ML18024A977
| ML18024A977 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 06/25/1979 |
| From: | Moon B, Ruhlman W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18024A973 | List: |
| References | |
| 50-259-79-14, 50-260-79-14, 50-296-79-14, NUDOCS 7908270209 | |
| Download: ML18024A977 (10) | |
See also: IR 05000259/1979014
Text
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UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTAST., N.W., SUITE 3100
ATLANTA,GEORGIA 30303
Report Nos. 50-259/79-14,
50-339/79-14
and 50-260/79-14
Licensee:
Valley Authority
818 Power Building
Chattanooga,
37401
Facility Name:
Browns Ferry Units 1,
2 and
3
Docket Nos. 50-259,
50-339,
and 50-260
License
Nos.
and DPR-68
Inspection at Browns Ferry Site near Athens,
Inspector:
B.
.
oon
.,.d.,IW A
W. A. Ruhlman, Acting Section Chief,
RONSB
Date Signed
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Da
Signed
sumev
Inspection
on Hay 21-25,
1979
Areas Inspected
This routine,
unannounced
inspection involved 46 inspector-hours
on-site in the
areas of plant procedures
for Units 1,
2 and
3 and
a review of valve/breaker/
switch alignments for all accessible
components
in all ESF systems
for Unit 3 to
verify the licensee
actions
taken in response
Results
Of the two areas
inspected,
no apparent
items of noncompliance
or deviations
were identified in one area;
one apparent
item of noncompliance
was found in one
area {infraction-failure to follow procedures,
paragraph
6),
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DETAILS
1.
Persons
Contacted
Licensee
Employees
- J.
G. Dewease,
Plant Superintendent
-R. Cole,
OP/A Resident
="J. L. Harvess,
Staff QA Supervisor,
BFNP
-R. Hunkpillar, Assistant Operation Supervisor
Other licensee
employees
contacted
included four operators
and
one office
personnel.
NRC Resident
Inspector
+R. F. Sullivan
-Attended exit interview.
2.
Exit Interview
The inspection
scope
and findings were summarized
on May 25,
1979 with
those
persons
indicated in Paragraph
1 above.
During the inspection,
seven
items were identified which require further action by the licensee
manage-
ment for acceptable
resolution.
These
items are contained in paragraphs
5.(l), 5.(2), 5.(3), 6., 6.(l), 6.(2),
and 6.(3).
The licensee
acknowledged
the inspectors'omments
on these
items.
3.
Licensee Action on Previous
Inspection Findings
Not inspected.
4.
Unresolved
Items
Unresolved
items are matters
about which more information is required to
determine
whether they are acceptable
or may involve noncompliance
or
deviations.
New unresolved
items identified during thjs inspection are
discussed
in paragraph 5.(3).
5.
Unit 1,
2 and
3 Procedures
The inspector
conducted
a review of procedures
and documentations
as follows:
EOI-1, "Recirculation
Pump Trip"
EOI-5, "Electrical System Failure"
EOI-8, "Reactor Water Level High/Low"
EOI-36, "Loss of Coolant Accident"
EOI-44, "Suppression
Chamber Water High Temperature"
NOI-65, "Standby Gas-Treatment
System"
NOI-68, "Reactor Water Recirculation System"
NOI-71, "Reactor
Core Isolation Cooling System"
NOI-73, "High-Pressure
Coolant Injection System"
NOI-74,
"RHR System"
NOI-91, "Pressure
Suppression
Pool Level Control"
NOI-92, "Neutron Monitoring System"
GOI-100-1, "Integrated Plant Operation"
GOI-100-5, "Process
Computer"
SI-2, "Instrument Checks
and Observations"
Technical Specification
Amendment Nos.
9,
11,
12,
13,
14,
16,
17,
18,
19,
31, 32, 34, 35, 36, 37, 38, 39, 40, 41, 42, 43, 44, 45, 46,
and
48
TVA Operational
equality Assurance
Manual
The above procedures
were reviewed to verify that:
Reviews,
approvals
and changes
covering the activities were in accor-
dance with Technical Specifications
Procedure
changes
were
made to reflect changes
required
by selected
Technical Specification revisions
Procedure
changes
were in conformance with 10 CFR 50.59(a)
and (b) re-
quirements.
Procedure
contents
were in accordance
with Technical Specification
and
applicable
standards
Selected
procedure
contents
were adequate
to control required safety
related operations
Controls were established
to prevent
the freezing of functional fluid
systems
during maintenance
The inspector
used
one or more of the following acceptance
criteria for
evaluating the above
items in the procedure
review:
Technical Specifications
Topical Report TVA-TR75-1A/TVA equality Assurance
Manual
FSAR, Section 13.6,
Appendix D and Response
to AEC questions
dated
March 25,
1971
(BRNP 24)
(1976)
Regulatory
Guide 1.33 (1978,
Rev.
2)
10 CFR 50, Appendix
B
Within the areas
inspected
no items of noncompliance
or deviations
were
identified.
The following items are
open pending corrective action by the licensee
and
verification by NRC at subsequent
inspections:
Technical Specification 3.7.f requires that reactor shall be scramed
when suppression
chamber water temperature
reaches
110 degrees
Farenheit.
(2).
This scram requirement is not currently incorporated in the procedure
EOI44 "Suppression
Chamber
Water High Temp".
Th'is appears
to be
a
typographical error made at the time of the procedure
revision (2/28/78).
The licensee
committed to incorporate
the requirement in the procedure
by June
25,
1979
(259/260/296/79-14-01).
The LPCI loop selection logics are
no longer functional for Unit
1 and
2 as
a result of current design modification.
The procedure
EOI8
"Reactor Water Level High/Low" still lists this action
as
an automatic
action to verify.
The licensee
committed to revise the procedure
by June
26,
1979
(259/260/79-14-02).
(3)
Technical Specification 6.3.A.l and
4 require that detailed written
procedures
covering normal operation
and emergency
condition shall
be
prepared.
Browns Ferry is committed, by FSAR Responses
to AEC (}ues-
tion 13.11 dated
March 25,
1971, to have
an abnormal
and emergency
operating procedure titled "Reactor Water Level High/Low".
The Pro-
cedure
EOI 8 "Reactor Water Level High/Low" does not contain appro-
priate procedures
for LO-LO-LO water level setpoint.
The licensee
stated that they will review to determine if the LO-LO-LO
level situation is covered
somewhere
else.
This item will remain
unresolved
(259/260/296/79-14-03).
-4-
6.
Onsite Inspection of Engineered
Safety Features
(ESF) - Unit 3 Only
The inspector
conducted
an independent
examination
on actions
taken by the
licensee in response
to IE Bulletin 79-08 for the following ESF systems:
- High Pressure
Coolant Injection System
(HPCI)
- Reactor
Core Isolation Cooling System
(RCIC)
- Core Spray System
(CS)
(LPCI)
- Automatic Depressurization
System
(ADS)
- Standby Liquid Control System
(SLCS)
- Emergency
Equipment Cooling Water System
(EECW)
- RHR Service Mater System
(RHRSW)
- Diesel Generator
System
(DG)
The inspector
reviewed the following licensee
procedures
and drawings to
verify safety valve/breaker/switch
alignments
to ensure
the
ESF systems
are
and that licensee's
procedures
and administrative controls provide
adequate
assurance
of continued operability.
Control Room Panel Checklists for HPCI, RCIC, LPCI,
CS,
SBLCS,
ADS and
EECW per OI-73, -71, -74, -75, -63, -1 and -67
RHR System
"RHR Systems"
Browns Ferry Flow Diagram Drawings
(Nos.
and 47W812-1)
The inspector also verified the standby
readiness
of four diesels
by
checking control room indications
and local panel switch alignment
as well
as the breaker positions.
The inspector
used
one or more of the following acceptance
criteria for the
above items:
- Technical Specifications,
Section
6
- Topical Report TVA-TR75-lA/TVAQuality Assurance
Manual
- Operating Instruction 74,
"RHR Systems"
- Browns Ferry Drawing No. 47W811-1,
"Flow Diagram
RHR Systems"
- Browns Ferry Drawing No. 47W859-1, "Flor Diagram
EECW"
- 10 CFR 50, Appendix
B
The above
systems
were inspected
to verify that:
Valve/breaker/switch
alignments
were aligned in accordance
with plant
procedures
and drawings
The alignment procedures
are adequate.
0
-5"
Comparison of alignment procedures
for the above
systems with current
PAID's and single-line diagrams
Standby
readiness
of Diesel Generator
System
was established
Within the areas
inspected,
one apparent
item of noncompliance
was iden-
tified.
The inspector identified that two outlet valves
(HCV 74-33 and
74-44) at the discharge
side of RHR Heat Exchangers
3B and 3D, were not
locked nor controlled during the Unit 3 operation
on May 24,
1972.
This is contrary to the requirements
in Technical Specification 6.3.A.1 and
operating Instruction
74
"RHR Systems"
which state in part that "Detailed
written procedures
including applicable checkoff lists covering normal and
shutdown operation shall be prepared,
approved
and adhered to" and that the
valves shall be "locked open" respectively
Another example of ESF valves not locked during the operation
were the
throttle valves
Nos.
561,
572 and
614 which are located at the
EECW flow
path for RHR pump Seal Heat Exchangers
3B and
3D.
These
examples
are designated
as
an item of noncompliance
(296/79-14-04).
The licensee
acknowledged
the inspector's
findings immediately and instruc-
ted the operator to lock the
NCV 74-33 and 74-44 during the inspection.
The inspector stated that the
EECW throttle valves
must be re-examined
by
the licensee
and locked after the flow is adjusted.
The following items remain open pending corrective action by the licensee
and verification by NRC at subsequent
inspections:
(1).
The
EECW RHR Pump Seal Heat Exchanger
Valves
(Nos. 561,
572 and 614)
on Drawing No.
47W859-1 "Flow Diagram Emergency
Equipment Cooling
Water" are required to be locked while Procedure
OI 74
"RHR Systems"
does not require locking.
This appears
to be
an inconsistant
procedure
requirement
(296/79-14-05).
(2). The
EECW RHR Pump Seal Heat Exchanger
Valves
605 and 616 are miss-
tagged
(interchanged)
with respect
to the
RHR pumps
3B and
3D and not
in accordance
with the Drawing No.
47W859-1 "Flow Diagram Emergency
Equipment Cooling Water".
The licensee
stated
they will correct the tags by June
24,
1979
(296/79-14-06).
(3). The RCIC System
Panel Checklist
(OI71) does not include FCV71-10
(Turbine Governing Valve) for position verification, although the
valve is in major flow path
as
shown
on Drawing No
~ 47W813-1,
"Flow
Diagram RCIC System".
This valve is normally open
and its position
must be verified with a Panel Checklist (296/79-14-07).
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