In addition to the SI revisions, the licensee
initiated a Technical Specification (TS) change (proposed TS change 78) to
delete the chlorine detection system from the plant and from TS. This TS
change was approved and issued as amendments 62 (Unit 1) and 54 (Unit 2)
on October 30, 1987. This item is closed.
(Closed) Violation 327/86-62-02 and 328/86-62-09; Failure To Properly
Control The Design Process During Hydrogen Analyzer Installation. A
re/iew of SQEP-13, Procedure For Transitional Design Change Control,
SQEP-39, Review And Approval Of Vendor Manuals / Revisions, and NEP-3.2,
Design Input, indicate that the vendor manuals / revisions will be properly
controlled and that vendor manuals will be reviewed for information
critical to the implementation and use of equipment. Corrective action
that was taken to resolve the specific problems associated with the
hydrogen analyzer installation, identified as item 328/86-62-09 was
reviewed during a subsequent inspection (IR 327,328/87-42) and left open,
as the inspector tied the specific issue to the overall vendor manual
problem as defined in generic letter 83-28. The generic aspects of vendor
manual control and update is a separate issue and is being followed as
part of the generic letter 83-28 followup. The use of vendor manuals as
input to the design process was reviewed during the -design control
fcilowup inspection (IR 327,328/87-42), and that inspection determined,
that for the modifications reviewed, the licensee was properly using
vendor manuals in the design change process. This item is closed.
(Closed) Unresolved Item 327,328/86-68-05, section 2.4.10, U-2.4-2;
Flamastic Thickness On Cable Trays. This item was reviewed in inspection
report 87-65 and was lef t open pending review of the final- TVA response.
TVA final response memo B25 87 1029 008, Commitment Verification of
Tracking Numbers NCO 870029002 and 9C0860484005, has been reviewed and
appears adequate. This item is closed.
(Closed) Unresolved Item 327,328/87-30-05; Qualification And Training Of
Testing Persornel. The inspector reviewed the following licensee actions
pertaining to this item:
Administrative instruction AI-47, Conduct Of Testing, has been issued
for implementation at Sequoyah. Section 5.0 of this instruction
references qualification and training requirements for personnel
directing and performing tests.
On September 1, 1987, a directive from the plant manager
(S53-870831-871) mandated that, as of September 1, 1987, only
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personnel trained to AI-47 will be allowed to direct testing
activities. The power operations training center (P0TC) course
number for this required training is MTS201.005.
The above referenced directive states that five hundred eighty-six
individuals had successfully completed the required-training at that
time. The inspector reviewed a sarapie of the applicable training
records and determined that they are adequate.
This item is closed.
(Closed) Violation 327,328/86-43-02; Failure To Properly Change Plant
Approved Procedures. This i tem involved the _ licensee's failure to
properly revise plant approved procedures when it was determined that a
change was needed to complete work in progress. The problems associated
with the licensee's temporary change process were discussed in detail in
inspection report 327,328/86-43. The inspector determined that the
licensee was abusing the temporary change process allowed by Technical
Specifications to correct a basic problem of how to expedite a change, l
either intert or non-intent, through the TVA review process which
included, in some cases, a plant operations review committee (PORC)
review. The licensee's original response to the violation dated
October 31, 1986, and supplemental response dated February 11, 1987, were
reviewed. These responses describe the interim changes that the licensee
took and those changes planned to revise administrative instruction (AI)
AI-4, "
Preparation, Review, Approval And Use Of Site Procedures /
Instructions". Additionally, the licensee indicated that a change to the
TS was planned to allow use of the qualified reviewer process, in-lieu of
requiring a PORC review of procedures and procedure changes. The
inspector questioned the TVA quality assurance (QA) organization as to
their plans to perform an implementation audit of the changes to AI-4 as a
result of the violation and TS changes. The QA implementat1on review
identified additional AI-4 changes that were needed. Revision 64 of AI-4
was reviewed to ascertain if the commitments contained in the licensee's
,. responses were implemented. The inspector's review determined that
currently the licensee's program meets the requi rements of TS and
therefore this item is closed. However, since the original issue was a
result of program implementation, the inspector will continue to mor itor
the use of temporary changes as part of the routine inspection program.
(Closed) Violation 327,328/87-42-02; Failure To Implement Upper Tier
Procedure Requirements In The Specified Ninety Day Time Period. TVA
corrective action for this item was a review of program manual procedures
which affect division of nuclear engineering activities to assure that all
requirements had been implemented in nuclear engineering procedures
(NEPs). Engineering assurance has performed the above review and one
additional revision to the NEPs was required. This revision has been
issued. The licensee's corrective actions appear adequate. This item is
closed.
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(Closed) Unresolved Item 327,328/86-44-01; NRC Inspection Report
Experience Review. This item was determined to be a subset of the issues
involved in the still outstanding NRC order EA 85-49. NRC has scheduled a
special inspection, which is intended to evaluate the licensee's
corrective actions t.nd close the order. Closure of the order will be
necessary prior to the startup (entry into Mode 2) of either Sequoyah
unit. The order identified a breakdown in- the management controls for
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evaluating and reporting of potentially significant safety conditions.
Inherent in this issue was the failure of the licensee to relay to the
appropriate levels of management, adequate information with regard to
specific issues for safety significance.
One aspect of this issue was inspected and addressed in inspection report
327,328/87-30. The area inspected was the licensee's nuclear experience
review program. No discrepancies were identified with respect to this
program. Several possible weaknesses were identified for long term NRC
review.
The licensee currently has a corporate commitment tracking system (CCTS)
which is responsible for the review of NRC inspection reports. The CCTS
was found to be acceptable in NRC inspection report 327,328/86-37.
The issues reviewed in reports 327,328/86-37 and 327,328/87-30 are subsets
of the order. As a result, Unresolved Item 327,328/86-44-01 is. closed and
the remaining issues will be followed under the resolution of the order.
(Closed) Violation 327,328/86-56-01; Failure To Report SQNCEB8409. In
inspection report 327,328/87-54, item A of this violation was closed. In
inspection report 327,328/87-60, items C1 and C2 were closed. The
licensee's response to item B dated March 23, 1987, stated that a
controlled walkdown was performed to identify and document hazards.
ECNL6770 was written to accomplish corrective actions for deficiencies
found. This item is being tracked as an nuclear performance plan (NPP)
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commitment. This item is closed.
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. No unresolved items were identified during this inspection.
5. Operational Safety Verification (71707)
a. Plant Tours
The inspectors observed control room operations, reviewed applicable
logs, conducted discussions with control room operators, observed
shif t turnovers, and confirmed operability of instrumentation. The
inspectors verified the operability of selected emergency systems, i
and verified compliance with Technical Specification (TS) Limiting l
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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.
Tours of the diesel generator, auxilia ry, 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.
The inspectors walked down a portion of the Unit 2 residual heat
removal system to verify operability and proper valve alignment.
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 l
controls; searching of personnel and packages; escorting of visitors;
patrols and compensatory posts; and badge issuance and retrieval. ;
In addition, the inspectors observed protected area lighting,
protected and vital areas barrier integrity. The inspectors verified
an interface between the security organization and operations or
maintenance. Specifically, the resident inspectors: interviewed
individuals with security concerns; inspected security during
outages; visited central or secondary alarm station; verified
protection of safeguards information; and verified onsite/offsite
communication capabilities.
No violations or deviations were identified,
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c. Radiation Protection
The inspectors observed health physics (HP) practices and verified
implementation of radiati a protection control. On a regular basis,
radiation work permits (KWPs) were reviewed and specific work
activities were monitored to ensure the activities were being
conducted in accordance with applicable RWPs. Selected radiation
protection instruments were verified operable and calibration
frequencies were reviewed.
No violations or deviations were identified.
6. Monthly Surveillance Observations (61726)
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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. The following
surveillance instructions (SI) were reviewed:
a. SI-7, Electrical Power Systems: Diesel Generators. The inspector
observed / reviewed portions of SI-7 on the 1A-A diesel generator.
During the performance of this surveillance, the operator blocked the
automatic initiation of the carbon dioxide fire deluge system for the
room in which the test was being conducted. The carbon dioxide
deluge system is described in TS 3.7.11.3. TS 3.7.11.3 states that
with one or more of the required carbon dioxide systems inonerable,
establish a continuous fire watch with backup fire suppression
equipment for those areas in which redundant systems or components
could be damaged within one hour.
Physical security instruction (PHYSI) -13, Fire, addresses the use of
fire watches and compensatory measures. The inspector reviewed +.he
applicable attachment F of PHYSI-13 for the 1A-A diesel generator
room. The attachment stated that a continuous fire watch was to be
posted in the diesel generator room. When the inspector entered the
diesel generator room, SI-7 was in progress and there was no fire ;
watch present in the room. The fire watch wa,s sitting in an outer
passage way outside of the diesel generator room. The fire watch
failed to implement PHYSI-13 adequately. There was a lack of safety
significance because the TS does not require a continuous fire watch
in this case, and because redundant diesel generators would not have
been affected by a fire in the 1A-A diesel generator room. This
issue was discussed with the plant manager on November 20, 1987.
Because of the lack of safety significance, the consideration that
this incident was an anomaly, and the prompt corrective action by the
test director and the shift engineer, a violation will not be issued.
b. On November 19, 1987, the licensee performed SI-240, Functional Test
Of Control Room Air Intake Chlorine Detection System, on train A of
Unit 1. At 11:00 a.m. the technicians completed testing train A and
returned it to service. At 11:22 a.m., the technicians discovered
that the train B monitor had been in the test position during the
testing of train A. This situation would have required entry into
LC0 3.3.3.6.b. The LCO requires that with both chlorine detection
systems inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> initiate and maintain operation of
the control room emergency ventilation system in the recirculation
mode of operation. The licensee initiated PRO 1-87-427 to evaluate
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the reportability of the event. The inspector will track this item
under the PRO and LER to determine any regulatory issues.
c. The inspector observed portions of the performance of SI-90.82,
Reactor Trip Instrumentation Monthly Functional Test. The SI
functionally tests and calibrates the A train solid state protection
system (SSPS). The testing addresses that portion of the system from
the input relays through the undervoltage coils, master relays, and
multiplexers. The testing is intended to meet the surveillan
requirements of TS 4.3.1.1.1, 4.3.1.1.2, 4.3.2.1.1, ando4.3.2.1.2 .or
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the reactor trip system and engineered safety feature actuation
system (ESFAS) instrumentation . channels and interlocks. No
discrepancies were noted.
d. The inspector observed the performance of SI-130.2, Motor Driven
Auxiliary Pumps, conducted on the 2A-A motor driven auxiliary feed
pump. The surveillance is performed to demonstrate pump operability
by verifying pump inlet pressure, differential pressure, discharge
pressure, flow rate, vibration, and lube oil level within acceptable
range. No discrepancies were noted.
e. The inspector observed a portion of SI-291.1, Readjustment Of
Setpoint For Frequently Used Radiation Monitors With Variable
Setpoints, performed on radiation monitor 0-RM-90-225. No
discrepancies were noted.
No violations or dcviations were identified.
7. Monthly Maintenance Observations (62703)
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.
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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 cstablished 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.
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a. The inspector observed / reviewed portions of preventive maintenance
(PM) 0412-090, radiation monitor recorder. No discrepancies were
noted.
b. The inspector observed portions of ongoing _ maintenance associated
with piping replacement of several sections of the ERCW-system in the
auxiliary building due to microbiologically induced corrosion (MIC).
The following work requests were reviewed:
WR B240546, replacement of a portion of section 3, class 2, ASME 312,
TP316, 6-inch, stainless steel piping which included an elbow and
weld numbers 12191 and 12192. New welds 12191A, 12191B, 12192A, and
121928 were made using detailed weld procedure GT88-0-3, Rev. 2. The
inspector visually inspected welds 121918 and 12192A, and observed
ongoing work. Additionally, portions of the hydrostatic testing was
observed.
WR 8251598, replacement of a straight piping portion of section 3, ,
class 2, ASME 312, TP316, 6-inch, stainless steel piping which
included weld number 14422. New welds 144218 and 14421C were made '
using detailed weld procedure GT88-0-3, Rev. 2. The inspector
visually inspected both welds and observed ongoing work. l
WR B240527, replacement of a portion of section 3, class 2, ASME 312,
TP316, 6-inch, stainless steel piping which included 2 elbows and
weld numbers 14351 and 14348. New welds 14351C, 143510, 14350A,
14349A, 14348A, and 143488 were made using detailed weld procedure
GT88-0-3, Rev. 2. The inspector visually inspected welds 12191B and
12192A, and observed ongoing work.
In all cases, welding was acceptable and 51-265.0, Hydrostatic
Testing, was performed to verify system integrity.
c The inspectors followed the progress of the repairs performed on the
Unit 2 residual heat removal (RHR) hot leg suction isolation valve
2-FCV-74-2; a 14-inch motor operated gate valve. The valve had been
experiencing excessive packing leakage with boric acid buildup on the
valve bonnet, possibly causing corrosion of the bonnet-bolting. The
licensee planned to replace the valve stem which was thought to be
bent. Both loops of the RHR system were taken out 'of service at
, 6:23 a.m., on November 3, 1987, entering the LC0 for Technical
Specification 3.4.1.4. Operations personnel commenced SI-127, Heatup
Rate, in accordance with S01-74.2. The valve stem and packing were
replaced, backseat lapped, and other internals cleaned and inspected.
The valve was reassembled later that evening and the system filled
and returned to service at 7:50 a.m. on November 4,1987. No change
in RCS temperature (stayed at 112 degrees F) or source range counts
(stayed at 2 cps) occurred during the tbae that the system was out of
service.
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In conjunction with the work on 2-FCV-74-2, the inspector reviewed
temporary alteration control form (TACF) 0-87-031-31A which was
implemented to seal off normal air flow to the cable spreading room.
The spreading room air supply fan was placed under a hold order in
support _of this TACF. This temporary alteration was installed to
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make the control room emergency ventilation system operable because
of ongoing work which caused a break in the system flowpath.
SI-144.2, Control Room Emergency Ventilation Test, was performed with
the temporary alteration in place to demonstrate operability of the
control room emergency ventilation system. These actions were
performed by the licensee to comply with Technical Specification .
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3.7.7.b which would not otherwise allow any positive reactivity
changes.
During the performance of the work on valve 2-FCV-74-2, the stem was
replaced as planned, the backseat lapped to remove rough spots and
the valve returned to service. New nuts were installed and the old
valve body studs inspected and reused with the exception of one stud
that required replacement due to galling which occurred during the
work. One problem that was noted by the inspector was that a blank
flange / cover made by the licensee did not initially fit. The flange
had been made by the licensee and was intended to be used as an
emergency measure to return the RHR system to service (with the valve
bonnet removed) in the event that core decay heat removal was
required. The holes that were drilled ir, the flange for the valve
body studs were not sized large enough to allow proper fit. Work was
temporarily stopped until the cover could be reworked to fit properly
(i.e. the valve bonnet was not taken apart and could have been placed
back on the valve body if the need occurred).
The inspector reviewed work request WR B288251 and special
maintenance instruction, SMI-2-74-1, and found both provided
sufficient controls and contained adequate technical instructions for
the intended work. Post maintenance testing specified included
e performance of SI-166.6, Valve Stroke Timing, MI-10.43, Movats
Testing; and SI-166.18, Leak Rate Testing. Additionally, SI-137.1,
RCS unidentified leakage testing was performed with satisfactory
results (.557 gpm) prior to returning the system to service. The
inspector had no further questions.
No violations or deviations were identified.
8. Licensee Event Report (LER) Followup (92700)
The following LERs were reviewed and closed. The inspector verified that- t
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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 violations of
regulations or Technical Specification conditions had been identified.
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LERs Unit 1
(Closed) LER 327/87-063, Control Room Isolation. This LER addressed the
inadvertent control room isolation which was caused by the misplacement of
an alligator clip. The alligator clip was being used to make electrical
connection during the performance of surveillance instruction (SI)-82,
Functional Tests For The Radiation Monitoring System. The licensee
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determined the root cause to be a combination of personnel error and lack
of procedure clarity. The inspector reviewed the licensee's corrective
action and determined that it was adequate. This item is closed.
(Closed) LER 327/87-004, Inadequate Determination Of Filter Train Charcoal
Adsorber Bank Efficiency Due To A Personnel Error. The licensee has
revised technical instruction (TI)-9, Test Methods For Nuclear Air
Cleaning Systems, to specify the use of "least squares fit" to the data to
extrapolate to time zero. Personnel have been trained in the use of least
squares fit and both the emergency gas treatment system (EGTS) and
auxiliary building gas treatment system (ABGTS) trains A and B have been
tested for Unit 2. The B trains of EGTS and ABGTS have been
satisfactorily completed. The A trains of EGTS and ABGTS have completed
testing and the data is being reviewed. The licensee's corrective actions
appear to be adequate. This LER is closed.
(Closed) LER 327/87-032, Chlorine Detection System. Problems. Addressed
problems resulted in an engineered safety feature actuation and a
violation of TS.due to a component failure and personnel error. During
performance of surveillance instruction (SI)-168, Calibration Of The "A"
Train Control Room Chlorine Detector, a control room isolation (CRI) ,
occurred due to a blown fuse (caused by test personnel actions). While l
the train "A" detector was out of service the train "B" detector was
declared inoperable dJe to an insufficient drip rate and a CRI was
initiated as required by the TS. However, the operator failed to shut off
the emergency pressurization fans as required by TS. Operations personnel
were counseled regarding the fan misalignment. TVA has submitted, and the
NRC has approved, a TS change deleting the chlorine detectors at Sequoyah.
The licensee's actions appear to be adequate. This LER is closed.
(Closed) LER 327,328/87-002, Lack Of Determining Tritium's Contribution To
The Monthly Dose Calculation Due To Procedural Inadequacy. A TS change
which occurred in January 1986, was not incorporated in the Sequoyah
procedures SI-423.1, Monthly Appendix I Dose Calculation Gaseous l
Ef fluents, and the Off site Dose Calculations Model (ODCM). Due to this i
failure to change the procedures, the tritium contribution to the monthly
dose calculation was omitted.
SI-423.1 was revised and approved on March 19, 1987, to include tritium in
the monthly dose calculations. The ODCM was revised and approved on
January 16, 1987, at a meeting of the radiological assessment review
committee (RARC). The licensee evaluated the quarterly dose results and
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11
determined that the limits of TS 3.11.2.3 were not exceeded. The
licencee's corrective action appears to be adequate. This item is closed.
(Closed) LER 327/87-013, Reactor Coolant System (RCS) Spillage While In
Mode 5 Due To A Procedural Inadequacy. In February 1987, approximately
3000 gallons of RCS water spilled from a steam generator manway during
conduct of SI-166, Stroke Testing Category A and B Valves During Cold
Shutdown. No personnel injuries or contamination resulted from this
event. The licensee established a special task force to investigate this
event. The root cause of this event was determined to be procedural
inadequacy and operator error. The following corrective actions have been
taken by the licensee to preclude similar future events:
AI-30, Conduct Of Operations, was revised to require operators to
review system flow diagrams of safety related equipment.
AI-43, Independent Qualified Review, was written to ensure an
adequate independent review process for procedures which affect plant
nuclear safety or changes to these procedures and to ensure
independent review of proposed changes or modifications to plant-
nuclear safety related structures, systems, or components.
MI-3.2, Method Of Plugging Steam Generator Tubes, was revised to
detail controls on potential RCS level change evolutions during steam
generator work.
Special training by site management for all operating crews was
conducted between May and August 1987, to ensure all personnel were
aware of the lessons learned from this and similar events.
Procedural compliance during all operations was heavily emphasized
during normal license requalification training. Special attention
was given to the sensitivity of level (volume) changes to the RCS
during conditions when the RCS is open.
r
A letter, dated August 24, 1987, was sent to all operating personnel i
that discussed the sensitivity of the RCS to level changes when the
RCS is open and the requirement for all operators to review flow
diagrams when performing safety related non routine valve alignments.
This letter was issued to supplement the above training.
Operations performed a walkdown of the SI-166 series for
procedural adequacy.
Based on the inspectors review of the above actions and discussions with
licensee personnel, the corrective actions for this LER appear to be
adequate. This item is closed.
(Closed) LER 327/87-015, Shutdown Margin Calculation Discrepancy. This
LER is an informational report on a discrepancy identified with the point
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xenon model used in shutdown margin (SDM) calculations. Westinghouse
notified TVA in February 1987, of a discrepancy in the SDM calculation
following transient oparating histories. Technical instruction (TI)-22,
SDM Calculation, has been revised to incorporate the corrective action
recommended by Westinghouse. Additionally, TVA has modified the SDM
computer calculation program to reflect the Westinghouse recommendation.
The licensee's corrective actions appear to be adequate. This item is
closed.
(Closed) LER 327/87-016, Diesel Generator (DG) Started On Loss Of Voltage
On 6.9 KV Shutdown Board Due To Personnel Error. During performance of
function testing procedure SMI-1-SB18-22, a technician incorrectly
connected a jumper to the wrong PK block. The technician has been
counseled by his supervisor on the importance of ensuring proper
identification of test devices as described in the instructions. The
above procedure was adequate as written, however the licensee is changing
the sequence of the steps to allow for double verification of the jumper
connection prior to tripping the relay to preclude recurrence of this
event. The revised procedure is scheduled to be effective on or before
December 31, 1987. The licensee's corrective action appears to be
adequate. This item is closed.
(Closed) LER 327/87-020, Potential For Water Spray On IE Electrical
Equipment. In May 1984, TVA identified a design oversight (SQNCEB8409)
which described a condition where class 1E equipment was potentially
unprotected from water spray resulting from failure of piping not
seismically qualified for pressure boundary integrity. In June 1984, a
failure evaluation / engineering report concluded that the condition was
unacceptable but no functional impairment of components was likely. No
immediate actions were taken. In January 1987, a v,iolation was issued for
failure to report this condition per the requirements of 10 CFR 50.73.
TVA calculated a probabilistic risk assessment to determine the frequency
of an earthquake which could cause an unqualified pipe break. A
consultant reviewed the study and concluded the f ailure probability was
within acceptable design limits. An engineering consulting firm was hired
to identify plant areas where a piping hazard existed. These areas were
identified on piping hazard identification data sheets (PHIDS). The PHIDS
were reviewed to determine the 1E electrical equipment in those areas for
Unit 2. Engineering change notice (ECN) L6770 provices the following
corrective actions:
1. Piping and support modification to prevent breaks. I
'
2. Anchoring of mechanical equipment to prevent breaks.
3. Sealing of conduits into 1E electrical equipment to prevent
water from shorting out the equipment.
All work for Unit 2 was completed on July 1987, with quality assurance
(QA) verification completed on July 20, 1987.
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The inspector walked down several of the PHIDS identified areas to
'
determine _the adequacy of the 1E electrical sealant program with the
following results:
-
One conduit on the sealant required list (2A6500) was not sealed
(near the upper head injection (UHI) isolation valve). TVA is
preparing a work request (WR-B-251081) to seal this conduit.
Other items not specifically on the list of conduit to be sealed per the
work plan, but appearing to be related to the LER are as follows:
-
At conduit 2A6525, near Unit 2 UHI isolation valve, the L Box cover
is loose with no gasket installed.
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At conduit 2V35251, near Unit 2_UHI isolation valve, the C box cover
is loose.
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Conduits 2PM3324, 2M31395, 2M3138S, and 2562255, located near the-
Unit 2 AFW turbine local control panel, are not sealed. These are
noted for documentation only, and not as a discrepancy due to work in
progress on this panel during the inspection.
An inspection of portions of the mechanical corrective actions i
accomplished by ECN 6770, Relocation Of Fire Main Nozzles And Restraint Of '
Qualified Piping / Fixtures, indicated the corrective action was adequate.
The few discrepancies identified by the inspector appeared to be
anomalies. Based on the above and discussions with licensee personnel the l
corrective action for this LER appears to be adequate. This item is l
closed.
'
(Closed) LER 327/87-019, Diesel Generator Started From A Loss Of Voltage
On The 6.9 KV Shutdown eoard During A Degraded Voltage Timer Functional
Test Due To An Inadequate Procedure. Surveillance instruction (SI)-307.2,
Degraded Voltage Relay Response Time Test And Timer Calibration, has been
e revised and a test has been run with no initiation of the diesel
generators. .SI-307.1, (the same test for Unit 1) has also been revised.
SI 220.1, Automatic Load Sequence Timer Function Test, for Unit I and
SI-220.2 (the same test for Unit 2) have been revised. Thus far, SI-220.2
has been run with no apparent problems. The licensee's actions appear to
be adequate. This item is closed.
(Closed) LER 327/87-028, Potential Loss Of Control Air. Reported was a
condition which could cause a loss of all control air due to a design
oversight. The licensee reviewed the electrical control logic and
determined a design flaw prevented the auxiliary control air compressors
from starting on a loss of offsite power signal. The corrective action
taken was to install jumpers around the seal-in contacts as a temporary
fix (NC0 870208001 completed July 29, 1987) and to have a final fix by
January 1,1990, to be tracked by NCO 870208002. The licensee's actions
appear to be adequate. This item is closed.
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14
(Closed) LER 327/87-044, Potential Problem with IE Electrical Equipment.
Inadequate communication between design organizations resulting in
potential problem with IE electrical equipment due to unanalyzed flooding
effects. Significant condition report (SCR) SQN NEB 8617 R3 lists items,
identified by Sergeant and Lundy Engineers report SL-4424 dated,
August 29, 1986, revised December 12, 1986, that have the potential to be
adverse to a safe shutdown, due to medium energy line breaks. The SCR
lists those items that TVA has identified to be completed prior to Unit 2
restart and those items justified by: B25 86 1212 008, D. W. Wilson's
memorandum to SQEP files dated December 12, 1986; B25 86 1014 001,
D. W. Wilson's memorandum to DNE files dated October 14, 1986; and B45 87
0327 426, NEB calculations, justification for continued operation with
unimplemented corrective actions for moderate energy line breaks,
SQN-SQS4-0088. The items identified as pre restart have been completed
and verified. The justifications for the pre-restart items have been
reviewed and appear to be adequate. The post-restart items will be
tracked by NCO 870 260 001 and NCO 870 260 002. The licensee's actions
appear to be adequate. This item is closed.
(Closed) LER 327/87-050, Containment Spray Pumps Will Not Deliver The
Design Basis Flow Rate. The TS has been revised (TS change 87-36 revision
1) to require a minimum differential pressure across the pump and a
minimum flow rate through the test loop. Surveillance instruction
(SI)-37, Containment Spray Pump Test, has been deleted and 51-37.1,
Containment Spray Pump 1A-A test; SI-37.2, Containment Spray Pump 18-B
Test; SI-37.3, Containment Spray Pump 2A-A Test; and SI-37.4, Containment
Spray Pump 28-B Test, have been generated to allow testing of eacn pump to
the criteria generated by TVA department of nuclear engineering (DNE) for
each loop. A review of these documents indicates that the licensee's
actions appear to be adequate. This item is closed,.
The following LERs remain open pending licensee corrective action.
(0 pen) LER 327/87-34, Containment Penetrations Identified As Not Meeting
General Design Criteria (GDC) 56. Thirteen containment penetrations were
identified which do not meet GDC 56 due to design criteria
misinterpretation resulting in a potential TS violation. This LER
identified that 13 containment pressure indicating lines had been
installed with test connections sealed with only a threaded tubing cap.
The NRC interpretation of 10 CFR 50, Appendix A, design Criteria 56,
primary containment isolation, did not consider threaded tubing caps to be
acceptable. Corrective action was to install isolation valves in the test
lines, revise surveillance instructicns (SI)-14.1 and 14.2, verification
of containment integrity, to include the additional valves and perform ,
these sis prior to unit startup. In addition, SI-158.1, Containment
Isolation Valve Leak Rate Test, was to be revised to include the
additional valves and the valves were to be tested prior to entering
mode 4 for each unit.
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15
The inspector reviewed the. revisions to SI-14.2 and 158.1. SI-14.2 tests-
have not yet been performed. SI-158.1 tests had been performed
satisfactorily.
The inspector reviewed the procedures and drawings related to the new
valve installation and inspected completed work in the plant. The
following deficiencies were identified:
TVA design criteria SQN-0C-V-2.15, Containment Isolation System,
paragraph 5.1.2.2.b, states that all test vent, drains and test
connections shall, as a _ minimum, include one manual valve and a
capped, threaded nipple. The inspector identified during a plant
walkdown that test connection lines for instruments30-468, 47B and
48B do not have a cap installed. No threaded cap was specified in
design change notice (DCN) X00028A or the implementing workplan (WP)
12630, which installed the isolation valves.
During the plant walkdown the inspector noted that the test lines for
the containment purge system penetrations (ten lines) also did not '
have the required caps installed. This system had been walked down
and evaluated against design criteria DC-V-2.15 during the design
basis verification program (DBVP), but these discrepancies were not
identified.
DCN X00028A and WP 12635 indicate that heating and ventilating air- .
flow diagram 47W866-1 is for reference only and thus does not require '
markup in the control room. However, administrative instruction
(AI)-25, Drawing Control After Unit Licensing, lists this diagram as
a "critical" drawing requiring prompt as-built markup in the control
room. ,
Flow diagram 47W866-1 is inaccurate in that the root valves for all !
containment pressure instruments are not shown. Also, pressure j
instruments30-310 and 30-311 are not shown. )
Mechanical instrumentation and controls drawing 47W600-153 does not
accurately reflect the as built configuration of instrument lines for
30-46A, 47A and 48A in that the vacuum relief line automatic
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isolation valves are not shown as installed between the containment
and the instrument.
2
Potential reportable occurrence (PRO) 1-87-259 states in the generic
implementation evaluation section that Watts Bar does not have a
vacuum relief system; indicating that no further generic review is
required. Although this is a true statement, discussions with design ,
personnel indicate that the other containment pressure instrument i
lines, addressed in the LER as not being properly isolated, are ,
installed at Watts Bar. CAQR 871246, closed on August 18, 1987, i
states that no generic evaluation is required.
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The failure to implement design criteria SQN-DC-V-2.15 on the thirteen
containment penetration test lines is a violation of 10 CFR 50,
Appendix B, Criterion III, Design Control, and is identified as violation
327,328/87-71-01.
The failure to perform proper generic issue evaluations in PRO 1-87-259
and CAQR 871246 with regard to similar containment penetration lines
installed at Watts Bar and the failure to identify and correct the above
deficiencies during the LER review and approval process is identified as
violation 327,328/87-71-02; failure to affect adequate- corractive action
in accordance with 10 CFR 50, Appendix B, Criterion XVI.
(0 pen) LER 327/87-40, Shield Building Mechanical Penetration Seals Not
Qualified Oue To A Material Misapplication. This was issued as an
information LER to notify NRC of a problem related to reactor snield
building mechanical penetration sleeves that might not be hydraulically
leak tight. Therefore these seals might not meet final safety analysis
report requirements for in leakage to the annulus during external flooding
conditions. This problem was identified in significant condition report
(SCR) SQNMEB8702 issued in February 1987. This SCR identified that a
review of previous calculations conducted during the licensee's design
basis verification program indicated that mechanical seals would fail at
design basis flood hydrostatic pressures and that pipe movement would
create a gap between the pipe and the silicone foam seal material, thus
providing a leakage path. This SCR also identified the three basic safety
functions of the shield building mechanical seals to be: (1) preventing
inleakage of water, (2) allowing maintenance of a negative pressure in the
annulus, and (3) as a fire barrier.
TVA performed a probabilistic risk assessment (PRA) which indicated a j
probability of experiencing a design basis flood during the remaining core j
life for Unit 2 was 3.41 E-5. A structural evaluation proved that the ,
vessel could withstand flood level (elevation 724) pressures. Cables and !
cable splices are capable of operation for fourteen days when submerged,
well past the tocal projected flood period of 1 to 6 days.
As the seals in question are also required as fire barriers, TVA performed
an evaluation and determined that the foam sealant would expand when
heated. This evaluation also noted that fire barriers had been
periodically inspected to assure they were intact. Therefore, TVA
concluded that the seals would continue to function as a fire barrier.
TVA also concluded that the negative pressurization capability of the
shield building was not adversely affected by the potential leakage of the
penetration seals, because surveillance instruction (SI)-264, EGTS Annulus
Vacuum Draw Test, is periodically performed. TVA's overall conclusion was
that the worst case condition of the seals would not adversely affect
safety related equipment in the performance of their intended function.
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The specified corrective actions were: (1) to re-evaluate and upgrade
shield building mechanical seals (to be completed during the next
scheduled refueling outage for each Unit, (2) add notes to seal drawings
to define water tight requirements, and (3) establish a design standard
disallowing silicone foam as a seal against liquids and requiring
consideration of other types of seals where pipe movements exceed one
eighth-inch.
The inspector reviewed the various documents associated with this-problem
and interviewed responsible engineering and licensing personnel. The
following deficiencies were identified:
The LER does not address the fact that mechanical penetration seals
were found to be improperly installed (or not installed) as
documented in condition adverse to quality report (CAQR) SQN 870157
issued in March 1987. Improper installation would significantly
impact any safety evaluations relied upon to accept current j
installations with regard to air in-leakage and fire . barrier
functions. !
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The LER conclusions that the fire barrier and air leakage (radiation)
sealing functions of the installed foam seals remain effective, are
based on inappropriate plant testing conditions. Pipe movement which j
would cause the pipe-to-seal gaps, noted in the laboratory tests, is ;
thermal growth at system operating temperatures. The fire barrier !'
inspections and EGTS vacuum tests have been/are performed during cold
shutdown or refueling modes (i.e., not at system operating )
temperature). Thus the ability of the plant to meet Technical !
Specification requirements related to EGTS (3.6.1.9.d.5) and fire
barriers (3.7.12) may not be assured after system heatup.
The LER does not address many of the sixteen recommended corrective
actions specified in significant condition report (SCR) SQN CEB8721
issued in February 1987. This SCR documented the seal inadequacies
, related to the pressure, fire, hydraulic and radiation sealing of
pipe through penetrations. These corrective actions addressed the
adequacies of previous tests, the expected pipe moments, seal
repairs, QA/QC, inspection and test changes, procedure and design
criteria changes, etc. for penetration seals located throughout the
plant, including the shield building.
The operability and generic issue sections of CAQR SQN 870157 have
been completed indicating that operability is not affected and no I
generic review is required. This CAQR also indicates that corrective
action is scheduled to be completed prior to the unit 2 cycle 4
outage and emphasizes that this CAQ and the corrective actions are i
not restart items. However, operability of seals may be affected as l
discussed above, and similar penetration problems have been
identified at Watts Bar (May 1987).
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Potentially reportable occurrence (PRO) report 1-87-254 associated ~
with SCR MEB8702 and this LER indicates that no CAQR should be
written and 'that no unreviewed safety question determination (USQD)
is required. Considering only the question of the flood protection
function of these seals, a USQD would be required to accept seal
installations that do not meet the design criteria specified in FSAR
section 6.2 until TVA could re-evaluate and upgrade these seals by
the next refueling outage.
In summary, .this LER does not accurately reflect the magnitude of the
problem with mechanical penetrations seals. The evaluations .and
conclusions in this LER, PRO 1-87-254 and'CAQR 870157, with respect to
safety function operability, were inadequate. The aforementioned
evaluations and conclusions do not appear to be consistent with the
extensive list of tests, evaluations and program changes specified in the
previously issued SCR SQN-CEB-8721. Resolution of the LER is still under
NRC review, and is not Unit 2 heatup related. However, this LER will be
resolved during the closure of order EA85-49 prior to the startup of
Sequoyah Unit 2.
(0 pen) LER 327/87-025, Failure To Meet The Surveillance Requirements For
The Hydrogen Mitigation System Due To An Inadequate Procedure. Unit.2
surveillance instruction (SI)-305.2, Hydrogen Mitigation System
Operability, has been revised to define specific regions for redundant i
circuit operability determinations. SI-305.1 (for Unit 1) has not yet
been revised, but will .>e prior to unit startup. The plant operations
review staff has cetermined that TS surveillance requirement (SR) 3.6.4.3.a allows two igniters per train to be inoperable, where
previous NRC documents (NUREG-0011, supplement 6 and the safety evaluation
for TS change 24) allow for only one igniter per train to be inoperable.
The acceptance criteria in SI-305.1 and 305.2 requires 33 of 34 igniter,
to be operable (i .e. , the more conservative position). A change to
SR 3.6.4.3.a is being initiated by TVA. Actions required for Unit 2
restart have been completed. Pending revision of TS ana $1-305.1 for
Unit I this item remains open.
(0 pen) LER 327/87-037, Engineered Safety Features Equipment Coolers'
Capacity Determined Inadequate For LOCA Conditions Due To HVAC Calculation
Errors. The deficiencies identified in this LER resulted from the use of
Watts Bar heat load figures in the original Sequoyah calculations.
Recalculation for safety related applications using the latest Sequoyah
data indicated that eight coolers, supplying three plant areas, had
inadequate capacity. Corrective actions were to: (1) update drawings with
the proper environmental conditions and (2) with required air and cooling
water flow rates, perform tests and adjustments to ERCW flow to assure the
required cooling and revise surveillance instruction (SI)-566, ERCW Flow
Verification Test, to reflect the required ERCW flows. In some cases, to
obtain the required flows, modifications and maintenance on the coolers
was required. SI-566 has been revised. Consolidation of design functions
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19
and a strengthened design change control program should prevent
recurrence.
TVA has identified additional area coolers that have insufficient air flow
rates and will submit a revision to this LER to document these findings
and the actions required.
In addition, this LER reported that non-safety related (non-seismically
qualified) backdraf t dampers had been found installed in the auxiliary
building elevation 714 penetration rooms (two dampers per unit). This
improper installation had been identified during repair / replacement
efforts to correct severe deterioration of these dampers. The current
issue of this LER does not include this damper problem as part of the
title or the abstract and analysis sections of the report. This omission
will be corrected by TVA in the revision to this LER. The stated cause of
the improper damper installation was that the purchase requisition did not
distinguish between safety related and non-safety related equipment. It
is apparent to the inspector that there also were inadequate controls on
the damper installation and inspection process. NRC inspector review of
condition adverse to quality report (CAQR) SQP870935 and discussions with
responsible personnel indicated that no further research, inspection, or )
evaluations had been performed to determine if this problem could exist j
in other safety related damper locations. This particular purchase l
requisition involved 87 backdraf t dampers, both safety and non-safety I
related. TVA has committed to a more detailed evaluation of this CAQR and
to address this issue in the revision to the LER. This item will remain
open until TVA completes this evaluation and submits the revised LER.
(0 pen) LER 327/87-033, Inadequate Design Calculations For Vital Battery V.
Adequate design calculations are unavailable for the 125 VDC vital
Battery V, causing potential degrading of the system. The licensee has
issued a hold order (1606) on 125 volt DC battery board No. V. This hold ,
order will be in effect until the deficiencies associated with vital l
Battery V have been resolved. This LER is therefore, no longer considered i
e a restart item. This LER will remain open pending resolution of
deficiencies.
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(0 pen) LER 327/87-048, Failure Of Silicone Rubber Insulated Cabies During l
Testing. Failure of the cables is the subject of continuing TVA-NRC
inspection / analysis. Pending final resolution of the corrective actions
required by the inspection / analysis this LER remains cpen.
9. Event Followup (93702, 62703)
The licensee identified an unused and uncapped 6-inch penetration sleeve
located through the auxiliary building roof directly above the spent fuel
pool at elevation 731.75 directly below the roof mounted demineralized
water storage tank. The sleeve pipe stub was taped over with duct tape
and covered with a black foam bonnet such that normal visual observation
failed to reveal that the sleeve was not capped. The penetration was
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discovered as part of an ongoing licensee program to inspect penetration /
sleeve seals (CAQR SQN8713161DI). Based on the condition of the rubber
bonnet and the fact that the associated drawing deficiency as shown on TVA
drawing 47W491-1 has e> isted since 1975 (Rev. 3), the licensee believes
that the condition has existed since at least that time. The penetration
was believed to have been abandoned during the construction phase of the
plant and has been in its present condition since that time. This item
has been documented by the licensee on CAQR SQP871624. The inspectors
will follow the corrective action of this item and look at performance of
surveillance testing on the auxiliary building gas treatment system in the
future.
No deviations or violations were identified.
10. Inspector Followup Items
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Inspector Followup Items (IFI) are matters of concern to the inspector l
which are documented and tracked in inspection reports to allow further i
review and evaluation by the inspector. The following IFIs have been i
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reviewed and evaluated oy 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.
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(Closed) IFI 327,328/85-26-07, Follow-up Licensee Directives For j
Interdi sciplinary Groups Reporting Plant Conditions. A technician had
returned power range (PR) nuclear instrument (NI) drawer N-42 to what he
assumed to be an operable condition. He commenced surveillance activities
on PR NI N-41 without reporting to the RO or SR0 that he had not complied
with an assigned procedure and had af fected safety,-related equipment not
authorized for surveillance activities. This caused a unit trip from
power.
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Administrative instruction AI-12, Adverse Conditions And Corrective
Actions, states that plant personnel shall report any suspected abnormal
plant condition adverse to quality in the performance of their regular l
work duties. A f ailure to follow procedure is defined in section 4.0 of
AI-12 as a condition adverse to quality.
On June IS, 1987, a memo was issued from the plant manager to the line
managers on tne requirements for conduct of testing. Administrative
instruction AI-47, Conduct Of Testing, defines the role of the test
director. The test director is an individual with thorough knowledge and
adequate abilities gained through formal training and/or experience to
perform and/or direct the test. Part of the training and knowledge of a
test director is to report abnormal plant conditions to the shift
engineer. The inspectors have watched numerous testing efforts since
AI-47 training was conducted by the licensee and have not noted any
additional problems.
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Inspection report 87-30 reviewed the conduct of the testing program at
Sequoyah. The conduct of the testing program will be addressed under item l
327,328/87-30-05. IFI 327,328/85-26-07 is closed based on the specific
corrective actions affected by the licensee. ,
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(Closed) IFI 327,328/85-47-06, Review Of Non-Safety Related Air Supply To
Spent Fuel Pit / Transfer Canal. Barrier Door. This item was opened to track
an event documented as non-safety related and not reportable. During a
subsequent NRC inspection of this same event and TVA's evaluation of it,
IFI 327,328/86-60-04 was opened to track ' this and similar events. As
discussed below, IFI 327,328/85-47-06 has been evaluated and closed as
part of the closure process of IFI 327,328/86-60-04.
(Closed) IFI 327,328/86-60-04, Non-Safety Related Activities That Affect
Safety Related Activities. This item provided for tracking incidents i
where non-safety related systems and activities related to them have l
challenged or had some adverse affect on safety related systems. This. !
item resulted from events such as the December 1985 spent fuel pit i
draindown and the auxiliary building isolation reported on LER 327/86-03. l
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The inspector reviewed approximately 200 Unit 2 potentially reportable
occurrence (PRO) reports from 1986 and 1987, and approximately 50 Unit 1 l
PRO reports from 1987 for similar occurrences. A plant operations review
staff (PORS) supervisor and other NRC inspectors were interviewed l
regarding their perception of this issue. Although a number of incidents i
have occurred during the last two years involving non-safety related y
equipment affecting safety-related equipment (e.g., PRO-2-86-36,
PRO-1-87-147 and 301) . These incidents appear to be isolated cases and '
are not indicative of any major programmatic weakness. Although incidents
of these types are still of concern to the NRC and will continue to be
followed by NRC inspectors, formal tracking is no longer considered I
necessary. This item is closed. I
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(0 pen) IFI 328/86-62-02, Closure Of Engineering Change Notices (ECN) Prior
To Restart. This item was identified and discussed in inspection report
327,328/86-62 and involves TVA's ECN closecut schedule. Specifically, the
inspectors reviewed the ECN closure process and determined that as of
November 1986, the licensee had closed only approximately 100 of the then
approximately 1000 field completed ECNs. The licensee was requested in a
letter from G. Zech to S. White (dated December 18, 1986) to address their
justification for not requiring the complete closure of all open ECNs
prior to restart. In their February 3, 1987 response, TVA addressed the
fact that although a large number of ECNs were not completely closed out, I
the physical work authorized by these ECNs was accomplished through the
modification process (i.e. , work plan) and the work plans, if identified
as restart issues, have been or will be verified as field complete prior I
to restart. Additionally, the licensee pointed out that the problems that
were identified during the closure of the 100, randomly selected, ECNs
were mostly minor drawing errors and would not impact system operability.
The licensee also stated that in addition to the 100 ECNs reviewed, the
closure group reviewed the primary drawings associated with approximately l
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- - - - -- - -. .. . . - - . - - -
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450 additional safety related ECNs and initiated corrections as required
to ensure control room drawings reflected the as installed system. j
Af ter reviewing the above submittal the NRC met with TVA and requested a
sucplemental response to address the following: (1) how representative the
100 closed ECNs were to the general population of ECNs, as to discipline i
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and complexity; (2) had TVA identified any common problem with drawing
update (i.e., type of drawings involved); and (3) provide TVA's manpower
allocations required to support the closure of the ECN backlog at Sequoyah
by October 1988 as committed to in TVA's original response. The licensee ,
provided the requested information in their May 22, 1987 supplemental l
response. The information provided satisfied the above request and is
currently being reviewed by the NRC and approval of the licensee's
position will be the subject of separate correspondence. This restart item
remains open pending approval of the licensee's position that ECN closure
is not a restart issue.
'11. Plant Operations Review Committee (PORC) (40700)
The inspector conducted a functional audit of the PORC which performs as
the onsite safety review committee. This functional review was intended ,
'
to evaluate if the activities of PORC could support the heatup and even-
tual startup of Unit 2. The site has implemented a change to a qualified !
reviewer concept as a result of TS change 87-34. The inspectors observed l
the PORC review proposals which affected nuclear safety including issuance
of plant procedures and changes thereof, modifications to systems and
equipment, and unreviewed safety question determinations (USQD).
The following documents were reviewed in connection with the functional
audit of PORC activities.
Administrative instruction (AI)-4, Preparation, Review, Approval
And Use Of Site Procedures And Instructions
AI-19, Part IV, Plant Modifications After Licensing
AI-43, Independent Qualified Review
AI-48, Plant Operations Review Committee Charter Standard Practice
(SQA)-21, Review Implementation And Reporting Of Nuclear Experience
Review Information
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SQA-119, Unreviewed Safety Question Determination
TS change 87-43, Revision 62, Amendment 53, dated September 10,1987.
As a result of reviewing the above documents and observing PORC activi-
ties, it was determined that the licensee was adequately implementing the
current TS requirements. The only observation that remains concerns the
number of regular members of PORC and the PORC quorum. TS change 87-43
. - - -
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increased the total number of PORC members from the original seven to st
least eleven. However, the quorum remained at five. Observed PORC
activities indicate that the chairman was routinely substituted for by an
alternate and there is a potential for a loss of continuity in PORC
activities. This issue was discussed with the NRC licensing project
manager and OSP management on November 17, 1987. In addition, this issue
was discussed with the new plant manager (Mr. Smith). .His stated inten-
tions were to strengthen PORC and to attempt to chair as many PORC
meetings as possible. The following violations are closed as a result of
the above review:
(Closed) violation 327,328/86-20-10, Improper PORC Composition. The
inspector reviewed the licensee's corrective actions stated in the viola-
tion response (Shell, Grace) dated May 23, 1986. Based on a review of
licensee corrective actions and the reviews stated in the above
paragraphs, this item is closed.
(Closed) violation 327,328/86-46-07, Review Of PORC Documents By Temporary
Members. The inspector reviewed the licensee's corrective actions stated
in the violation response (Gridley, Grace) dated December 5,1986. Based
on a review of licensee corrective actions and the reviews stated in the
above paragraphs, this item is closed.
12. Independent Safety Engineering Group (ISEG) (40700)
The inspector reviewed the activities of the ISEG for the period of July ,
1, 1987 through August 31, 1987. The activities appear to be consistent l
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with the charter stated in the Technical Specifications. The inspector
had no further questions.
13. Independent Inspection (61700)
The inspector reviewed the licensee's actions taken to implement correc-
tive action for a condition identified in CAQR SQP-870525. The identified
condition was that surveillance instruction SI-118 did not include the
, testing of the auxiliary feedwater system (AFW) bypass level control
valves, as is required by TS 4.7.1.2.b.1. The stated corrective action was i
to revise SI-118 to include the testing of these valves. J
The inspector reviewed the current revision of SI-118, and determined that
this instruction now adequately satisfies the requirements of the above
referenced Technical Specification. However, the licensee's review of the
CAQR for repo: .ibility is determined to be inadequate. The following is a
discussion of this review-
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a. On March 25, 1987, PRO 1-87-129 was initiated to identify that
the bypass level control valves (LCVs) were not being tested.
b. On March 21, 1987, the "operator action taken on event" section
of the PRO was filled in to state that a revision to SI-118 had ;
. . . . - _ - - - - - - . _ .
_ . _ . __. _ __ . . . _. _ __ _
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been initiated to include testing of the bypass LCVs, and the l
"NRC notified" block was checked "no". l
c. On April 2, 1987, Part "B" of the PRO was checked as
"reportable". l
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d. On April 3, 1987, the CAQR was initiated.
e. On April 9, 1987, CAQR operability /reportability assessment
sheet, Rev. O, was completed, stating potential reportability to
be "indeterminate".
,
f. On April 15, 1987, the "reportability" block on sheet 2, Rev. O,
of the CAQR was checked '*yes" by the responsible organization
(OPS). i
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g. On April 16, 1987, CAQR operability /reportability assessment I
sheet, Rev. 1, was completed, stating potential reportability to I
be "no" with the justification that "SI-118 has and does meet
the intent of SR 4.7.1.2.b.1". l
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h. On May 21, 1987, the "reportability" block on sheet 2, Rev. :1,
,
of the CAQR was checked "no", and therefore, this condition was l
not reported to NRC. '
The inspector has reviewed SI-118, Rev.14 (current at the time) and this
review revealed the following. TS surveillance requirement 4.7.1.2.b.1
states to verify "that each automatic valve in the flow path actuates to !
its correct position upon receipt of an auxiliary feedwater actuation test l
signal". Although SI-118, Rev.14 required the bypass LCVs to be tested
to actuate on receipt of a Low-Low Steam Generator Level signal, they were
not being tested to actuate on receipt of the other four . auxiliary
feedwater actuation test signals (i .e. , safety injection, blackout, both
main feedwater pump turbines (MFPT) tripped, and one MFPT tripped with the
plant at or above 80%). Contrary to Technical Specification 4.7.1.2.b.1,
the testing required to verify operability of the AFW LCVs had not been
performed. Violation 327,328/87-71-03 is being issued for failure of
SI-118 to include all appliccole surveillance testing requirements for the
AFW LCVs. Although the above procedure inadequacy was originally identi-
fied by licensee personnel, this cannot be considered a Licensee identi- ,
fied violation in accordance with NRC policy as stated in 10 CFR 2,
Appendix C,Section V.G. , because the violation was not reported to NRC '
due to an inadequate technical review for reportability. Contrary to
10 CFR 50.73(a)(2)(1)(B), the licensee has not reported a violation of
Technical Specifications to NRC via issuance of an LER. Violation
327,328/87-71-04 is being issued for failure to report a violation of a
Technical Specification.
In addition, during this review of AFW level control valves, the inspector
noted the following inconsistencies between statements contained in
section 10.4.7.2.5 of the FSAR and the actual operation of the auxiliary
.
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feedwater system: (1) The FSAR states that the three AFW pumps start and
the flow path opens upon receipt of four stated signals. Also stated, is
that level control valves receive an "opening signal" on any of the pump
start signals. The actual operation of the system is that the pumps start
upon receipt of any of the four stated signals and the level control
valves receive an "arming" or "enabling" signal on any of the pump start
signals. Fowever, the level control valves only open upon receipt of a
low-low level signal from the steam generator (s), thereby establishing an
open flow path; (2) The FSAR implies that each pump discharge line
contains a bypass level control valve. In fact, these valves exist only in
the discharge lines from the two motor-driven pumps; and (3) Table
10.4.7-1 of the FSAR does not list the four bypass level control valves as
being seismic category I valves within the AFW system. Conversations with
various licensee personnel, in conjunction with a review of the perfor-
mance requirements of the AFW system, indicate that the system functions
properly. The licensee has, therefore, initiated action to update the
applicable section and table in the FSAR to accurately describe the
functioning of the system and the valves involved. This action is docu-
mented via form SQA180, Attachment 1, which was initiated on November 19,
1987.
14. Generic Letter 85-28, TI 2500/20
This inspection utilized the guidance provided in NRC temporary instruc-
tion (TI) 2500/20, generic letter (GL) 85-28 and 10 CFR 50.62.
TVA committed to installation and operation of the 10 CFR 50.62 required
modification by letter (TVA to NRC), dated October 11, 1985. Unit 1 is
scheduled for completion by the end of the cycle 4 refuel outage. Unit 2
is scheduled for completion by the end of the cycle 3 refuel outage. The
Unit 2 schedule was revised in TVA letter to NRC dated October 31, 1986.
The revised Unit 2 schedule requires the installation be ccmplete by the
end of the cycle 4 refuel outage. The design selected, per the October
11, 1985 letter, reflects the first of the generic functional designs
developed by the Westinghouse owners group (WOG). NRC informed TVA by
letter dated, September 24, 1986, that the WOG topical report (WCAP-
10858), generic design package is acceptable; however, plant speci fic
details were needed to complete the review. TVA agreed by letter dated,
October 31, 1986, to provide the requested details by February 15, 1987.
TVA provided to NRC, by letter dated February 17, 1987, the plant specific i
details. The TVA plan appears to be adequate. However, NRC's response to I
the plant specific details has not been received.
TVA's nuclear quality assurance manual, part 1, section 1.3, defines the
TVA requirements for limited QA program requirements. Sequoyah standard
practice instruction SQA-172, QA Requirements for ATWS Equipment That Is <
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Not Safety related, implements the above portion of the NQAM. The
purpose of SQA-172 is to describe the QA controls required to meet the
intent of 10 CFR 50.62. SQA-172 was established following the guidance of
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GL 85-28 and appears adequate. In general SQA-172 utilizes existing
procedures.
Design criteria SQA-DC-V-20.0 provides the design basis for the AMSAC
system. Section 3.1 specifies that the implementation of this system
shall not degrade existing safety related systems. The design for the
AMSAC system is approximately 98*, complete with a minor change required as
proposed by human factors review. Procurement of required devices is in
progress.
Equipment change notice (ECN) number L6478 is assigned to accomplish this
modification. However, the design group has not transferred the design
package to the modifications group for the development of the work
package.
TVA does not plan any Technical Specification changes to support the
10 CFR 50.62 rule. Plant procedures and administrative controls will be
used to ensure reliable system operation.
The inspection of TI2500/20 is incomplete due to the stage of implemen-
tation and the schedule for start of actual work to accomplish the
modification. Future inspection will be conducted when the work package
for ECN L6478 is developed. Due to the accepted long term schedule for
completion of this modification, this item is not considered a restart !
item. l
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15. Restart Test Program Review l
,
The inspector reviewed the functional review matrix (FRM) for system 67, j
emergency raw cooling water. The system is described in Section 9.2.2 of i
the FSAR and Section 3.7.4 of the Technical Specifications. 30 safety
related functions were randomly selected f rom the list of 715 functions
identified by the licensee. The applicable references (surveillance
instructions, special test instructions, post maintenance / modification
testing or preoperational testing) associated with each of these 30
functions were then reviewed to determine the adequacy of the licensees
review. The inspector noted that in each case the function was properly
verified by the applicable referenced test.
During interviews with the two restart test engineers (RTEs), the
inspector determined that they were qualified to perform the reviews and
that their work satisfied program requirements. Both engineers demon-
strated a thorough knowledge of the system. Additionaily, the inspector
determined that any identified test requirements were ounchlisted against l
the final package approval and would be tested and resolved prior to
closure.
The inspector attended each of the 3 joint test group (JTG) meetings held
associated with this system. The system being larger and more complex
than other systems required more preparation effort and involved a rather
extensive review by JTG. Although numerous administrative and word
- - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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changes were made to the package prior to final acceptance by JTG, no
major technical issues were identified. The inspector had no further
questions.
16. Review of General Operating Instructions (61700)
The inspector reviewed G01-1, Plant Startup From Cold Shutdown To Hot
Standby, and G01-2, Plant Startup From Hot Standby To Minimum Load. The
review was performed to determine the adequacy o' f actoring procedural
revisions into the operator training program and to ensure operators have
been trained during this extended outage to support plant heatup.
The inspector noted on page 11 of GOI-2, that the minimum source range
meter (SRM) count rate requirement for startup had been changed from 2 cps
to 1/2 cps. This change was based on guidance on initial criticality
obtained from regulacory guide 1.68, Initial Test Programs For Water-
Cooled Nuclear Powee Plants. However, the inspector noted that control
room SRM count rate instrumentation channels do not read below 1 count.
Licensee personnel stated that minimum count rate could be verified by
other means and clarifying instructions would be added to the GOI prior to
plant start-up.
Two inspectors attended a portion of the cycle 5 licensed operator and STA
classroom training. Material covered included G01-2 and restart test
instruction, RTI-3, initial criticality, and a review of subcritical
multiplication, effects of extended shutdown on core reactivity. The
inspector determined from the observed training, review of lesson plan
EGT222, and interviews with licensed operators and training personnel that
the licensee's training prograni was adequate for covering outstanding
revi sions to G01-2. However, no training has been conducted on GOI-1.
The inspector determined from conversations with licensee training
personnel that the licensee plans to conduct the training on shif t prior
to plant heatup. As this training is not scheduled, the inspectors plan
to continue to inspect in this area during subsequent inspections. The
inspector had no further questions.
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