ML20235T051
| ML20235T051 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 09/23/1987 |
| From: | Brooks C, Ignatonis A, Menning J, Patterson C, Paulk G, Postonbrown M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20235T033 | List: |
| References | |
| 50-259-87-30, 50-260-87-30, 50-296-87-30, NUDOCS 8710090244 | |
| Download: ML20235T051 (16) | |
See also: IR 05000259/1987030
Text
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UNIT ED STATES
[ p atop 'o .
NUCLEAR REGULATORY COMMISSION
[" g REGION 11
g
j 101 MARIETTA STREET. N.W.
's ATLANT A, GEORGI A 30323
\...../
Report Nos.: 50-259/87-30, 50-260/87-30, and 50-296/87-30
Licensee: ' Tennessee Valley Authority
6N 38A Lookout Place
1101 Market Street
Chattanooga, TN 37402-2801
Docket Nos.: 50-259, 50-260, and 50-296
License Nos.. DPR-33, DPR-52, and DPR-68
Facility Name: Browns Ferry Nuclear Plant
Inspection at Browns Ferry Site near Athens, Alabama
Inspection Conducted: August 1 - 31, 1987
Inspectors: \. Ob [ 23 f7
G. L. Paulk, Senior Resident Inspector Date Sig'ned~
O. <r . hem.
C. A. Patterson, Resident Inspector
9lb 3 10
Date Signed
N '
'/
C. R. Brooks, Resident Inspector D' ate Signed
bL1- %
J. Mennir@, Resident Itlspecubr, Hatch Nuclear Plant
[41 Or
23/F)
a te signed
%_ Ac--t 9/d2/P'7
M. Poston ' Brown, Project (hginep Date Signed
Approved by: 8 d. Iw1, S/A3/K7
A. J. IdjdatonisT/Section Chief Date Si'gn96
Inspection Programs
TVA Projects Division
SUMMARY
Scope: This routine inspection was in the areas of operational safety,
maintenance observation, surveillance testing observation, reportable
occurrences, action on previous enforcement matters, Surveillance Instruction
upgrade program, Restart Test Program, Employee Concerns Program, Restart
Review Board, FSAR updates and facility modifications.
Results: One deviation was identified for failure to perform and make
available for review adequate justification for changes to the FSAR.
8710090244 G70929
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PDR ADOCK 05000259
G PD9
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REPORT DETAILS
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1. Licensee Employees Contacted:
H. G. Pomrehn, Site Director i
- J. G. Walker, Plant Manager
P. J. Speidel, Project Engineer
- J. D. Martin, Assistant to the Plant Manager
R. M. McKeon, Superintendent - Unit 2
J. S. Olsen, Superintendent - Units 1 and 3
T. F. Ziegler, Superintendent - Maintenance
D. C. Mims, Technical Services Supervisor
J. G. Turner, Manager - Site Quality Assurance
- M. J. May, Manager - Site Licensing
- P. P. Carier, Compliance Supervisor
A. W. Sorrell, Health Physics Supervisor
R. M. Tuttle, Site Security Manager
J. R. Kern, Fire Protection Supervisor .
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- D. A. Pullen, Office of Nuclear Power, Site Representative
Other licensee employees contacted included licensed reactor operators,
auxiliary operators, craftsmen, technicians, public safety officers,
quality assurance, design and engineering personnel.
- Attended exit interview
2. Exit Interview (30703)
The inspection scope and findings were summarized on August 31, 1987, with
the Plant Manager and/or Superintendents and other members of his staff.
The licensee acknowledged the findings and took no exceptions. The
licensee did not identify as proprietary any of the materials provided to
or reviewed by the inspectors during this inspection.
3. Licensee Action on Previous Enforcement Matters (92702)
(Closed) Open Item (259/79-04-02) This item concerned ground water leakage
into the reactor building. TVA revised its response to this item in a
letter from J. A. Domer to J. Nelson Grace dated January 28, 1987. This
letter gave a brief history of the yard dewatering pumps situation and
stated TVA's intention of assessing groundwater inleakage with the
objective of permanently removing the dewatering pumps from service. On
April 30, 1987, TVA submitted a followup response in a letter to the NRC
Document Control Desk from R. L. Gridley. TVA decided to permanently
remove the dewatering pumps " sm service. In the reactor building, 370
feet of concrete cracks wet . repaired by pressure injection grouting.
Regular surveys of the observed inleakage have been too small to be
measured. TVA has assessed that the groundwater inleakage during a
probable maximum flood would not be large enough to affect the operation
of safety-related equipment. This item is closed.
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(Closed) Violation (259, 260, 296/84-15-10) This item was that old fuel
- racks were removed from the Unit 2 fuel pool without the use of a detailed
or adequate work plan to address the task action requirements or
procedural steps. The licensee revised the workplan to clarify work -
instructions. Browns Ferry Standard Practice 8.3 concerning modifications
was revised to provide clearer responsibilities for all parties involved "'
in work plans and to clarify the work process from initiation to
completion. This item is closed.
(Closed) Inspector Followup Item (259, 260, 296/84-32-01) This item was
a concern regarding the reference point used to determine water level in
the scram discharge instrument volume. The inspector questioned if the
l reference point was based on survey marks located on the adjacent building
l structure, or were based on poir.ts measured from the floor. The licensee
l was in the process of establishing new survey marks based on permanent
tank components. Calibration reference points were established using the
center line of the lower element sensing line as the zero point. The
other points are then calculated using the elevations referenced on
drawings. The inspector reviewed plant procedures SCI-185.9 and SI4.1.A-8
for the method of calibration. This item is closed.
(Closed) Unresolved Item (259, 260, 296/86-32-09) This item was whether
the " Rigging Fundamentals" course discussed rigging selection techniques
for lifts which involve I-beams. Mechanical Maintenance Instruction
MMI-102, Rigging Equipment Program contains a caution concerning rated
capacities fgr certain size wire rope and loads. The licensee revised the
lesson plan for rigging to include discussion of this caution for lifts
which involve I-beams. This item is closed.
(Closed) Inspector Followup Item (259, 260, 296/86-32-11) This item
concerned potential problems found in the rigging cage at Elevation 639 on
Unit 2. Some hooks were identified that did not have safety latches and
were not tagged with a defective equipment tag. One generic defective tag
was attached to the rigging cage to cover all defective items in the cage. '
Also, the rated capacity table for wire ropes posted at Elevation 639 was
in error in that the line of ratings corresponding to 1/2 inch ropes was
labeled 1/4 inch. The licensee removed or tagged all defective hooks
separately. The error on the capacity table was corrected. This item is
closed.
4. Unresolved Items * (92701)
Two Unresolved Items were identified during this inspection. The first
one relates to an LER which reported that the Control Room Emergency
Ventilation System (CREV) could have failed to perform its intended
function. This is addressed in paragraph 8. The second one relates to
the lack of formalized methods for implementing the ECP policy and is
discussed in paragraph 10. ,
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- An Unresolved Item is a matter about which more information is required to
determine whether it is acceptable or may involve a violation or deviation.
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5. Operational Safety (71707, 71710)
The inspector . were kept informed of the overall plant status and any
significant safety matters related to plant operations on a daily basis.
Weekly discussions were held with plant management and various members of
the plant operating staff.
The inspectors made routine visits to the control rooms when an inspector
was on site. Observations included instrument readings, setpoints and
recordings; status of operating systems; status and alignments of emergency
standby systems;- onsite and offsite emergency power sources available for
automatic operation; purpose of temporary tags on equipment controls and
switches; annunciator alarm status; adherence to procedures; adherence to
limiting conditions for operations; nuclear instruments operable;
temporary alterations in effect; daily journals and logs; stack monitor
recorder traces; and control room manning. This inspection activity also
included numerous discussions with operators and their supervisors.
General plant tours were conducted on at least a weekly basis. Portions
of the turbine building, each reactor building and outside areas were
visited. Observations included valve positions and system alignment;
snubber and hanger conditions; containment isolation alignments;
instrument readings; housekeeping; proper power supply and breaker
alignments; radiation area controls; tag controls on equipment; work
activities in progress; and radiation protection controls. Informal
discussions were held with selected plant personnel in their functional
areas during these tours.
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 to include; protected and vital areas access controls,
searching of personnel, packages and vehicles, badge issuance and
retrieval, escorting of visitors, patrols and compensatory posts. In
addition, the inspectors observed protected area lighting, protected and
vital areas barrier integrity.
The inspectors witnessed performance of Special Test 87-25, Disposal of
Sodium Pentaborate Solution into Discharge Canal of August 28, 1987.
Several of the limitations, prerequisites, and precautions were noted to
be vague and subject to interpretation. For example, one statement
required that at least two condenser circulation water pumps (CCW) be
operating. Although, the intent of this step was to assure that two pumps
were operating on the Unit 3 discharge conduit, the performer merely
verified that four pumps were running without regard to the lineup (it is
possible to have up to six CCW pumps running without any being on the
Unit 3 discharge conduit). Although, one prerequisite of the special test
required that the Shift Engineer verify that two CCW pumps were on and
would remain on throughout the test, the Shift Engineer was not aware that
the special test was in progress when qui::ed by the inspector. The
inspector also noted that the special test did not comply with existing
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writers guides and style guides currently in use for the procedure upgrade
program. There were no page numbers, no indication of when the last page
was reached, no sign off steps for notification to shift engineers or
reactor operators, and no second party verification of calculations.
These issues were discussed with licensee management personnel as items of
generic concerns.
6. Maintenance Observation (62703)
Plant maintenance activities of selected safety-related systems and
components were observed / reviewed to ascertain that they were conducted in
accordance with requirements. The following items were considered during
this review: the limiting conditions for operations were met; activities
were accomplished using approved procedures; functional testing and/or
calibrations were performed prior to returning components or system to
service; quality control records were maintained; activities were
accomplished by qualified personnel; parts and materials used were
properly certified; proper tagout clearance procedures were adhered to;
Technical Specification adherence; and radiological controls were
implemented as required.
Maintenance requests were reviewed to determine status of outstanding jobs
and to assure that priority was assigned to safety-related equipment
maintenance which might affect plant safety. The inspectors observed the
below listed maintenance activities during this report period:
a. Post maintenance testing on C3 Residual Heat Removal Service Water
(RHRSW) pump. During this test, the licensee detected a problem with
the electrical integrity of the motor windings. Subsequent followup
activity determined that a junction box containing splices had become
flooded apparently from the ' open conduit associated with the pump
motor maintenance. Troubleshooting activities are still ongoing.
b. Discharge of Sodium Pentaborate waste water to Unit 3 discharge CCW l
conduit.
No violations or oeviations were observed in this area.
7. Review of Surveillance Instructions (61726) l
The inspector reviewed selected surveillance instructions (sis) that are
being evaluated by the licensee in accordance with procedure SDSP-2.14,
" Surveillance Instruction Evaluation." This procedure provides for the
verification, independent review, walkdown, and validation of sis. The
licensee indicated to the inspector that 467 sis have been identified as
requiring evaluation prior to Unit 2 restart. As of August 19, 1987, 54
of these sis had been evaluated to the point of being PORC reviewed and
approved. The inspector's review of the sis focussed on documentation for
the SDSP-2.14 evaluations, technical adequacy, and compliance with
technical specification requirements. The below-listed SI's were reviewed
by the inspector.
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Procedure Number Title Revision
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SI-4.2.K.2A Reactor Building Vent Exhaust 0
Radiation Monitor (2-RM-90-250)
Functional Test
SI-4.2.K-07 Reactor Building Vent Exhaust 0
Monitor (2-RM-90-250) Calibration (
SI-4.7.A.2.g-3/36b Primary Containment Local Leak Rate 0
Test Reactor Feedwater Line B:
Penetration X-9B (
SI-4.7.A.2.g-3/32a Primary Containment Local Leak Rate 0 .
Test Control Air: Penetration X-48
SI-4.4.A.1 Standby Liquid Control Pump
Functional Test
0 0}
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SI-4.4.A.2 Standby Liquid Control System 0
Functional Test
SI-4.9.A.2.a-2 Weekly Check for Shutdown Board C & D 1
Batteries Surveillance Instruction
SI-4.9.A.2.b-2 Quarterly Check for Shutdown Board 1
C and D Batteries Surveillance
Instruction
SI-4.9.A.2.a Weekly Check for Diesel Generator A, 1
B, C, and D Batteries Surveillance
Instruction
SI-4.9.A.2.6 Quarterly Check for Diesei Generator 1
A,B,C, and D Batteries Surveillance
Instruction
The SI evaluation documentation provided to the inspector along with these
instructions was in some cases either incomplete or inconsistent. For ,
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instance, the documentation for 2-SI-4.2.K.2A indicated that the
instruction had been validated on June 13, 1987, with no evidence of prior
verification, independent review, or wal kdown. The documentation for ,
2-SI-4.7.A.2.g-3/36b and 2-SI-4.7.A.2.g-3/32a did not provide evidence of
verification, independent review, or walkdown. The status and tracking
sheets for these SI's indicated that these evaluation processes had been
completed. The documentation for 2-SI-4.4.A.1 and 2-SI-4.4.A.2 only
reflected PORC review and approval. The inspector understood that the
licensee is in the early stages of the SI evaluation program and did not
expect to find complete documentation for all SI's currently in process.
However, the inspector was concerned that the licensee would not be able
to ultimately assemble evaluation packages following SI validations unless
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a more concerted effort is made to collect and maintain evaluation
documentation. This concern was expressed to licensee representatives.
In reviewing 2-SI-4.9.A.2.a-2, 2-SI-4.9.A.2.b.2, 0-SI-4.9.A.2.a, and
0-SI-4.9. A.2.b, the inspector noted that the same individual verified,
independently reviewed, and walked down these procedures. This practice
does not give the licensee the benefit of independent reviews and is
inconsistent with the guidance of SDSP-2.14. The inspector also noted
that Section 5.0, "Special Tools and Equipment Recommended", of each of
these procedures does not identify the device (s) to be used to measure the
temperatures of the battery pilot cells. Other required measuring
instruments are listed. These matters were discussed with the licensee
and will be tracked as Inspector Followup Item 50-260/87-30-05 (Review of
Surveillance Instructions).
The inspector observed, while reviewing 2-SI-4.7.A.2.g-3/36b and
2-SI-4.7. A.2.g-3/32a, that Precaution 3.1 of these instructions specifies
that personnel performing the instructions will be qualified to at least
NDE Level I in the rotameter method. These instructions should require
certification and should identify the document and specific edition to
which certification is to be made. This matter was discussed with the
licensee and will be tracked as part of Inspector Followup Item
50-260/87-30-05.
In reviewing 2-SI-4.4. A.2., the inspector noted that this instruction may
not fully cQmply with Technical Specification requirements. Technical
Specification 4.4. A.2.b requires that at least once per operating cycle
the Standbly Liquid Control System be manually initiated except for the
explosive valves. Boron solution is to be pumped through the
recirculation path and back to the Standby Liquid Control Solution Tank,
and a minimum pump flow rate of 39 gpm against a system head of 1,275 psig
is to be verified. The SI, as written, measures flow at the required
discharge pressure with suction being taken from the Standby Liquid
Control Test Tank and flow discharged via a vent line. The hydraulic
circuit used for the flow determination appears to differ from that
specified in TS 4.4.A.2.b. This concern was discussed with the licensee
and will be tracked as part of Inspector Followup Item 50-260/87-30-05.
During the performance of Surveillance Instruction SI 4.8.B.2-2b,
Particulate Filter Monthly Alpha Composite, the licensee determined that
the sample was inadvertently discarded following analysis on August 26,
1987. This monthly analysis is performed by compositing samples taken
during weekly collections throughout the month. Once the monthly analysis
is completed, the sample should have been retained for the purpose of
preparing a quarterly composite. It is this quarterly composite which is
required by Technical Specification Table 4.8-B, quarterly scan for
Strontium-89 and Strontium-90, which cannot be completed due to the
discarded sample. This is the second occurrence of this event within
about a years time frame. On June 27, 1986, the sample was improperly
discarded by a trainee. That event was attributed to a procedural error
since the procedure did not require saving the monthly sample for use in
the quarterly composite and inexperience of the trainee. The procedure
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was revised as part of the corrective action for that event. The recent i
event has been preliminarily assessed as a communication breakdown during
shift turnover. The prior event was not reported as an LER due to an
interpretation that although a Technical Specification surveillance
requirement was missed, no LCO had been violated. This item will be
tracked as an Inspector Followup Item (259,260,296/87-3,0-01) to review the .
corrective actions resulting from the recent event and to resolve the
deportability questions related to both events.
8. Reportable Occurrences (90712,92700)
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The below listed licensee events reports (LERs) were reviewed to determine
! if the information provided met NRC requirements. The determination
included: adequacy of event description, verification of compliance with
technical specifications and regulatory requirements, corrective action
taken, existence of potential generic problems, reporting requirements
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satisfied, and the relative safety significance of each event. Additional
in plant reviews and discussion with plant personnel, as appropriate, were
conducted for those reports indicated by an asterisk. The following
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licensee event reports are closed:
LER No. Date Event
- 259/87-014 8/4/87 Control Room Emergency
Ventilation System Flow Rate
Too High
- 259/87-004 3/20/87 Inadvertent Initiation of
Control Room Emergency
Ventilation
- 260/87-003 5/29/87 Reactor Vessel Water
Chemistry Excursion
- 259/85-013 5/14/85 Temporary startup test
panel installation
LER 87-14 was reported under the provisions of 10 CFR 50.73(a)(2)(1) for
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deviation from Technical Specifications and 10 CFR 50.73 (a)(2)(v) for a
condition that alone could have prevented the fulfillment of the Control
Room Emergency Ventilation System (CREV) safety function. The LER did not
address how the deficient condition (throttling damper incorrectly
adjusted due to inappropriate flow test method) could have prevented the
safety function. Neither did the LER contain an adequate discussion of
the safety consequences which could have resulted from this failure and
the potential compensation from other available systems and programs. The
LER also did not contain sufficient dates and times of events such that
the duration that the equipment was inoperable could be determined.
The LER failed to address that tighter administrative controls have been
imposed upon ventilation system components and alignments in the control
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room in order to prevent recurrence of this event. The LER should address
the magnitude of the reduction in filter efficiency resulting from the
increased flow and assess the corresponding increase in post-accident
radiation dose for control room and technical support center (TSC) per-
sonnel. Further a discussion should also be provided on other system
availability to compensate for CREV system inoperability or degradation
under postulated accident conditions. This may include the licensee's
post-accident dose assessment and monitoring program for operating " cc q['
personnel in order to determine whether undetected overexposure could
occur. Since this is a licensee identified violation, all of this infor- a
mation is necessary in order to properly assess the safety significance
and consequence of this event. This will be carried as an Unresolved Item
pending a thorough evaluation by the licensee (259,260,296/87-30-02). - - -
The cause of the inadvertent initiation of Control Room Emergency
Ventilation (LER 259/87-004) was the moving of a control bay air inlet
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radiation monitor by an Assistant Shift Engineer. The ASE was counseled
for not paying closer attention to his work. The equipment functioned as
designed.
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The water chemistry excursion (LER 260/87-003) is believed to have been
caused by atmospheric nitrogen combining with reactor water in the
presence of radiation energy with the gamma flux of the reactor providing
energy to drive the reaction. An evaluation was performed by the licensee
which determined this event could be minimized by use of a feed and bleed
process while filling the vessel and placing RWCU in service. Future
procedures will attempt to use these alternatives.
Problems were identified by Engineering regarding the existing
configuration of the startup test instrumentation panels (LER 259/85-013).
Corrective actions taken by the licensee included a modification of the
panel installation, and a design change request.
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9. Restart Test Program
The Restart Test Program (RTP) is intended to confirm the safe and
reliable operation of systems involved in the operation of Browns Ferry
Unit 2. All systems are addressed using a graded approach, i.e. , some
systems will receive a more extensive review and test than others. The
graded approach is implemented by separating all systems into one of three
groups according to the following general guidelines:
Group 1: Systems critical to safe operation or shutdown of plant will be
included in this group. Testing requirements are determined
primarily by Design Baseline Evaluation Program.
Group 2: Systems which provide support to plant operation are categorized
as Group 2. Few test requirements specified by Design Baseline
Evaluation Program; the majority of test requirements are
determined by the RTP system review.
Group 3: Systems not directly supporting plant operation and not
important to safety. Generally, no testing will be required.
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Groups 1 and 2 will have a System. Test Specification (STS) prepared per
Site Director Standard Practice (SDSP) 12.2, and they will have a RTP Test
Instruction and Test Results Package prepared as described in SDSP 12.1.
Group 3 systems will not have a STS prepared but will be addressed by the
requirements of the System Checklist, which is part of the Test Results
Package as required for Groups 1 and 2.
The RTP includes integrated system tests. The procedures for the
integrated tests are written and conducted per the requirements of SDSP
12.1. Test procedures and results are reviewed by the Joint Test Group
(JTG) as specified in Browns Ferry Standard Practice BF-1.10, Plant -
Operations Review Committee (PORC).
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The JTG is a group of site personnel acting as a Plant Operations Review
Committee (PORC) subcommittee with authority to review STS's, RTP tests,
and revisions to RTP Tests and STS's as described in Standard Practice
BF-1.10.
System Test Specifications (STS's) are used to stipulate the test
requirements for selected systems and other interrelated systems for the
BFN Unit 2 Startup. System Test Specifications are developed from
elements of the Design Baseline Evaluation Program Test Requirements and
from a system review by the RTP Engineer.
The STS is the document that defines the functions of a system that must
be tested, and the basis for the requirement. These test requirements are
implemented by test instructions as described by SDSP-12.1, " Restart Test
Program".
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The following procedures were developed by TVA to provide direction in the
conduct of the Restart Test Program. The inspector reviewed the
procedures to assure that program control and implementation adequately
met the commitments in Volume III of the Performance Plan.
(a) Section Instruction Letter (SIL) - 002, Training and Qualification of
Restart Test Program Personnel
(b) SIL-005- System Punch List Program
(c) SIL-006- System Check List Preparation
(d) SIL-001- Preparation and Use of Division of Nuclear Engineering
"DNE Need Sheets" -
(e) SIL-007- Review Documentation Reports
(f) SIL-003- RTP Instruction EXAMPLE FORMATS
The inspector attended a JTG meeting on August 27, 1987. The meeting was
conducted as prescribed by SDSP 12.1,12.2, BF 1.10, and as discussed in
the Performance Plan Volume III, Section 8.0. Agenda items discussed
included:
(a) Review and approval of JTG meeting minutes87-029 and 87-030.
(b) Non-intent changes for:
(1) 2 BFN-STS-073, High Pressure Coolant Injection System
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' (2) 2 BFN-RTP-027, Condenser Circulating Water System
(3) 2 BFN-STS-090, Radiation Monitoring System
(4) 2 BFN-RTP-025, Raw Service Water System
(5) 2 BFN-RTP-032, Control Air /Drywell Control Air System 4
(c) Review and approval of STS-099, Reactor Protection System.
The following conclusions were reached by the inspector during the review: 3
(a) Test control and conduct is closely coordinated between the restart
test group and operations staff to ensure Technical Specification
requirements are adhered to during testing.
(b) The Restart Test Program described in the Performance Plan Volume III
was fully implemented and controlled by site director and plant
procedures.
(c) Test directors are fully trained and certified to conduct system
tests. The inspectors consider the test directors knowledge and
test conduct as a positive attribute of this program.
(d) The JTG reviews and approves program procedures, and the PORC and
Plant Manager reviews and approves the procedures which detail the ;
program scope, as well as the specific system special test
procedures.
On August 20,' 1987, inspectors witnessed portions of testing performed per
SI-4.9.A.4.b, "4-KV Shutdown Board Undervoltage Start of Diesel
Generator". More specifically, they witnessed the undervoltage on 4-KV
shutdown board A and start of diesel generator A performed per Section 4.0
of this proceaure. This testing was performed as part of the restart test
program in support of both 2-BFN-RTP-57-5 and 2-BFN-RTP-082. The
inspectors observed that testing was being performed in accordance with
approved procedures and that the procedures were being followed by test ,
personnel. ]
During the performance of Restart Test Procedure 2-BFN-RTP-032 an
unplanned Engineered Safety Feature (ESF) actuation occurred; the start of
Standby Gas Treatment trains A and C, and the start of Control Room
Emergency Ventilation train A. The testing included the verification of
closure of valve 2-FCV-32-62 in response to a Primary Containment
Isolation Signal . Jumpers were being installed to prevent inadvertent
actuation of Primary Containment Isolation System controlled components
which were controlled by the same relay, 16A-K37, as valve 2-FCV-32-62.
During the jumper installation process, the associated coil of the relay
contacts being jumpered, 16A-K37, was electrically shorted. This caused
de-energization of the relay and subsequent actuation of the ESF's
mentioned. The Unit 1 Unit Operator received the "Rx Bldg Vent Abnormal" i
alarm in the Main Control Room. He then determined that CREV A, SBGT A, I
and the Refuel Zone Ventilation had isolated and associated fans had
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tripped. He placed the fans in the off position. No high radiation
signal or alarm was present and no reason for continued CREV or SBGT
operation was in evidence. The Unit 1 Unit Operator then called the Unit
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2 Unit Operator and informed him of.the SBGT A and CREV A start. The Unit
2 Unit Operator informed the Unit 1 Unit Operator that the probable cause
of initiation was test procedure 2-BFN-RTP-032. Upon the concurrence of *
the Shift Engineer, the Unit 2 Operator secured SBGT C after resetting the
PCIS signal. Likewise, upon concurrence of the Shift Engineer, the Unit 1
Unit Operator reset the PCIS signal, secured SBGT A and CREV A, and
restarted the Refuel Zone Fans. Operations and test personnel actions
were deemed adequate and proper by the inspector. Recurrence prevention
i control will be evaluated by the licensee.
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10. Employee Concerns Program
The inspector performed a review of the Employee Concerns Programs at
Browns Ferry. The inspection included a general overview of the program
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- as well as a review of several case files. I
_
The overall program is described in the Office of Nuclear Power (ONP) TVA
Employee Concern Program Instructions Manual. This manual considers of a
i line organization procedure; which describes the purpose, scope and policy
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of the ONP Employee Concern Program as well as a basic description of how ,
the program should work; and a set of ECP Instructions which describe
implementation of the program. ECP Instruction 1 (Site Representative
Procedure) states that to implement the policy of the ECP, "an ECP shall
be established ... to include ... methods to receive, classify,
investigate, and refer employee concerns to other responsible authorities
when appropriate to ensure an effective program". However, investigation
by the inspector appears to indicate that except for the restart safety
significance category classifications and identification of potential for
referral to IG, no formal methods have oeen established. Rather, the
program appears to rely on the experience of the ECP staff members to make
these determinations.
Instruction 1 also requires methods to identify instances of abuse of the
ECP system (i.e. concerns more adequately handled through the
labor / management grievance process, repeated submittal of concerns on a
subject TVA has already taken a stand on, or concerns made in an effort to
harrass management). These methods also appear to be left up to the
discretion of the individual ECP staff member. The lack of formalized
methods to fully implement the program as described in the Employee
Concerns Program Instructions Manual was discussed with the ECP Site
Representative (ECP-SR). The ECP-SR felt that the selection criteria,
used to pick his staff, was rigorous enough to ensure his people are
qualified enough to make the decisions required to implement the program
without having formal written methods. The inspector has identified the
lack of formalized methods for implementation of ECP policy as an item of
concern. The inspector was also concerned regarding the ECP practice for
referral of Management and Personnel (MP) issues to line management,
especially in cases where the MP issue relates to harassment or
intimidation by the referred supervisor. Although the above inspector
concerns are not covered by regulatory requirements, the ECP is a TVA
commitment to the NRC, and warrants NRC management review to determine if
the current licensee's practice is acceptable. This item will be tracked
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as an Unresolved Item (259,260,296/87-30-03) pending further NRC review on
l the significance of the matter.
1
Instruction 1 also identifies two types of files kept by the ECP. They are
open files and concerns. Based on the inspector's discussion with the
ECP-SR and those examples provided by the manual, an open file appears to
be an item which is not significant enough to be classified as a concern.
The examples are: (1) an employee contacted ECP but is willing to work
l with line management to resolve the issue; (2) an anonymous issue where
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referral to line management is appropriate; (3) potentially generic item i
that was substantiated and not safety significant; (4) potentially generic 4
1
l item referred by another site representative; (5) exit interview by
employee (with no concerns) requiring followup by ECP; (6) employee
i request help from ECP but doesn't identify a concern; (7) information not
adequate to identify a concern; and (8) at discretion of ECP-SR employee
issues that are better handled cn an informal basis.
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The inspector reviewed some open files and some concern files to determine i'
the difference between the two types of files. Those files reviewed were
as follows:
File Number File Type
ECP-87-BF-489 OPEN
ECP-87-BF-094-01 CONJERN
ECP-87-QF-952 OPEN
ECP-87-BF-305 OPEN
ECP-87-BF-313 OPEN
ECP-86-BF-819 OPEN
ECP-86-BF-567 CONCERN
ECP-87-BF-588 OPEN
ECP-87-BF-413-01 OPEN
ECP-87-BF-389 OPEN
ECP-86-BF-889 OPEN
ECP-86-BF-868 OPEN
ECP-87-BF-893 OPEN
ECP-86-BF-199-01 CONCERN ,
ECP-86-BF-621 CONCERN
ECP-86-BF-620 CONCERN
ECP-86-BF-A55 CONCERN
The documentation in the concern files is adequate to support resolution, '
where as the documentation in the open files was found in some cases to be
insufficient to support resolution (ECP-87-BF-489, ECP-87-BF-305). This i
discrepancy was discussed with the ECP-SR. Other problems identified in l
the open files were lack of investigation into more technical issues
(ECP-87-BF-489, ECP-87-BF-588) and referral of MP issues to line
management for resolution (ECP-86-BF-868).
11. NSRS/SWEC Reports I
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13
The inspector reviewed the status. of the Nuclear Safety Review Staff
l (NSRS) and the Stone and Webster Engineering Corporation (SWEC) reports at
Browns Ferry.
The 24 NSRS open items identified as requiring closure by ECTG before Unit
2 start up have been incorporated into NSRS-BFN-01 and the corrective
actions required to close these 24 items are being tracked on a computer
system by the Employee Concerns Task Group (ECTG) staff who are also
responsible for maintaining the NSRS files.
The 85 items identified as not being required before restart have been
incorporated into NSRS-BFN-02. The ECTG staff is currently in the process
of adding the CATDs (Corrective Action Tracking Documents) related to this
report in the computer system. NSRS items are also tracked by the Plant
Operations and Review Staff (PORS) on the Plant Manager's Action Tracking
System. Information on NSRS reports is transmitted from ETCG to PORS on a
monthly basis.
SWEC reports are also maintained by the ECTG staff. Currently, the 64
SWEC element reports have been incorporated into a single SWEC subcategory
report for Browns Ferry. As with the NSRS reports the corrective actions
required to address the SWEC items will be tracked on the ECTG computer
system. The 14 SWEC items identified as not having sufficient evidence
for corrective action have been assigned CATD's but as of this inpsection
none of those responses was available for review. The status of the
CATD's for SWEC and NSRS reports will be reviewed in future inspections.
12. Restart Review Board
The inspector observed the Restart Review Board (RRB) meeting conducted on
August 27, 1987. Items for review were selected from Division of Nuclear
Engineering (DNE) conditions adverse to quality (CAQR) reports. The
inspector observed the RRB using both the TVA and NRC draf t proposed
restart criteria and referring to these criteria frequently. The RRB was
recording reasons for concluding that items were not required for restart.
The RRB was additionally cognizant of the commitment contained in the
Browns Ferry Nuclear Performance Plan to have the Manager of Nuclear Power
give explicit approval prior to reversing a prior decision to resolve an
item prior to restart. At least one example was observed where the Board
was to recommend reversal to the Manager of Nuclear Power.
The inspector observed several examples of a particularly difficult
problem to analyze using the restart criteria. The RRB was inconsistent
in its approach to this type of problem which involves potential problems
for problems for which the scope of the generic implication is not yet
fully determined. For example, a CAQR may identify a programmatic
breakdown of control where a potential exists for hardware deficiencies
but yet no sampling program has yet been carried out that located specific
examples. One philosphy espoused during the meeting was that a problem
did not exist until a specific deficiency was found and shown to be a
problem. Another competing philosphy was that until it can be shown that
no examples exist in the field, the potential problem must be addressed as
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t. 14
if the lax controls had indeed led to material deficiencies in the plant.
The inspector made known his opinion that inconsistencies in designating
this type of item as "in progress" or "YES" (meaning prior-to-restart) or
" BLANK" (meaning it will be addressed at a later time) were of no
consequence as long as one of these types of problems doesn't get
classified "N0" in which case the complete ramifications and consequences
would not be investigated until after restart. Of the three examples of
this category of problem witnessed by the inspector, some discussion was
held in this general direction; however, none were finally designated as
not required for restart. These examples were:
(1) Potential welding deficiencies in main steam relief valve (MSRV)
tailpipe restraints. There may be insufficient weld metal to carry
the required load. This condition has been evaluated as generic to
various other welds in the drywell.
(2) Some errors have been found in calculations used in the torus
attached piping program. Examples may be generic, yet it is not
known whether ' the errors affect the final acceptability of the
asssociated hardware.
(3) An obscure classification of non-safety-related cables were allowed
to be routed along with safety-related divisional cables. These
cables, were not uncontrolled during installation such that they may
have crossed other division's cables in violation of the divisional
separation criteria.
13. Facility Modifications (37701)
In the summer of 1985, the Unit 2 jet pump instrumentation safe ends were
replaced following indication of Intergranular Stress Corrosion Cracking
(IGSCC) on two welds. This activity was controlled and documented under
Engineering Change Nctice (ECN) P 0720 and Workplan 2069. Although, the 1
work is complete, the ECN and Workplan remain open pending satisfactory
inspection of the welds during the reactor vessel hydrostatic test.
Recent problems in backfilling the jet pump instrument lines have called
into question the modification activity associated with this safe end. i
Three jet pump sensing lines have been found to be plugged to the extent i
that no water flows through the lines at a pressure up to 700 psig. The
workplan did not call for any flushes or filling operations following safe ]
i
end replacement. Troubleshooting activities by the licensee has narrowed l
speculation about the probable cause to the soluble rice paper purge dams )
placed in the one quarter inch sensing line near the welds to the safe l
end.
14. FSAR Annual Update
s
The licensee recently submitted the annual FSAR update as required by 10 I
CFR 50.71(e). This update (Amendment 5) was reviewed along with the
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licensee's commitment made in response to Unresolved Item 86-05-01. The l
Unresolved Item related to indiscriminate changes to commitments contained l
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in the FSAR without documented justification. In a letter to the NRC
dated May 20, 1986, the licensee stated that future changes to the FSAR ,
would be formally evaluated, justified and available for review. Standard l
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15
Practice 1.13, Final Safety Analysis Report and Tecnnical Specifications,
controls this process and requires that each . change to the FSAR must be
addressed in a USQD. The inspector selected eight specific changes made
in Amendment 5 for review. Two of the changes were said to be covered by ,
a blanket USQD for administrative changes to correct typographical errers)
omissions, clarifications and miscellaneous changes of a "non-intent"
nature. This was inappropriate for these two instances.
The first change deleted a statement in Section 2.2, Site Description.
This paragraph previously contained a commitment to periodically examine
this portion of the FSAR to maintain a reasonable representation of the
area population and land use. No justification existed for deletion of
-
this commitment nor was it clear whether the licensee was eliminating the
periodic updates to this portion of the FSAR or whether they were -no
longer going to review the surrounding area for changes in population and
land use which might significantly exceed the original projections.
A second change which was considered ar, administrative "non-intent" type
of change was made to Section 13.3.1, Training Program Description. The
previous version of the FSAR contained a statement that training visits to
operating nuclear plants and other nuclear facilities have been and will
continue to be scheduled as needed for selected supervisors. This sentence
was deleted by Amendment 5. No evaluation could be kcated that assessed
l the intent of this original commitment and whether' the corrent training
program adequately compensates for the loss of dross-beveding, fresh
ideas, and evaluation of different ways of doing,4,hings.
One additional change was located that lacked a specific USQD. Section
2.6.2.1, atomspheric monitoring, was revised to delete a sentence which
described the capabilities of the local and perimeter environmental
monitors. The previous version stated that heavy particle fallout and
rainwater are also collected at these locations. This statement was
deleted by Amendment 5. It is unclear whether this change resulted from a
modification or planned modification to the equipment or whether this type
of sampling would simply be discontinued with the capability to be left
intact. ,
Failure to have performed and make available for review the formal
evaluation and justification of these changes to the FSAR is a deviation
from the commitment made by the licensee in its May 20, 1986 letter
(259,260,296/87-30-04).
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