ML19309D921
| ML19309D921 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 07/10/1974 |
| From: | Kidd M, Robert Lewis NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19309D918 | List: |
| References | |
| 50-313-74-09, 50-313-74-9, NUDOCS 8004110746 | |
| Download: ML19309D921 (48) | |
See also: IR 05000313/1974009
Text
._
_
__
_
. . _ _
_
. _ _ _ _
.
...
'
s
UNITED STATES
[
l ,gt
4
.
ATOMIC ENERGY COMMISSION
,
fc B{. i
7
.s
DIiEC NHATE OF RSCOLATGW OPESATICE3
s
-
d D-
l
-
54El.e0 H II - Se a s f F_ 818
2 41 P h AC.H i n E E ST H t. E 1. NOR T u rt E.51
, , , ,
,,
AT L a ta r a. Gt.OHG e A M30 3
JUL 1 2 1974
RO Inspection Report No. 50-313/74-9
Licensee: Arkansas Power and Light Company
Sixth and Pine Streets
Pine Bluff, Arkansas
71601
i
Facility Name: Arkansas Nuclear One, Unit 1
Docket No.:
50-313
-
License No.:
CPPR-57
Category:
B1
,
i
Location: Russellville, Arkansas
Type of License: B&W, PWR, 2568 Mwt
'.W e of Inspection: Routine, Announced
Dates of Inspection: April 16-19, 24-26, May 1-3 and 13-17, 1974
Dates of Previous Inspection: March 26-29 and April 16-18, 1974
t
,
Principal Inspector:
M. S. Kidd, Reactor Inspector
4
Facilities Test and Startup Branch
Accompanying Inspectors:
D. J. Burke, Reactor Inspector
Facilities Test and Startup Branch
-
-
F. Jape, Reactor Inspector
Facilities Test and Startup Branch
,
_.
.
R. C. Parker, Reactor Inspector
Facilities Test and Startup Branch
W. W. Peery, Radiation Specialist
Radiological and Environmental Protection Branch
K. W. Whitt, Reactor Inspector
Facilities Test and Startup Branch
Other Accompanying Personnel: None
Principal Inspector:
b
/Id[Y[
.
M. S. Kidd,' Reactor Inspector
Date
Facilities Test and Startup Branch
Reviewed By: 8, C
7
M
Date -
s
R. C. Lewis, Senior Inspector
COM C C :
3
Facilities Test and Startup Branch
?DR
-
'
8
. .
..
_ .
. - _. .
0427o7y p
.
.
'
A
^
,
'
.,
'
RO Rpt. No. 50-313/74-9
-2-
'
SUMMARY OF FINDINGS
I.
Enforcement Matters
A.
Violations
Contrary to Section 13 of the FSAR, the control room "Halon"
fire protection system test was not conducted per a procedure
which had been reviewed and approved by AP&L.
(Details I,
paragraph 26.d)
.
B.
Safety Items
None
II.
Licensee Action on Previously Identified Enforcement Matters
'
A.
Violations
-
1.
RO Report Nos. 50-313/74-2 and 50-313/74-4
Licensee responses have been received on the apparent
s
violations discussed in Section I.A of the above reports.
These items are considered closed.
(Details I, paragraphs
3.a through 3.h)
2.
Calibration of Test Equipment (R0 Report No. 50-313/74-7)
.
Not inspected.
B.
Safety Items
None
III. New Unresolved Items
74-9/1 Steam Generator Hydrotest
The steam generator secondary side is to be rehydroed prior
to initial criticality.
(Details I, paragraph 2.b(1))
,
.
e
i
.
,
, -
,.
-
- , - - . .
-
a--,
- ,
.a~<,
-,
-
-v.,
- ---
-ne
,, - - -+ v
+---
9
,
..s
p ..
.
.
.
' 's _,/
-
s
RO Rpt. No. 50-313/74-9
-3-
74-9/2 Checkout of Reactor Coolant Pumps (RCP)
The RCP's are to be checked out again prior to initial
criticality.
(Details I, paragraph 2.b(.2))
'
74-9/3 Pressurizer Code Safety Valve
The seat of one pressurizer code safety valve was replaced.
The setpoint is to be recheched prior to initirl criticality.
(Details I, paragraph 2.b(3))
74-9/4 Service Water Isolation Valves
'
Air accumulators are to be installed on certain SW iso 2ation
valves prior to initial criticality.
(Details I, paragraph
2.b (4))
74-9/5 Main Steam Piping Restraints
.
Shims are to b'e installed on the main steam piping restraints
outside of the reactor building prior to initial criticality.
(s_
(Details I, paragraph 2.b(5))
74-9/6 Grating Over Steam Generators
Storage areas and tie-downs for the grating over the steam
generators must be provided prior to initial criticality.
(Details I, paragraph 2.b(6)
74-9/7 Quench Tank Rupture Disk Deflector
,
A deflector is to be installed over the quench tank rupture
disk prior to initial criticality.
(Details I, paragraph
2.b(7))
74-9/8 DC Panel Transfer Switches
The manual transfer switches for DC distribution panels
D11 and D12 are to be replaced prior to initial criticality.
(Details I, paragraph 2.b(8)
74-9/9 Preoperational Test Deficiencies
Fifteen preoperational tests contained deficiencies which
must be resolved prior to initical criticality.
(Details I,
-~s
paragraph 2.b(9))
,
)
ss_ -
.
I
\\
!
l
!
'
. ~
_
.
.
..
-
.
.
,
n-
,
.
O
RO Rpt. No. 50-313/74-9
-4-
.
IV.
Status of Previously Reported Unresolved Items
72-12/2 Valve Wall Thickness Verification Program
Not inspected. This item must be resolved prior to initial
criticality.
73-8/l Procedural Coverage Per Regulatory Guide 1.33
All procedures recommended by RG 1.33 for operations of
the facility have been written and approved with the except-
ion of two surveillance test procedures which will be
required for the first refueling.
This item is considered
'
closed.
(Details V, paragraph 2)
73-10/6 Respiratory Protection Program and Procedures
Training in the use of respirators and storage of them
has been completed. This item is resolved.
(Details I,
paragraph 4 and Details III, paragraph 2)
73-10/7 Representative Sampling of Gaseous Waste
,_,
Insulation of the gaseous stack vent sample lines has been
completed. This item is considered resolved.
(Details III,
paragraph 3)
,
,
. 73-12/2 Diesel Generator Trips
Modifications to the emergency diesel generators and
,
inverters and retesting have been completed. This item
is resolved.
(Details I, paragraph 5)
73-12/3 Control Rod Trip Test
At power trip tests will be conducted using an operational
test procedure, in which the inspectors comments have been
resolved. This item is closed.
(Details I, paragraph 6)
73-14/2 Initial Core Load Procedure
This procedure has been approved and previous comments
resolved. This item is closed.
(Details I, paragraph 7)
,
v
.
l
_
4
_
_
_.
l
l
'
'
m
'
.
,
s_s
.R0 Rpt. No. 50-313/74-9
5-
-
73-17/1 Pressurizer Electromatic Relief Valve
.
The inspector was informed that the design of this valve
is adequate for its intended service. This item is
considered closed.
(Details I, paragraph 8)
73-17/2 Emergency Operating Procedures
All of these procedures have been written and approved.
In
,
addition, RO comments have been resolved. This item is
l
closed.
(Details I, paragraph 9)
'
!
73-17/3 Operational Test Program
Schedules to be used in conducting surveillance testing
have been prepared and all procedures needed for initial
operations have been approved. This item is considered
. resolved.
(De, tails I, paragraph 10)
.
73-18/1 Emergency Planning
-
'
Implementation of the emergency plan has been completed;
therefore, this item is considered resolved.
(Details I,
paragraph 11 and Details III, paragraph 4)
73-18/3 Calibration of Radiation Monitors
The calibration and functional checkout of area and process
monitors have been completed. This item is closed.
(Details
'
I, paragraph 12 and Details III, paragraph 5)
73-19/2 Reactor Building Ventilation System Ductwork Design
A final report per 10 CFR 50.55(e) has been received and
,
all modifications to the ductwork completed. This ite.m
is considered resolved.
(Details I, paragraph 13)
74-2/1 Vibrations -in Reactor Vessel / Reactor Coolant Systens
The cause of the noises and vibrations observed during hot
functional testing has been determined to be a loose radiation
i
specimen holder which is attached to the core barrel. This
item is resolved.
(Details I, paragraph 14)
!
)
J
-
,
,
I.
4
.e--e-y.-
+r,--
_
.-..p,
,-.,_,.y-
_ - , ,
---m,,,-
_ -
%e-
- - _
,e.--w
--y-
- - - -- -
_
. _ ___ .__- . __
-
f
~
.
.
.
--
.~
s
'
'
'
4
.
R0 Rpc. No. 50-313/74-9
-6-
74-2/2 Control of Maintenance Activities
Controls for all types of maintenacce activities have been
redefined and strengthened were appropriate. This item is
resolved.
(Details I, paragraph 15)
74-2/3 Control of Temporary Circuit Modifications
Controls for the use of all types of jumpers and bypasses
have been strengthened. This item is considered resolved.
,
(Details I, paragraph 16)
74-4/1 Leak in P36B Recirculation Line Flow Orifice
The flow orifice has been redesigned and two of three
have been replaced.
This item remains open and will be
resolved prior to initial. criticality.
(Details I,
-
paragraph 17).
74-4/2 Control Rod Drive (CRD) Position Indicators
I
A final report in this item has been received which
concludes that the problems will not have an adverse effect
on the safe operation of the plant. This item is closed.
(Details I, paragraph 18)
.
74-5/1 Organization
,
.
This item is closed.
(Details IV)
.
74-5/2 Quality Assurance Program
This item is closed.
(Details IV)
74-5/3 Design control
Ihis item is closed.
(Details IV)
74-5/4 Procurement Document Control
This item is closed.
(Details IV)
~
74-5/5 Instructions, Procedures and Drawings
This item is closed.
(Details IV)
s.
-
s
4
-,-ee,,
w -
m-.
+
w-
< --
, , ,
%.
v---
- - ----
,-y-r
_
_
_
'
~ N
'
.
,
,.
O
-
RO Rpt. No. 50-313/74-9
-7-
74-5/6 Document Control
This item is closed.
(Details IV)
74-5/7 Control of Purchased Material, Equipment and Services
This item is closed.
(Details IV)
74-5/8 Identification-and Control of Materials, Parts and Components
This item is closed.
(Details IV)
,
4
74-5/9 Control of Special Processes
i
,
This item is closed.
(Details IV)
74-5/10 Inspection
.
This item is closed.
(Details IV)
i
74-5/11 Test Control
-
This item is clos.ed.
(Details IV)
74-5/12 control of Measuring and Test Equipment
This item is closed.
(Details IV)
.
74-5/13 Handling, Storage and Shipping
This item is closed.
(Details IV)
74-5/14 Inspection, Test and Operating Status
j
This item is closed.
(Details IV)
74-5/15 Nonconforming Materials, Parts or Components
!
This item is closed.
(Details IV)
74-5/16 Corrective Action
This-item is closed.
(. Details IV)
,
~
-
+
+
- - - .
-
--
,----,,--,v.--
y--e,
,
w
ye+-----
e
- - =
---m--v-v
---
%or
. - -
-v-w--
-
- -- . _ - .
_
.
.
5
'
7-g-
.
,
,
/
'
R0 Rpt. No. 50-313/74-9
-8-
.
74-5/17 Quality Assurance Records
This item is closed.
(Details IV)
74-5/18 Audits
This item is closed.
(Details IV)
74-6/1 Replacement of Quality Breakers With Non-Quality Documented
Ones
A report per 10 CFR 50.55(e) has been received and proper
,
breakers have been installed. This item is closed.
(Details I,
paragraph 19)
74-6/2 Sensitivity of Plant Leak Detection Systems
.
The licensee has agreed to revise the FSAR to better define
the capabilities of the inventory balance leak detection
-
method. This item remains open.
(Details I, paragraph 20)
r(
74-6/3 Availability of Zero Power Physics and Powe'r Ascension Test
Procedures
All of these procedures have been reviewed and approved by
AP&L. This item is resolved.
(Details I, paragraph 21)
74-6/4 Quality Documentation of Valves
1
The eight small valves in the main steam system were replaced
- l
because they were not
"N" stamped, not because of a lack of
documentation. This item is considered closed.
(Details I,
paragraph 22)
74-7/1 Preoperational (Baseline) Inspection Data
Problem areas in this data are to be addressed in the Baseline
Inspection Report to be submitted per technical specifications.
AP&L has committed to resolving these problems prior to
initial criticality.. This item remains open.
(Details I,
paragraph 23)
4
.
e
.
.
-
.-
-
--
.
_
__
~
.
-
q
,
.
I
RO Rpt. No. 50-313/74-9
-.-9-
-
74-7/2 Mechanized Ultrasonic Data Verification and Repeatability
Not inspected. This item is to be resolved upon submission
of the Baseline Inspection Report.
74-7/3 Regulatory Operations Bulletin No. 74-3
A response to this ROB has been received. This item is
considered closed.
(Details I, paragraph 24)
V.
Design Changes
,
A.
Reactor Building Ventilation System Ductwork
This ductwork has been strengthened by adding stiffeners and
hangers and by beefing up the welding between existing stiffeners
and hangers and the ductwork.
(Details I, paragraph 13)
.
B.
Service Water Pumps
Restart logics for the service water pump moto'rs have been
modified such that they will automatically restart following
,
s_,
a loss of offsite power as well as upon receipt of an
engineered safeguards- (ES) signal.
(Details I, paragraph 28)
VI.
Unusual Occurrences
.
A.
Service Water Pump Motor Housing Cracks
On April 24, 1974, AP&L reported a significant deficiency
regarding cracks found in a structural member of service water
pump motor PM4A lower motor base (end bell). The cracking was
apparently due to rough handling during realignment.
(Details I,
paragraph 27)
B.
Paralleling of ES Buses A3 and A4
,
During the conduct of the Integrated ES Test (TP310.03) on May 11,
1974, service water pump motor PM4B vas fed power simultaneously
,
from redundant vital buses A3 and A4, resulting in a paralleling
of these buses. A drawing error associated with the design
change in V.B. above was the cause of this problem.
(Details I,
paragraph 28)
,
1
,~
\\
.
!
.-. .
. , , - . .
-.
',
- ,_ ,
, _ _ . _ . , _ _ -
-
. .
. .- .
. - . -
- - -
. . -
- ~-
-
_
'
3
s
p-
.
.
. s~_ >
R0 Rpt. No. 50-313/74-9
-10-
C.
Inability to Insert Incore Monitor Detectors
On May 10, 1974, it was discovered that certain of the incore
detectors could not be inserted into their guide tubes due to
penetration of shop welds through the pipe walls. The excessive
'
weld penetrations were reamed out to provide sufficient
clearance for the detectors.
(Details I, paragraph 2.a(1))
VII. Other Significant Findings
'
A.
Review of Test Results
,
RO Review of a sampling of test results packages was completed.
(Details I, paragraph 26; Details II, paragraphs 3-10; Details V,
paragraph 4)
B.
Annunciator Response Procedures
.
A review of a sampling of annunciator response procedures was
completed with no major comments outstanding.
(Details V,
paragraph 3)
x
C.
Radiological Waste Disposal Systems,
,
The review of radiological waste disposal systems and comparison
of them with FSAR descriptions was completed.
(Details III,
paragraph 6)
.
,
D.
Project Status
This prelicensing inspection was concluded onsite May 17, 1974,
after commitments were obtained on those items to be completed
prior to licensing and also prior to initial criticality. The
'
completion of those items required prior to licensing was con-
firmed by telecon to the inspector May 21, 1974, and Operating
License DPR-51 was issued for Unit 1 on that date.
(Details I,
paragraphs 2, 17 and 23)
I
E.
Welding Review
l
On May 16-17, 1974, a review of welding for Units 1 and 2 was
made to determine if the peening of a root pass of a piping
'
weld in Unit 2 (R0 Report No. 50-368/74-2, Details IV) was an
.
isolated incident and to determine if the same welder had
welded nuclear class piping in Unit 1.
The results are
documented in report No. 50-368/74-3. Evidence indicated
that the welder had not performed welding on safety systems in
\\s_
Unit 1.
~
1
..
- _ _ -
.
___
_______
1
-
'
'
\\
~
,
,
\\
x -l
'
R0 Rpt. No. 50-313/74-9
-11-
' l
YIIL Management Interview
Management interviews were held with J. W. Anderson, Plant Superintendent,
and members of his staff after each segment of the inspection to
discuss findings of the inspection.
Interim interviews were conducted
April 19 (Details I), April 24 (Details IV), April 26 (Details I),
j
May 2 (Details III), and May 3, 1974, (Details I and III). A final
interview was conducted May 17, 1974, with Mr. Anderson and
W. Cavanaugh, III, Manager of Nuclear Services, to discuss all findings
of the prelicensing inspection.
(Details I and V)
!
'
The apparent violation in Section I was discussed.
(Details I,
paragraph 26.d)
The previously identified enforcement. matters in Section II were
discussed.
(Details I, paragraph 3)
f
The new unresolved items in Section III were discussed.
(Details I,
paragrapr 2)
The status of previously identified unresolved items was discussed.
'
[\\ss)
(Details I, paragraphs 4-24, and Details III, paragraphs 2-5)
.
)
9
l
\\
.
N__/
-
1
.
-w-
, , - , - - - -
, - _ - - -
,-me
m-
_
- - _ .
-
-
M
'
'
'3
.
.
.
V
Ro Rpt. No. 50-313/74-9
I-l
-
DETAILS I
Prepared By:
[Y8-2d
M. S. Kidd, Reactor Inspector
Date
Facilities Test and Startup Branch
.
Dates of Inspection: April 16-19, April 24-26, May 1-3,
and May 13-17, 1974
Reviewed By:
84.
M
f//o/7%
R. C. Lewis, Acting Chief
Date
Facilities Test and Startup Branch
1.
Persons contacted
The following persons were contacted during the inspection:
Arkansas Power and Light Company (AP&L)
P. L. Almond - Reactor Technician
J. W. Anderson - Plant. Superintendent
T. C. Baker - Chemical and Radiation Protection Engineer
A. Bland - QA inspector - Civil
'
R. G. Carroll - Chemical and Radiation Protection Engineer
"
!
W. Cavanaugh, III - Manager of Nuclear Services
T. H. Cogburn - Nuclear Engineer
R. R. Culp - Test Administrator
}
T. Green - Production Department Training Coordinator
j
, C. A. Halbert - Technical Support Engineer
l
H. Hollis - Administrative Assistant
}
J. Marlin - Assistant Engineer - Production Department
G. H. Miller - Assistant Plant Superintendent
i
N. A. Moore - Manager of Quality Assurance
<
J. L. Orlicek - Quality Control Engineer
D. R. Sikes - Results Engineer
i
M. H. Shanbhag - Plant . Staff Engineer
C. N. Shively - Procedure Administrator
B. A. Terwilliger - Operations Supervisor
J. Vincent - Arkansas Nuclear One Training Coordinator
J. H. Woodward - Director of Power Production
.
Babcock and Wilcox Company (B&W)
F. J. Sattler - Manager, Inservice Inspection
4-
Bechtel Corporation (Bechtel)
L. Tilley - Field Engineer
'
-
.
.
-
..
-
-.
-
-
-
._.
.-
.
_
_
~
.
.
~
^'
- -
.
,
RO Rpt. No. 50-313/74-9
I-2
.
i
'
2.
Project Status
The ~ inspection was concluded May 17,-1974, with a management interview
in which commitments were obtained on those items to be completed /
resolved prior to licensing and prior to initial criticality. Those
commitments were as follows:
'
a. _ Items to be Completed Prior to Licensing
(1) Incore Guide Tube Hydro
On May 10, 1974, AP&L discovered that certain of the incore
monitor detectors could not be inserted into their guide
tub es .
(See FSAR Section 7.3.3 and Figure 7-24) Discussions
with licensee personnel revealed the following history.
The guide tubes are schedule 80, one half inch piping with
'
six shop welds. Apparently, the shop welds were not bored
out sufficiently at the shop in view of a subsequent change
to the'incore detectors. This change involved the addition
-
of a protective sheath over a portion of the detector to
'
O
- prevent vibrational wear and was made to all detectors at B&W
177 fuel assembly plants following observation of a wearing
,
problem at Oconee 1.
The outside diameter (OD) of the
i
detector was approximately .292 inches without the sheath
and increased to approximately .420 af ter addition of the
'
i
sheath.
!
The ANO 1 guide tubes were checked with a .375 inch probe in
-
the fabricator's shop prior to the sheath addition. After the
sheath addition, the Oconee 1 detectors were reinserted with
,-
no problems; therefore, AP&L felt that there should be no
,
4
problems at ANO 1.
Revised detector drawings showing
sheaths were provided to AP&L in March of 1974, but these
,
did not give dimensions for the sheath OD.
AP&L elected to ream out the guide tubes to allow the passage
of a .517 inch probe. A procedure, OP 1701.01, " Repair of
,
Incore Instrument Tubes ," was written and approved on May 12,
l
!
1974, to control the reaming operation. The reaming operation
was designed to reduce the excess penetration of the shop
'
welds while leaving the tube walls intact.
(ID approximately
.547 inches)
.The reaming procedure was discussed with licensee personnel
.
on May 17, 1974. It was noted that no accpetance criteria
relating to the integrity of the welds or pipe walls had been
l.
established.
It was concluded that there was no method of
.
t
Ij'-
-
-
.
-
..- . . - .
._
__ .
, _ _ _
_ , _ _ _
.
.
'
-
r.
,
,
.
R0 Rpt. No. 50-313/74-9
I-3
verifying the adequacy without rehydroing the tubes, repeating
the original hydro. Licensee personnel stated that this
would be accomplished prior to licensing.
(2) Security
The physical security systems had been completed, but certain
items had not been implemented.
(3) Completion of Test Deficiencies
,
(a) TP 351.30
" Computer and Control Room Heating and Ventilation
Test"
Isolation damper CV-7906 would not close in response to
signal because of a pinch in the sensing line to the
pressure switch. The pinch was to be located and
-
'
corrected or a new sensing line run.
('
The emergency air reservoirs for dampers CV-7905, CV-7906,
N--
and CV-7907 had not been installed. The reservoirs are
designed to provide air to operate the dampers in the
event of loss of instrument air. Licensee personnel stated
that pressurized nitrogen bottles would be connected to the
dampers until the reservoirs are received and installed.
.
Two smoke detectors were found to be reversed. This error
would have caused the normal system to isolate and circulate
the smoke within the control. The detectors were to be
relocated and tested.
(b) TP 362.01, " Effluent Monitoring Check at Power"
i
The data to be gathered at background levels prior to
fuel loading had been taken, but had not been reviewed and
approved by AF&L. This review and approval was to be
completed prior to licensing.
A licensee representative informed the inspector by telephone
May 21, 1974, that the items discussed above had been satisfactorily
completed.
.
)
.~-
.
.
.
-
. - .
-.
-
- _ . . - . . _
. , .
_ _ . . _ _ _ . -
-
'
.-
/- .,
.
\\'
'
R0 Rpt. No. 50-313/74-9
I-4
.
b.
Items to be Completed / Resolved Prior to Initial Criticality
(1) Steam Generator Hydro
As discussed in R0 Report No. 50-313/74-6, Details I, paragraph
3.e., AP&L installed block valves upstream of the atmospheric
dump valves on each main steam header. This requires a
rehydro of the steam generator secondary side and main steam
lines. This can be accomplished with least delay after fuel
loading in that it requires the RCS to be heated slightly
such that both sides of the steam generators are above ambient
temperatures.
(2) Checkout of RCP's
i
Subsequent to hot functional testing (HFT), the RCP seals
were removed for inspection and replaced.
In addition,
quick disconnect instrumentation cables to pumps and motors
were replaced with cables terminating in conventional
-
terminal boxes to improve leak-tightness of the connections.
In that the RCP's are not required by Technical Specifications
until preparation for criticality is made,' AP&L committed
to checking them out prior to initial criticality.
>
(3) Pressurizer Code Safety Valve
During HFT, as the pressurizer safeties were being tested, one
of them did not reseat properly. The resultant steam flow
-
across the seat of the valve eroded it slightly. After
cooldown, the seat was replaced. The setpoint on this valve
is to be rechecked after heatup and pressurization in
,
preparation for initial criticality.
(4) Service Water Isolation Valves
Air accumulators are to be installed on service water system
valves CV-3812, 3813, 3814, and 3815. These valves isolate
the feed to the reactor building coolers. The accumulators
will enable the valves to remain closed in the event of loss
of instrument air concurrent with a high radiation signal.
(5) Main Steam Piping Restraints
Shims are to be installed on the main steam line piping
-
restraints outside of reactor building (inside building
complete) prior to initial criticality. The shims are to
be added as a result of evaluations of pipe deflections
g
q
)
observed during HFT.
_
.
___
'
-
n
S
,
'
RO Rpt. No. 50-313/74-9
I-5
j
(6) Grating Over Steam Generators
l
The licensee determined that it was desirable to remove
'
the grating directly over the steam generators for operations
to increase cooling air flow in the compartments. Seismic
tie-downs are to be provided and the grating appropriately
stored prior to initial criticality,
j
(7) Quench Tank Rupture Disk Deflector
A deflector is to be installed over the quench tank rupture
disk in order to protect pressurizer instrumentation li s
directly overhead in. the event of rupture. This is to be
accomplished prior to initial criticality.
(8) DC Panel Transfer Switches
The licensee decided to replace the: manual transfer switches
for DC distribution panels D11 and D12 (See FSAR Figure 8-1)
with what is felt to be a more reliable type switch. This
is to be accomplished prior to initial criticality.
(9) Preoperational Test Deficiencies
All preoperational tests which had not received final endorse-
ment by AP&L were reviewed and discussed with licensee
personnel.
(Details I, paragraph 26.d and Details V,
paragraph 4) Of those which had not been endorsed, the
.
following contained deficiencies which must be resolved prior
to initial criticality.
(a) 165.01
" Filter Tests"
(b) 172.01
"Penerration Room Ventilation Test"
'
(c) 201.03
" Core Flood System Functional Test"
(d) 231.69
" Dirty Radwaste System Test"
(e) 234.01
" Resin Sluicing Tests"
,
(f) 240.14
"ICW System Test"
(g) 273.36
" Aux. & Emerg. W System Test"
,
(h) 276.44
" Condensate System Test"
}
(i) 330.04
"CRD Integrated Test"
..)
~
.
~
i
i
\\
'
.m
o
,
fO)
'
RO Rpt. No. 50-313/74-9
I-6
'
l
(j) 351.30
" Computer and control Room Air Conditioning Pre-Op"
.
(k) 370.01
" Reactor Building Hydrogen Removal System"
)
(1) 500.03
" Initial Radio Chemisit ' Test"
i
(m) 600.13
" Pressurizer Operation and Spray Test"
l
'
(n) 600.14
" Pipe / Components Hot Deflection Test"
(o) 600.17
"CRD Operation Test"
.
In addition to the items above, licensee representatives committed to
resolving / completing the previously identified unresolved items in
paragraphs 17 and 23 of this Details section and Item 72-12/2, the
thin walled valve program. Licensee representatives stated that a
letter of commitment to DL on these items would be presented to DL
i
at the time an operating license was requested.
3.
Status of Previously Id'entified Enforcement Matters
a.
Failure to Follow Test Procedure 200.08, " Pressurizer Relief Valve Test"
I
This item was discussed in R0 Report No. 50-313/74-2, Details II,
paragraph 3.a. , and involved the failure to conduct the test at the
reactor coolant system temperature specifie.d in the procedure.
The licensee response to the notice of violation on this item states
,
that it ic an AP&L policy to conduct all tests according to a
written and approved procedure. Also, it indicates that this was
an isolated case and that continued emphasis was being placed on
adherence to procedures.
i
This test was rerun using a completely revised procedure. The test
)
was rerun because of the problem discussed above and the fact that the
setpoints were marginal. During the rerun, conducted February 26,
1974, one of the two code valves did not reseat properly leading to
the disc face being cut slightly. The disc has been replaced and
the setpoinc is to be rechecked prior to initial criticality. This
item is considered closed.
'
b.
Procedures Not at Work Location
This item was discussed in R0 Report No. 50-313/74-2, Details III,
1
paragraph 2.a.
During the current inspection, the inspector
verified that a copy of procedure 1004.11. " Handling, Storage, and
Shipping of Q Listed Materials" had been provided in the storeroom.
j
All quality control and administrative procedures are provided
).
in the storekeepar's office. This item is considered closed.
,
,
1
L
'
_
_
_
__
.
. .
.
_ _ .
_
_
_
_
_
_
_
e
I
A
.
"
.
.
'
RO Rpt. No. . 50-313/74-9
I-7
.
J
c.
Changes to Procedures
.
This item was discussed in R0 Report No. 50-313/74-2, Details III,
'
paragraph 2.c.
As discussed in the licensee response dated
March 28, 1974,. a memorandum to all ANO employees dated February 19,
1974, restores the authority for approval of major changes to test
procedures to the plant superintendent or his representative
'
(assistant superintendent or department head). The memo provides
for verbal approval, if necessary, to be followed by written
approval, normally within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. This item is c,onsidered closed,
d.
Design Changes
,
This item was discussed in RO Report No. 50-313/74-2, Details III,
paragraph 2.d. , and involved controls over a change to the power
,
supply boards for the pressurizer level and reactor coolant flow
i
transmitters. The response in this item states that after
reviewing the matter again AP&L determined again that the change
i
made to power supply board did not constitute a design change.
-
The inspector discussed the power supply board change again during
the current inspection and after a detailed explanation of the exact
j
change made, concurred that a design change had not been made.
!
This item is considered closed.
e.
Spent Fuel Area Radiation Monitor
The failure to calibrate this monitor at timely intervals was
discussed in R0 Report No. 50-313/74-4, Details III, paragraph 3.
The response from AP&L stated that a program for maintaining the
.
instrument had been set up.
Plant records reveal that the monitor
l
was recalibrated' February 8,1974, the day after the problem was
. -
'
identified. It was calibrated using preoperational test procedure
360.79, " Radiation Monitoring System Preop," which utilizes
the calibration procedure recommended by the manufacturer. The
instrument is to calibrated quarterly, the same frequency as given
'
i
in the proposed technical specifications. This item is closed.
f.
Review of the Station Log
This matter was discussed in R0 Report No. 50-313/74-2, Details III,
paragraph 2.e.
AP 1005.01
" Administrative Controls Manual," had
'
been revised to require only a weekly review of log, whereas Section
l
12.5 of the FSAR requires a daily review. The licensee response
stated that 1005.01, would be revised by April 1,1974, to agree
-
with the FSAR. A review of the procedure on May 1,1974, revealed
that this change had not been made.
The inspector brought this
fact to the attention of plant personnel and it was revised on the
s
,
4
---m=
4
--, -- -
-,v
+-
,
- -,
,,.--.w
- . . - - -
---2-
+me-
.----s.,i
- - . . . , - - - - . .
m-
.,,.w---s-vs--
- ---,eawee+ow.n--ew-=m
e
-
4
'
o)-
.
iV
R0 Rpt. No. 50-313/74-9
I-8
same date. The procedure, as revised, provides for a daily review
(except on weekends and holidays) by the operations supervisor or
his designated representative. This is in agreement with the FSAR.
This matter is considered closed.
g.
Reporting of Design Changes
The subject of reportability of the change to the power supply boards
_
discussed in paragraph 3. above and the change in lubricating oil
for the reactor building spray pumps were discussed in R0 Report No.
50-313/74-2, Details III, paragraph 2.d.
The fact that a design
change was not made to the power supply boards is noted in paragraph
d.
Regarding the spray pump lubricating oil, a review of the results
of TP 204.03, " Reactor Building Spray System Functional Test,"
revealed that the thrust bearings overheated slightly (173*F vs
recommended limit of 170*F). The pump vendor was contacted and
informed AP&L that the limit could be changed to 200*F with no
ill effects. B&W recommended a change to a less viscous oil, which
was done by AP&L after review by station personnel. After the oil
was replaced, bearing temperatures ranged from approximately 165'F
p
to 168'F, reflecting the slightly lower viscosity of the new oil.
The AP&L response dated March 28, 1974, states'that the matter was
handled properly under the design and construction phase controls
in effect at the time. Af ter review of the matter vd discussing
it with station personnel, the inspector stated that he had no
further questions and that the matter was considered closed.
h.
Reporting of Pressure Transmitter Deficiencies
Failure to promptly report the failure of the four narrow range
reactor coolant system pressure transmitters to meet specified
response times per 10 CFR 50.55 fe) was discussed in RO Report No.
50-313/74-4, Details I, paragrap i 3.
A licensee response, dated
March 28, 1974, has been recei<ed. The response indicates that
AP&L is dependent upon B&W and Bechtel to inform them of reportable
items, and states that these parties have been urged to notify
AP&L in a timely manner of such problems. A report per 10 CFR 50.55(e),
dated March 22, 1974, has also been received. This report states
that studies showed that the increased sensor time delay would
have caused the high pressure safety limit of 2750 psia to be
exceeded if two or more banks of control rods were moved out of
the core simultaneously.
The inspector was informed that the old transmitters which did not
-
meet the specified time delay of 250 milliseconds or less were made by
Foxboro and were model number EllGM-SAE1. The replacement transmitters
are by Westinghouse, are Veritrak Model 59 PH, and have response times
of 100 millisconds or less. This item is considered closed.
.
--
_ _ .
__ _
__
M'
.e
.
- , .
.--
.u_,
, , ~ ,
. _ .
d
.
a
.
.
'
'
'
RO.Rpt. No. 50-313/74-9
I-9
-
,
4.
Respiratory Protection Program
The status of this unresolved item as of May 2, 1974, is discussed in
Details III, paragraph 2, of this report. On May 14,1974, the inspector
confirmed that training in the use of respirators had been completed
-
by review of records and that all respirators had been stored. This
item is considered closed.
,
,
,5.
Diesel Generator Trips
'
This unresolved item was initially discussed in R0 Report No. 50-313/73-12,
Details I, paragraph 4.
A licensee report dated October 31, 1974, and
entitled " Loss of Power to Vital Buses" described corrective measures
.
to be taken to prevent a similar loss of power to the 120 volt A.C. vital
buses. During the current inspection, the inspector was informed that
modifications to the plant inverters and retesting had been completed
i
May 4, 1974.
The results of the additional testing performed were reviewed. Testing
- '
was accomplished using an addendum to test procedure 400.03,
"A.C.
Power System Energization." The purposes of this test were to
demonstrate the ability of the inverters to provide continuous
<
uninterrupted power to the vital buses; to demonstrate the ability to
s.
automatically' switch to the alternate 125 Volt DC input source upon
i
loss of 480 V A.C. input or inverter automatic lockout of the
!
480 V A.C. input due to high voltage; to demonstrate the ability
to transfer to the alternate emergency AC source through a static
transfer switch when an output undervoltage or output overcurrent
.
condition occurs; and to show that the inverters can maintain output
frequency within limits when the alternate emergency AC input .
frequency exceeds specified 1Laits,
t
Results verified the adequate of the plant inverters, including the
sp are. In addition to the testing reflect 2d above, load tests of
10, 25, 50, 75, and 100% were run on each inverter. The test
results had been reviewed and approved in accordance with plant
procedures. The inspector stated that he had no questions and that
the item was considered resolved.
6.
Control Rod Trip Test
This unresolved item was initially discussed in R0 Report No. 50-313/
73-12, Details I, paragraph 5.
After the inspection documented in
RO Report Nc. 50-313/74-4, (See Details I, paragraph 11) three
comments on TP 330.05 remained open. During the current inspection,
the inspector was informed that instead of rewriting TP 330.05, it had
been decided to use OP 1304.35, " Control Rod Drive Trip Test," to
. accomplish.the at-power trip tests. . Review of 1304.35 revealed that
the inspectors comments had been resolved as follows:
-
!
l
_.
.
t
4
-
%
,-_ _ _ _ - . - _ , _ _ _
,m
,
- . . _ . ..
. . .
- , , , , . _ .
, . . , _ . , _ , _ . , . - - -
. , , _ _ - _ _
-
O
.
m
.-
'
.
R0 Rpt. N . 50-313/74-9
I-10
'
o
a.
References have been incorporated in the procedure to OP 1103.15,
" Reactivity Balance Calculation," as the -means of calculating
b.
Plant conditions are specified for four tests.
If additional tests
are run, they will be run at one of the sets of conditions
already given,
c.
All rods mechanisms were tripped together during HFT and times
measured. Licensee personnel stated that if an adequate shutdown
margin could be maintained without undue boration, then all groups
(one through seven) would be tripped simultaneously during low power
testing also.
The inspector stated.that he had no further questions and in that
the testing planned appeared to meet that recommended by Regulatory
Guide 1.68,
This matrer was considered resolved.
7.
Initial Core Load Procedure
~
g'"N
This unresolved item was initially discussed in RO Report No. 50-313/
73-14, Details I, paragraph 4.
The status of the procedure (1502.04)
i\\s-
and RO concerns on it were discussed in R0 Report No. 50-313/74-6,
Details I, paragraph 9.
The procedure received final approval
April 18, 1974. Review of the procedure revealed that all previous
emaments had been resolved. Revision 1 of 1502.04 appears to cover
all commitments and requirements of Section 13.3.1 of the FSAR and
proposed Technical Specifications 3.8.1 - 3.8.12.
In addition, it
addresses the recommended prerequisites, limitations and actions,
and procedural details discussed in Section B of Appendix C to
RG 1.68, "Preoperational and Initial Startup Test Programs for Water-
'
Cooled Power Reactors." This iten is considered resolved.
8.
Pressurider Electromatic Relief Valve
This unresolved item was initially discussed in RO Report No. 50-313/
73-17, Details I, paragraph 4, after questions regarding the design
i
adequacy of this type valve were raised by RO inspectors at another
B&W facility. During the current inspection, licenseee personnel
stated that B&W had assured them that the valve for Unit 1 had been
l
constructed using the same procedures as those for other B&W
plants. Alsc, licensee personnel stated that a review of quality
documentation related to this valve had revealed that all documentation
requested on the B&W purchase order to Dresser had been received and was
on file. On May 17,1974, the inspector informed licensee personnel that
this item was considered resolved in that he had been informed by other R0
{V'~'}
personnel that the adequacy of design of the valve had been assured.
-
'
_.
_
_
___
.
~
.
.
e
-
m.'
.
)
,
s_ /
.
50-313/74-9
RO Rpt. No.
' 11
-
-
9.
Emergency Operating Procedures
General and specific comments of the plant emergency operating procedures
were documented in RO Report No. 50-313/73-17, Details III, paragraph 2.
Paragraph 24 of this report section itemizes the procedures which have
been reviewed since the initial review to determine that RO concerns
i
have been resolved. Review of the revised procedures revealed that
RO comments were considered and procedures revised accordingly. As
.
of May 15, 1974, all emergency operating procedures had been developed
and approved by AP&L and reviewed by RO:II, with no comments outstanding.
i
This item is considered closed.
i
10.
Operational Test Program
'
This item was initially discussed in RO Report No. 50-313/73-17,
Details III, paragraph 3.
As of that report date, November of
1973, several procedures which would implement the surveillance testing
testing to be required by technical specifications had not been
identified by AP&L. In addition, only a small portion of these
procedures which had been identified had been written and approved.
Os _,,
During the current inspection, it was verified that all test
procedures recommended by Regulatory Guide 1.33 and/or required by
technical specifications had been included on the station procedure
.
index. In addition, all surveillance test procedures had been written
and approved with the exception of two which will be used to conduct
refueling tests. Licensee personnel stated that those two would
. be approved in the near future.
Discussions with department heads responsible for surveillance testing
revealed that schedules had been developed for testing. Each group
,
had developed a listing of pre-critical tests which had been
incorporated into OP 1102.01, " Plant Preheatup and Precritical Check-
list." Also, computerized listings were available for the various
groups. The listings will cover testing required for each week and
also periods such as refuelings. These listings will be supplemented
,
by schedules maintained manually for a trial period. The inspector
stated that this item was considered resolved.
11. Emergency Planning
RO concerns regarding implementation of the ANO, Unit 1, emergency
plan were initially discussed in R0 Report No. 50-313/73-18,
Details I, paragraph 2.
The results of inspections conducted through
May 2,1974, to determine the degree of hnplementation are presented
in Details III, paragraph 4, of this report. On May 14, 1974, the
f- s
inspector verified that the emergency control center radio and
Q,_,))
.
(
.
._
_
.-- -
._.
-
-
- - - - - - . - - _ .
.-
._
.
-
WP
,
.
'
m
,
4
,
'
RO Rpt. No. 50-313/74-9_
I-12
.
telephe2. had been installed and were operable. In addition, he
noted tnat ainor-problems with health physics instruments in the
control center ^had been corrected. The inspector stated that this
item was considered closed.
12 .~ . Calibration of Radiation Monitors
'
.
This unresolved item was initially discussed in RO Report No. 50-313/
73-18, Details I, paragraph 4.
The status of these instruments on
May 2,1974, is discussed in paragraph 5 of Details III of this report.
.!
~
On May 16, 1974, the inspector reviewed the completed test package
i
for TP 360.79, " Radiation Monitoring System Preop," which was used
to calibrate and functionally check area and process monitors.
T
Results indicated that all monitors except RE-4642, the liquid
radwaste system process monitor, had been calibrated successfully.
This monitor had been calibrated earlier in the test program, but
its detector subsequently failed. This instrument will be' required
to pass a surveillance test within one week af ter initial criticality
per proposed Technical Specification 4.1.a.
The inspector stated
j
that he had no further questions and that the item was considered
resolved.-
,
l
13. Reactor Building Ventilation System Ductwork
s
This item was initially discussed in RO Report No. 50-313/73-19,
i
Details I, paragraph 5.
Bechtel had determined through a design
review that the Unit 1 ductwork might not withstand the differential
.
. pressure experienced during a loss-of-coolant accident. Destructive
.
tests were conducted on mock-ups of the ductwork by a subcontractor
in order to determine the amount of rework which would be needed.
l
A repcrt per 10 CFR 50.55(e) giving the results of two tests was
-
submitted April 1, 1974. A supplemental report dated April 26,
,
1974, gave the results of the third and final test. These reports
describe the procedures used to vacuum test the ducts and the test
results. In addition, corrective actions required to bring the
ductwork up to required strength are given.
<
.
During the current inspection, the inspector inspected modifications
>
which had been made in the field and verified, on a sampling basis,
,
'
that the actions required per Section 8.0 (Closecut Action) of the April
26, 1974, report had been' completed. The addition of selected stiffeners
specified in Section 8.1, the addition of angles to stiffeners and
hangers per Section' 8.2 and the addition of welding between hangers
and ducts and stiffeners and ducts per Section 8.3 were verified.
l
Also, several sections of round duct were observed to verify that addi-
~
tional hangers had been added where needed to assure a distance of twelve
feet or 'less between hangers per Table I of the April 26 report.
- O
-
!
i
'
.
_.
_
_
- 1
.
-
.
.
.
,
.
..
-
,
,
V
R0 Rpt. No. 50-313/74-9
I-13
The inspector informed licensee personnel that based on his observations
in the field, it appeared that all corrective action had been completed
and that the matter was considered resolved. He noted that DL had
been requested to evaluate the adequacy of design of the ductwork
af ter modification and that AP&L would be contacted by DL if questions
arose from that review.
14.
Vibrations in Reactor Vessel
This unresolved item was initially discussed in RO Report No. 50-313/
74-2, Details I, paragraph 2.
During the conduct of hot functional
testing (HFT) on February 4,1974, vibrations / noises were heard in
the upper and lower section of the reactor vessel on the loose parts
and vibration monitor (LPM) . Several tests were conducted involving
combinations of the three operable reactor coolant pumps (RCP)
at various temperature and pressure levels.
RCP "D" was inoperable
at th-* *4ae due to a seal staging problem. During the current
,
inst
.he inspector reviewed a report to J. W. Anderson from
the
, Test Program Coordinator, dated April 1,1974, which summarized
f "'S
the results of tests run. All test data and recordings from the LPM
e
g
had been analyzed by B&W. The inspector was informed that AP&L
(Plant Superintendent and the Plant Safety Committee) had reviewed
,
the report and concurred with its conclusion.
'
The report concluded that the noise was most prominent with one RCP
running alone, and was more severe with RCP "C" running alone than
- either "A" or "B" . Also, the noise diminished with a decrease in RCS
temperature. In addition, an accelerometer which had been placed on
the pressurizer spray valve revealed that this valve was not the source
of the noise.
It stated that after HFT, during the conduct of inspection
of the vessel intervals per TP 600.32, no loose parts were found in the
vessel or intervals.
~he west irradiation surveillance specimen tube
was found to be loose.
(See FSAR Section 4.4.5) The upper locking pin
,
would not seat properly, indicating that the spacing between pintles
was incorrect. Slight wear was seen on the locking pin and groove,
indicating that the tube had probably vibrated during HFT. This
problem was to be rectified prior to fuel loading. An inspection
of "B" steam generator upper tube sheet revealed no loose parts or
evidence of damage. The report concludes that thhe cause of the
noises / vibrations was the loose specimen tube.
Supporting data and
recorded observations appeared to confirm this conclusion.
The inspector was informed that further testing to obtain baseline data
for the LPM will be conducted after fuel loading and prior to initial
criticality. The inspector stated that this item was considered resolved.
,,_ \\
[Q
-
. . -
e
. . _ _ . _ -.-
__
_
_
__
- _ _ _
_ _ . _ . _
_ ___.
_-
.
.
.
.
..
i
-
.
~
~3
3
-'
s
'
.
R0 Rpt. No. 50-313/74-9
I-14
.
,
j.
15. Control of Maintenance Activities
l
This unresolved item was initially discussed in RO Report No. 50-313/
74-2, Details III, paragraph 2.e.
The Administrative Controls
Manual, OP 1005.01, discussed four levels of maintenance activities,
including a level 4 which required no procedure for activities that
'
are simple and routine, i.e. torque settings, any minor adjustment.
RO's concern was that for torque switch settings and similar activities
where an exact setting is important, some degree of control over that
'
in effect should be exercised. The manual was revised April 18, 1974,.
to better define the controls over all levels of maintenance activities.
Level 4 activities can be accomplished through the use of a Job Order,
-
but specific instructions for completing the job, such as torque
setting, are to be provided on the job order by the cognizant
'
supervisor. This item is considered closed.
16.
Control of Temporary Circuit Modifications
!
'
This unresolved item was initially discussed in F0 Report No.
.
50-313/74-2, Details III, paragraph 2.f.
Two wasicnesses in the controls
for the use of jumpers and bypasses were identified. Procedure 1005.04,
-
I
" Control and Use of Jtmapers (M 3ypasses," has b4 en revised to correct
.
these weaknesses.
It provides that all jumpers qnd bypasses applied
to equipment anywhere in the plant outside of shaps and labs are to be
recorded in the Bypass and Jumpers Log. Also, each log sheet is to
be ntmbered subquentially and all tags are to be serially numbered.
<
This item is considered e.losed.
'
j.
-
17. Leak in Pump P36B Recirculation Line
-
t'
During hot functional testing, a leak was observed in the recirculation
line flow orifice (FO-1242) for high pressure fajection (HPI) pump
P36B. A written report per 10 CFR 50.55(e) ent Ltled " Makeup Pump
Orifice Leak" was submitted April 15, 1974. This report attributes
the cause of the erosion of the orifice wall tt high exit flow from
the last stage'of the orifice. Also, the last stage exit was
located . eccentrically in the orifice shell suct that its discharge was
i
-near the wall'of the orifice. Corrective action included modifying the
orifice design to reduce the exit velocity from the last stage from
,
a maximum of 183 fps to 110 fps and.~ to locate the last stage exit
,
'
concentrically in the shell. As of May 17, 1974, the orifices for
i
pumps P36A and P36C had been replaced with the new design type.
- Hydrostatic tests had not been completed. Also, the welding for
the P36A orificc. contained a defect requiring repair.
l
m
'
.
l
l
~
_ - _.. _ , . _ .._
.
_ _ . -_
- - -
_
.
.
.
.
.
-
-
.
.
.
'
'
-
,
,
R0 Rpt. No. 50-313/74-9
I-15
'-
.
Repair could not be made in that craf ts were on strike. AP&L committed
to.a final _rasolution of this problem prior to initial criticality.
The inspector stated that this item would remain open.
18.
Erratic Behavior of CRD Position Indicators
AP&L reported a possible significant deficiency concerning the
failure of several relative position indicators (RPI) and absolute
position indicators (API) for control rods during hot functional
.
testing on March 4,1974. A written report dated April 5, 1974,
concludes that the problems experienced represent operational restrictions
but do not have safety significance. The inspector reviewed the proposed
Technical Specifications 4.7.1.2, 4.7.3.3, and 3.5.2.2 which will govern
operation with position indicator failures as well as true abnormal
conditions. These restrictions, coupled with instructions in procedure
1203.03, " Control Rod Drive Malfunction Action," should assure safety
of operation. The inspector stated that he concurred that the problem
was an operational one and not of safety significance. This matter is
considered closed.
-
19. Replacement of Quality Breakers with Non-Quality Documented Ones
0i
This item was initially discussed in R0 Report No. 50-313/74-6,
Details I, paragraph 5.
It involved the inadvertent replacement
of pressurizer heater breakers B-5144 and B-6144 with non-quality
'
documented breakers. A written report per 10 CFR 50.55(e), dated
May 1,1974, and entitled " Identification of Breakers as Non-Q"
states that maintenance personnel were not aware that the breakers
.
were "Q" items in that identification as safety related devices was
not apparent. These breakers isolate motor control centers BS1 and
B61 (safety related) from the pressurizer heaters (non-safety related)
in the event of faults in the heaters. The error in identification
was discovered through quality control review of documentation associated
with the replacement.
The inspector reviewed Job Order No. 251, dated March 18, 1974, which
was used to install the proper breakers. He also noted that a
Non-Conformance Report (NCR) had been written by quality control
personnel May 9,1974, due to the fact that QA documentation for the
new breakers had not yet arrived on site. A letter from Bechtel to
AP&L was attached to the NCR which stated that the required documenta-
tion was forthcoming. The inspector stated that although this practice
was in conformance with the ANO Quality Control Manual at the time of
use of the latter set of breakers, it would not be acceptable per
the manual as revised in April of 1974. Licensee personnel agreed
with this comment. This item is considered closed.
)
O
.
.
--
~
---
-. --
.
.
.
.
T
.
\\s_ >
'
R0 Rpt. No. 50-313/74-9
I-16
20.
Sensitivity of Leak Detection System
This unresolved item was discussed in R0 Report Ni. 50-313/74-6,
Details I, paragraph 3.f.
As explained in that report, the results
of hot functional testing did not appear to confirm the sensitivity
of the inventory balance method of primary system leak detection as
described in Section 4.2.3.8.b of the Unit 1 FSAR. Licensee personnel
stated that the FSAR would be revised to clarify the ability and
sensitivity of this leakage detection method or a known leakage test
similar to that conducted per TP 600.10, "RCS Hot Leakage Test," would
be rerun during power ascension testing. The inspector stated that
this item would remain open.
21.
Availability of Startup Test Procedures
The majority of zero power physics and power ascension test procedures
had not been approved on April 5,1974, when this matter was documented
as an unresolved item in RO Report No. 50-313/74-6, Details I,
ps
paragraph 6.
On May 17, 1974, the inspector was informed that all
(d)
of these procedures had been written and approved and were available
\\
for his review. This item is considered closed.
22.
Quality Documentation of Valves
This unresolved item was discussed in EO Report No. 50-313/74-6,
. Det.ils I, paragraph 3.e.
The test results package for TP 200.09,
" Steam Generator Secondary Side Hydro Test," contained a statement
that a rerun of the hydro was required because it was found that
certain small valves in the system did not have quality documentation.
.
Detailed information concerning these valves could not be obtained
during that inspection. During the current inspection, discussions
with AP&L QA personnel revealed that the eight instrument root valves
in question did indeed have documentation, but were not "N" Stamp
valves as required by code. A representative stated that earlier
in the test progras for Unit 1 AP&L decided to install several small,
non "N" stamp valves where "N" stamp ones could not be obtained
'
such that testing could proceed. Plans were to replace the valves
when the permanent ones were received and retesting done where needed.
Apparently the tags used to identify these valves as ones to be
replaced were lost. The fact that they were not "N" stamped and
should have been was discovered by Bechtel QC as part of their
program of tracing out all systems in the-plant to assure that all
appropriate components had stamps.
The inspector was assured that
sJ
,
.
.
[
l
!
l
.
l
,
.,
,_
-
.
.
.
'
e3
+
.
RO Rpt. No. 50-313/74-9
I-17
,
all components had been checked and found satisfactory.
In that the
'
licensee QA program had ' identified and corrected the problem and the
rehydro had been completed, the inspector stated that he had no
further questions and that this item was considered resolved.
23. Baseline Inspection Data
This item was initially discussed in R0 Report No. 50-313/74-7,
Details I, paragraph 2.b.
Discussions with licensee and B&W personnel
during the current inspection revealed that the problem areas
discussed in the referenced report can be resolved at the time the
final baseline report is submitted. AP&L committed to resolving these
problems prior to initial criticality. This item remains open.
24. Regulatory Operatidns Bulletin 74-3
A response to ROB 74-3, " Failure of Structural or Seismic Support Bolts
on Class I Components," has been received. The response, dated April.22,
1974, states that inspection of apprcpriate vecsel supports will be
conducted.during the first refueling and that detailed written programs
.
and procedures will be available for RO inspection prior to that time.
The inspector stated that he had no further questions on this matter.
This item is closed.
25. Procedure Reviews
,
i
.a.
Emergency Operating Procedures
Results of previous reviews of the station emergency operating
procedures were given in R0 Report Nos. 50- 313/73-17, Details II,
,
'
paragraph 2, 50-313/74-2, Details II, paragraph 2, and 50-313/74-6,
Details I, paragraph 4.b.
The current inspection effort was
directed toward determining whether RO comments had been
considered on procedures previously reviewed and to complete the
review of procedures not previously available. Findings were as
'
follows:
(1) Procedures Previously Reviewed
Review of these revised procedures revealed that weaknesses
which had been previously commented on had been corrected:
1202.07, " Moderator Dilution" - Rev. 3
1202.11, "High Activity in Reactor Coolant" - Rev. 1
1202.12, " Loss of Instrument Air" - Rev. 3
'
1202.13, " Loss of Service Water" - Rev. 3
.
l
,
f
- - - , - -
, - -
- - - ,
,
--
_.
_ _
4
.
.
.
-
,
l
'
'
m
s ,
y
.
.
l
I 18
\\-
RO Rpt. No. 50-313/74-9
1202.15, " Loss of Reactor Coolant Makeup" - Rev. 1
1202.16, "RC Pump and Motor Emergencies" - Rev. 1
1202.17, " Loss of Neutron Flux Indication" - Rev. 1
1202.18, " Emergency Shutdown" - Rev. 1
1202.23, " Steam Generater Tube Rupture" - Rev.1
1202.26, " Loss of Steam Generator Feed" - Rev. 1
i,
1202.29, " Pressurizer System Failure" - Rev. 2
1202.32, " Loss of Decay Heat Removal System" - Rev. 1
1202.33, " Emergency Operation of NSS" - Rev. 1
,
~
A review of Revision 1 of 1202.02, " Blackout," resulted in
minor comments. Licensee personnel stated the procedure
f
would be revised to reflect them.
(2) Results of. Initial Reviews
These procedures were compared to the guidelines of ANSI N.18.7,
" Standard for Administrative Controls for Nuclear Power Plants,"
as were the other emergency procedures. The inspector had
no comments on these procedures:
1202.24, " Steam Supply System Rupture" - Rev. 0
1202.40, " Refueling Accident" - Rev. 0
1202.45, " Gas Line Rupture" - Rev. 0
1203.01, "ICS Abnormal Operations" - Rev. 0
1203.04, " Reactor High Startup Rate" - Rev. 1
'
1203.05, " Loss of Containment Integrity" - Rev. 1
s
.
Review of 1202.05, " Degraded Power," (Rev. 0) resulted in
minor comments, which licensee personnel agreed to resolve.
Review of Revision 0 of 1202.38, " Fire or Explosion," revealed
that it did not speak to explosions. Also, it did not give
precautions for use of special techniques for certain fire areas,
such as nuclear fuel storage areas or fires near electrical
equipment. The procedure was rewritten while the inspector was
,
onsite to correct these weaknesses.
The inspector stated that all emergency operating procedures had been
reviewed and that there were no outstanding comments on them.
b.
Surveillance Test Procedures
f
Results of previous reviews of selected surveillance test procedures
were discussed-in R0 Report No. 50-313/74-6, Details I, paragraph 4.c.
'
Review of these revised procedures revealed that previous comments
and questions had been resolved:
'
.
Ns
-
.
.
-.
.
. .-
.
.
.
.
'
O
.-
,
,
i_/
-
s
RO Rpt, No. 50-313/74-9
I-19
.
.
.
1304.45, '"ESAS Digital Subsystem #1 Test" - Rev. 0
(previously 1304.05)
1304.49, "ESAS Analog #1 Test" - Rev. 0 (Previously 1304.02)
1304.55, "ESAS Coincidence and Manual Trip Test" - Rev. o
(previously 1304.07)
1304.08, " Integrated ES Systems Test" - Rev. 0
1304.22, " Pressurizer Level and Temp. Instrumentation Surv.
Test" - Rev. 1
1304.24, " Reactor Building Sump Level LT 1405" - Rev.1
General Operating Procedures
c.
As discussed in R0 Report No. 50-313/74-6, Details I, paragraph 4.a.,
the inspector had comments outstanding on two of the 1102 series
'T
procedures. Temporary changes were made to them to incorporate
those comments. Changes to 1102.01, " Plant Preheatup and Precritical
Checklist" (Rev.1) and 1102.08, " Approach to Criticality" (Rev. 2)
were reviewed and the inspector stated that he had no further
questions.
26. Test Results Reviewed
.
TP 600.03, " Soluble Poison Concentration Control Test"
a.
This test was witnessed by the inspector during hot functional
testing (HFT) and findings were documented in R0 Report No.
50-313/74-4, Details I, paragraph 2.
Results of the test were
endorsed by AP&L April 24, 1974. The purpose of the test was
to verify the ability of the boric acid change mechanism to
raise / lower boron concentration in the reactor coolant system
(RCS) and to obtain and maintain specified co'ncentrations. Also,
the calculational methods of OP 1103.04, " Soluble Poison Concentra-
tion Control Operating Procedures," were to be compared with
measured results and adjusted if necessary. Acceptance criteria
,
l
l
involved successful completion of these two purposes.
Several tests were conducted, including boration by feed and bleed,
boration by batch feed, and deboration by lead and bleed and by
batch feed (See Sections 9.1 and 9.2 of t'.v FSAR) . In addition,
one purification demineralizer was borar;d. Tests involving
boration indicated that the formulas u .d in 1103.04 will require
s
,
'
.
-,7----
,. - - - -
,_ --,
.
.
.
.
'
m
-
(n
<
.
.
'
s'
RO Rpt. No. 50-313/74-9
I-20
no alteration after the core is loaded. The final RCS concentra-
tion were slightly lower then predicted, and estimates show that
the reduction in the RCS volume by insertion of fuel elements will
give very good results. Deboration results revealed that final
concentrations were lower than expected, requiring a reduction
in the amount of condensate to be added as calculated by 1103.04.
OP 1103.04 has been revised to reflect this correction factor.
Two problems were encountered during the test. When the batch
controller is shutdown either automatically or manually, the boric
acid pumps discharge valves close automatically, and the pumps
,
are dead-headed unless stopped by the operator. This fact has been
'
highlighted in a revision to the procedure. On one occasion, two
valves were found to be misaligned. The procedure was revised to
caution operators to verify correct valve lineups each time an
evolution is started.
The inspector stated that he had no comments or questions on this
test.
b.
TP 600.32, " Post HFr Internals Inspection"
m)
.
This procedure was used to conduct an inspection of the reactor
vessel and its internals following HFT.in accordance with RG 1.20,
" Vibration Measurements on Reactor Internals." In that the ANO 1
design is identical to the B&W 177 fuel assembly plant prototype,
AP&L elected to perform this inspection rather than perform a
vibration measurement program per Section B of RG 1.20.
Specific
.
areas and components inspected were in accordance with Section D.2
of RG 1.20.
.
No debris, foreign material, or loose parts were found in the
reactor vessel. The only anomaly found in core support and
internals assemblies was a' loose surveillance specimen holder tube
(SSHT). As discussed in paragraph 14, AP&L concluded that this
was ,the source of noises / vibrations observed during HFT. A
B&W field change had been issued to correct the problem.
Numerous readings were taken on the clearances between the *.nermal
shield and its restraint blocks. Some of them did not fr.1 into
the range of 005 to .008 inches specified in the procedure. This
matter was under review by B&W. This problem alcag with the
loose SSHT were being carried as test deficiencies. The test
received an interim endorsement April 20, 1974.
In that AP&L had
classified these problems as restraints to fuel loading, the
inspector stated he.had no comments,
v
.
.
.
m
_ _
._ -
.
_ __
__
.
_
_ _ _
.
-
-
.
,
'
r3
,
s
'
.
,
,
RO Rpt. No. 50-313/74-9
I-21
)
i
,
.
.
f
c.
TP 420.01. " Communications System Test"
l
This test .was conducted during HFr. Audibility of.the plant alarm
4
and page systems were indicated on plant layout drawings. .The
results showed that the evacuation alarm could not be heard in
several locations. Numerous extra speakers were installed and ,a
second test was successfully conducted May 11, 1974, to assure
'
that the alarms could be heard in all areas necessary for fuel
,
loading at noise levels expected to be prevalent at that time.
.
A repeat of the test is to be conducted after fuel loading when
all equipment is in normal operation again. The inspector had
no outstanding comments on the results of this test.
j
d.
Control Room Fire Control System
i
.
!
In response to questions from DL regarding the use of polyvinyl
chloride (PVC) as a covering for flexible steel conduits in the
!
control room subfloor, the licensee stated in letters of
!
supplemental information to the license application dated August 1,
!
1973, and November 30, 1973, that a "Halon" fire suppression system
-s
would.be installed. This system would automatically release "Halon"
j
s
,
i
gas into the subfloor upon detection of a fire by fire and smoke
i
detectors. The systems installed was a Cardox high flow Halon 1301
by the Cardox Division of Chemtron Corporation.
During the ' current inspection, licensee personnel informed the
,
'
inspector that part of this system had been disabled during
testing. The introduction of approximately 1200 psig CO2 into
.
this 600 psig system had disabled a check valve on the redundant
,
"Halon" bottle bank train, this leaving the system with a one shot
capability rather than the redundancy required by insurance. Licensee
, :
!
personnel stated that the train not damaged had been tested
!
successfully and that a replacement check valve had been ordered and
'
would be installed as soon as it arrived, probably the week of
May 20,1974. The inspector asked that the test procedure for
'
this system be made available for his review and licensee personnel
!
stated that the only procedure used was one which Cardox had
i
provided. The inspector stated this appeared to be in violation of
l
Section 13 (Page 13-1) of ehe FSAR which states that tests will be
l
conducted using written, approved test procedures and defines the
'
nature of the test procedures.
It also states that the AEC
publication " Guide for the Planning of Preoperational Test Programs"
,
would be complied with.Section V.H of the appendix to the Guide
incorporates fire protection systems.
.
4
0
.:
!
.
-
-
.
-
. . . . . .
_ _ -
.
. .
.
-
.
-
- .
.
.
._,. -
.
.
.
.
.
.
f~' ,
m.
'
m
~
'
V
.
RO Rpt. No. 50-313/74-9
I-22
Licensee personnel agreed that a procedure for the test should
have been reviewed and approved by AP&L, but felt that this was
an isolated case,
Other Test Results Containing Deficiencies
e.
All test results packages which contained exceptions or deficiencies
.such that they had not been finally endorsed were reviewed to determine
that the systems required for fuel loading had been demonstrated
to be operable. Those which do not affect fuel loading but will be
required to be completed prior to initial criticality are given
below.
Examples of the types of problems to be corrected are.also
givea.
(1) TP 201.03 . " Core Flood System Functional Test"
Flow paths for the tank vents to the vent plenum and a drain line
to the clean liquid waste system had been verified. The radwaste
systems had not been completed at the time this test was run.
(2) 330.04
"CRD Integrated Test"
As_l
~
Several in-limit lights did not actuate properly during HFT. These
are to be repaired and retested after the vessel head and service
structure are installed after fuel loading.
(3) Others Containing Deficiencies
.
TP 231.69
" Dirty Radwaste System Test"
TP 234.01
" Resin Sluicing Test"
TP 273.36
" Auxiliary and E:nergency W"
TP 276.44
" Condensate System Test"
TP 370.01
" Reactor Building Hydrogen Re'moval System '
TP 500.03
" Initial Radiochemistry Test"
TP 600.14
" Pipe / Components Hot Defle.: f on Test"
27.
Service Water Pumo Motor Housing Crack
On April 24, 1974, the inspector received a verbal report regarding cracks
which had beer found in a structural member of the service water pump
p.
)
PM4A lower notor base (end bell) . A written report per 10 CFR 50.55(e),
(V
.
.
_ ,
. - . . . , .
._
_
_.
_ _ . . . _ .
d
.
--
A
N
,
,
,
.
RO Rpt. No. 50-313/74-9
I-23
.
l
dated May 10, 1974, attributes the cause of cracking to be rough handling
under alignment operations. The report also addresses safety implications
-
and corrective action taken. The defective end bell was replaced prior
to the conclusion of the inspection. The inspector stated that he had
'
no questions on this matter.
-
28. Paralleling of Vital Busses A3 and A4
On May 15,1974, a verbal report was received per 10 CFR 50.55(e)
'
regarding an inadvertent paralleling of vital buses A3 and A4 (See
FSAR Figure 8-1) . - The incident occurred during the conduct of TP 310.03,
" Integrated Engineered Safeguards Test," on May 11, 1974, and aas caused
j
j
by an error on a schematic drawing which had been revised as part of a
design change to allow automatic restart of the service water pumps
j
following loss of voltage without a concurrent ES signal. The inspector
i
{
noted that the revised drawing (by Bechtel) had received the review
required by the AP&L QA manual. The drawing error was caugitt by a
'
subsequent review by AP&L personnel. The error was correcued and the
test was successfully rerun prior to the conclusion of the inspection.
A written report was submitted May 16, 1974. The inspector stated that
he had no questions on this matter.
-
4
.
!
.
l
i
I
.
l
.
' O
.
T
'
'
- _ . . , , . _
. , , _ . , , , . .
,
_.
_ , , _ , , _ ,_
' * " -
--
.
_
_
_
_
.
_ _
-
.
.
.
..
.
,.
~
v
.
.
,
R0 Rpt.'No. 50-313/74-9
II-1
)
.
e
DETAILS II
Prepared by:
d (.4 %
- Shs-7 y
t
F. Ja'pe
47 /-
Date
Reactor Inspector
Facilities Test and
Startup Branch
!
Dates of Inspection: April 24-26, 1974
,
Reviewed by: [. C . bM
7/M/7h
'
R. C. Lewis
Date
Acting Chief
.
Facilitiee Test and
Startup Branch
,
.
1.
Individuals Contacted
,
J. W. Anderson - Plant Superintendent
R. R. Culp - Test Administrator
J. L. Orlicek - Quality Control Engineer
i
'
2.
General
.
I
The inspection was conducted to review preoperational test results.
l
Examination of eight completed test procedures by the inspector
'
revealed one general comment which has to do with revising the FSAR
whenever a limit or condition is accepted as a result of a test and
is different from that presented in the FSAR.
Licensee management stated that the FSAR would be revised as soon as
practicable to include changes as a result of testing. A brief
discussion of each test procedure examined by the inspector is
presented in the parascaphs below.
.
'
3.
This test was run on August 10, 1973, and was endorsed by the Test
Working Group (TWG) on March 29, 1974. Following completion of the
i
decay heat cross over piping modification, the hydro test was rerun
on November 25, 1973, to test the newly added section of piping. The
TWG endorsed these test results on March 29,,1974.
i-
,/
t
.
1
[
. . .,:
.
,
. .
.
.
.
.
.
,
,
__
. . . ..
.
.
~
,,.
.
.
.
_,
.
.
R0 Rpt. No. 50-313/74-9
II-2
'
During both runs of the test, a pressure of 3125 psig was held for
at least 10 minutes then reduced to 2500 psig and held during the
,
inapection for leaks. - A representative from the Kemper Insurance
Company witnessed the test and concurred that the test results met
the acceptance criteria.
The test pressure was 1.25 times system design pressure and was in
agreement with ANSI.B31.7-1969, " Nuclear Power Piping," Chapter I-VI,
" Examination and Test."
4._
TP 203.03, " Decay Heat System Functional Test"
The primary purpose of this test was to demonstrate the operability
l
of the decay heat system using related operating procedures.. The
test was started on January 27, 1973, and completed on January 15, 1974.
Final endorsement of test results by the TWG was obtained on
February 8, 1974. All flow paths in the decay heat system were
verified which meets the acceptance criteria. The instrumentation
and controls related to this system were determined to be operable.
i
'
5.
TP 201.01 " Core Flooding System Hydrostatic Test"
Both core flooding tanks were hydro tested on 1000 psig. The test
was completed on July 27, 1973, and the endorsement that all
i
acceptance criteria were met was obtained on September 19, 1973,
from the TWG. No deficiencies or discrepancies were noted. The
test was conducted as required by ANSI B31.7-1969, " Nuclear Power
Piping," which is in agreement with Table 6-2 of the FSAR.
'
.
I
6.* TP 201.02, " Core Flooding System Valves Electrical Test"
The primary purpose of this test was to demonstrate . proper operation
of electrically operated core flooding valves. The test was initiated
on January 10, 1973, and completed on January 16, 1973. The operation
,
of each valve and its position indicating light was verified which
t
satisfied the acceptance criteria. Valve opening and closing cycle
times were measured on required valves and 11 were within acceptance
limits. The torque switches on CV-2416, 2417, and 2420 were faulty
'and all were replaced with new-torque switches.
l
O
-
-
.--
..
.
.
.
,.
.
.
.
..
.
J
RO Rpt. No. 50-313/74-9
II-3
7.
TP 23.07, " Decay Heat Removal System Engineered Safeguards Test"
This test was performed to demonstrate the capabill.ty of the decay
heat removal system.- The test was approved by the PSC on October 24,
'
1972; SRC on December 4,1972; and the TWG on January 3,1973. The
test was conducted on November 1, 1973. Endorsement of the test,
indicating that all-acceptance criteria were. met, was given by the
.
TWG on March 20, 1974.
!
Evaluation of the test data indicated that the minimum required flow
of 3000 gpm was attained. Loop "A" flow was determined to be
3150 gpm and Loop "B" was 3197 gym. These data are within the
values given in Table 9-10 of the FSAR.
8.
TP 202.03, " Makeup and Purification Functional Test"
i
Test of the makeup and purification system began on May 3, 1973,
and was completed on January 5, 1974. All flow paths in the system
,
+
were verified and the alarms and interlocks were checked. The test
was approved by the Plant Review Committee and the Safety Review
Committee on March 19, 1973. Endorsement of the test results,-
indicating that all acceptance criteria were met, was obtained
on April 10, 1974, from the TWG.
'
This test incorporated the design features of the makeup and
purification system as described in Section 9.1 of the FSAR.
The important features of this system are to provide reactor
coolant pump seal water and to provide a means to adjust the
boric acid concentration of the reactor coolant. These
'
features were demonstrated.
9.
TP 230.68, " Clean Radwaste System"
The objective of this test was to demonstrate the ability of
the liquid radwaste system to collect, store and process clean
waste streams from the plant. The test was approved by the Plant
Safety Committee on February 8, 1973, the Safety Review Committee
on March 9,1973, and the Test Ucrking Group on March 28, 1973.
Testing ' began on February 6,1974, and was completed on
February 28, 1974.
'
.
.
,
_
_ _ .
._..y
..
. _ - _
_ . . - .
.
.
.
~.
.
.
-
, ~~
(s_,/
RO Rpt. No. 50-313/74-9
II-4
-
.
All acceptance criteria were met except for the minimum flow of the
treated waste monitor pumps, P47A and B.
The -acceptance criteria
for these pumps was a minimum flow of 100 gpm at 31 psig discharge
pressure. Actual flow was 85 gpm at 31 psig discharge pressure.
This item was reviewed by AP&L and Bechtel and it was concluded
that the 85 gpm was an acceptable flow rate and no further action
would be. required. Final endorsement of the test results was obtained
on March 27, 1974, by the TWG. Table 11-6 in the FSAR which lists the
, treated waste monitor pump capaicty was revised May 1,1974, to a capacity
of 85 gpm which agrees with the test results.
10. TP 500.01, " Reactor Coolant System Chemistry Test"
!
This test was approved by the Plant Safety Committee on March 15, 1973,
and the TWG on March 2.8, 1973. Testing was started on May 1, 1973, and
completed on March 6, 1974. The test demonstrated the ability to
establish proper water chemistry during various plant conditions.
The acceptance criteria for water chemistry agreed with values
presented in Table 4-10 of the FSAR. Results of the test showed
'
that water quality limits can be met.
- i
Final endorsement of the test results was obtained,from the TWG on
s
April 17, 1974.
.
I
e
I
'
,
. - .
-.
.
,
_.
.
__ _ _
_ _ _ _
~
,
.
.
'
-
..
m.
,
.
.
.
/~
.
(
.
RO Rpt. No. 50-313/74-9
III-1
DETAILS III
Prepared by:
[
ofn#
8 0
W. W. Peery, Rad %$ lion Specialist
/ Date
Radiological and Environmental
Protection Branch
l
,
l
Dates of Inspection: May 1-2, 1974
[d
'%
Reviewed by:
/
a
,
,47 T. Sutherland,~ Chief '
/ Datef
Radiological and Environmental
Protection Branch
,
l
'
1.
Individuals Contacted
J. W. Anderson - Plant Superintendent
'
G. H. Miller - Assistant Plant Superintendent
C. A. Halbert - Technical Support Engineer
'
J. L. Orlicek - Quality Control Engineer
T. C. Baker - Chemical and Radiation Protection Engineer
'
R. G. Carroll - Otemical and Radiation Protection Engineer
2.
Respiratory Protection Program
This unresolved item (73-10/6) was previously discussed in RO Report
,
Nos. 50-313/73-10, 50-313/73-18, and 50-313/74-6, Details II,
- paragraph 2.' - Training in the use of respirators is to be completed
on May 3, 1974. The respirators have been in use in the training,
therefore, they have not had a final cleaning', packaging and storage.
Storage shelves are available in a storage room near the health
.,
physics office, however, the shelves and room had not been cleaned
and made completely ready for the masks. Management verified with
i
staff members present that this-item will be entirely completed by
the week of May 13, 1974.
.
3.
Representative Sampling of Gaseous Wastes
l
This unresolved item was previously discussed in RO Report Nos.
50-313/73-10, 50-313/73-18, and 50-313/74-6, Details II, paragraph 3.
The inspection of gaseous sample delivery lines outside buildings
- revealed that insulation had been installed on the lines to minimize
sample losses due to condensation. This unresolved item is closed.
4.
Emergency Planning
This unresolved item was previously discussed in RO Report No.
50-313/73-18, and 50-313/74-6, Details II, paragraph 4.
As
,
.
4 M
-
~
,,., ,-- ,....--
e
.
,.,y,
.+,,.,4.
,e-,e-,
m
- - - . - - -
,g
w
-
.
.
-
, - ,
,
.
,
.
R0 Rpt. No. 50-313/74-9
III-2
a result of inadequacies revealed by Plant Safety Committee review
of a drill held on January 11, 1974, additional drills have been
held which corrected the inadequacies. A continuous telephone
cable had been provided from the plant to the emergency control
center but the telephone had not been installed. A management repre-
sentative stated that the telephone should be installed by May 3,1974.
Electric power is available in the building and lights are in
operation. A radio was available but it had not been connected to
the power source. Batteries were in place as an auxiliary power
supply but they had not been connected to the line supply for
charging. This was also expected by management to be completed
by May 3, 1974. Two emergency kits were provided in the emergency
control center,
In addition to protective clothing provided in the
kits, individually packaged protective clothing was stored in the
center.
Inspection of the contents of the emergency kits revealed
that the detector cable was not provided for one survey instrument
and another survey instrument did not appear to be operating properly.
Management agreed that the problems with the survey instruments
will be corrected promptly. It was also pointed out that the low
volume air sampler should have a screen backing for the filter
's
paper and a response check source provided for the survey instru-
)
ments. A management representative indicated that this would be
accomplished.
5.
Calibration of Radiation Monitors
This unresolved item was previously discussed in R0 Report No.
50-313/73-18 and 50-313/74-6, Details II, paragraph 5.
It was
-
determined that process monitors remaining to be calibrated were
as follows:
.
RE-3814 - Service Water 1
No detector
RE-4642 - Liquid Radwaste
No detector
R-2120 - Penetration Rm. Vent Sampler 7
R-2130 - Penetration Rm. Vent Sampler /
In progress,
R-7441 - Hydrogen Purge System Monitor
3 complete
R-7442 - Hydrogen Purge System Nbnitor
1 remaining
Licensee representatives determined that the replacement detectors
had been shipped and they were expected to be received by May 3,1974.
Management assigned specific responsibility for followup to complete
all repairs and calibrations as soon as possible with completion
expected no later than the week of May 6, 1974.
Area monitors not calibrated were as follows:
l
l
3
RE-8002
Relay Room
Faulty Calibration Pot
!
,/
RE-8006
dadio-Chem Lab.
Faulty Calibration Pot
,
s
.
-
.
.
.
.
~
.
p.
.
.
.
(
!
-
'- #
RO Rpt. No. 50-313/74-9
III-3
.
RE-8012
Piping Area I
Faulty Calibration Pot
RE-8017
Fuel Handling Equipment Faulty Calibration Pot
A licensee representative stated that replacement parts are on order
but the time of receipt was unkncwn. Management also emphasized
assignment of specific responsibility for followup and completion of
this item as soon as possible.
It was also pointed out by the inspector
that the calibration of RE-8015, Condensate Demineralizer and RE-8019,
Incore Instrument Tank revealed them to be barely within specifi-
cations. A licensee representative stated dhat replacement parts
ordered would be adequate to replace parts on these two instruments
to achieve recalibrations that better meet specifications.
6.
Radioactive Waste Systems
B
i
a.
Provision has not been made for spot check sampling of. normally
uncontaminated waste streams. This will be pursued post
operatively in terms of adequacy of determination of releases
to unrestricted areas.
e- s
b.
Upon completion of calibration of process monitors the ability
j
)
to detect anticipated concentrations should exist.
c.
Operating limits related to the use of treatment systems have
not been developed. A licensee representative stated that
this will be accomplished post operatively, after parameters
based on capabilities are established.
.
.
d.
Equipment associated with waste systems is operable with the
exception of resin waste system which is to be completed after
operation.
(See R0 Report No. 50-313/7)
e.
A licensee representativo stated that the efficiency and calibration
of sample collection media will be accomplished post operatively
in accordance with AEC Guides.
f.
A licensee represents 've stated that correlation will be made
between concentration at the sample probe and those indicated
at the monitor post operatively when activities have been generated.
g.
The noble gas monitor has been calibrated.
h.
The solid waste disposal systems are complete with the exception
,
l
of the spent resin system. This item will be completed post
'
operatively.
(R0 Report No. 50-313/74-4.) In the handling of
solid wastes, provision has been made for personnel exposure
fN
control, adequ-te storage, administrative limits to estimate curie
I
)
contene and es prevent overloading containers.
\\_./
~
.
.
.
.
-
,
'
.
>
,
RO Rpt. No. 50-313/74-9
IV-1
DETAILS IV
Prepared By:
.
(A
I-h-7[
.
R. C. Parker, Reactor Inspector
Date
Facilities Test and Startup Branch
LbN
h
W74
L W. Whitt, Reactor Inspector
Date
Facilities Test and Starcup Branch
Dates of Inspection: April 24-25, 1974
Reviewed By: [ (,
e M .)
M
R. C. Lewis, Acting Chief
Date
Facilities Test and Startup Branch
.
1.
Individuals contacted
Arkansas Power and Light Company (AP&L)
-
J. W. Anderson - Plant Superintendent
g
B. A. Terwillinger - Operations Supervisor
,
J. A. Orlicek - Quality Control Engineer
N. A. Moore - Manager of' Quality Assurance
H. Hollis - Administrative Assistant
2.
Discussion
.
This portion of the inspection was performed to determine if
deficiencies related to the implementation of the operational QA
progra:n had been corrected. Deficient areas inspected are identified
in RO Rpt. No. 50-313/74-5.
3.
Procedure Review
Resolution of the QA audit deficiencies was accomplished primarily
through implementing procedures. This required major changes or
complete revisions of many procedures as well as the development of
several additional procedures. The implementing procedures reviewed
during this inspection are listed numerically below:
a.
Quality Assurance Procedure (QAP) - ANO-14, " Operating Plant
General Audit"
b.
Quality Control Procedure (QCP) 1004.01, " Design Control"
c.
QCP1004.02, " Initiation and Processing of Trouble Tickets"
-
.
.
m
S
'
.
,
,
,
,
'
i
RO Rpt. No. 50-313/74-9
IV-2
N
d.
QCP1004.03, " Quality Control Training and Indoctrination"
e.
QCP 1004.04, " Turnover of QA Documentation from Construction
to AP&L"
f.
QCP 1004.05, " Purchase Requisition Preparation and Processing"
,
g.
QCP 1004.06, " Material Receiving and Inspection"
h.
QCP 1004.07, " Control of Special Processes"
1.
QCP 1004.08, "QC Inspections"
j. QCP 1004.10, " Calibration Control"
k.
QCP 1004.11, " Handling, Storage, and Shipping of Q-List Materials"
1.
QCP 1004.12, " Operational Test Control"
~
%
m.
QCP 1004.13, "Nonconformance and Corrective Action"
i
n.
QCP 1004.16, " Review and Verification of Documents"
o.
QCP 1004.17, "Onsite Fabrication and Modification Control"
p.
QCP 1004.18, " Material Identification"
.
q.
QCP 1004.19, " Hold, Caution, and QC Tagging Procedure"
QCP 1004.20, " Qualification and Certification of Quality
r.
Control Personnel"
s.
Administration Procedure (AP) 1005.01, " Administrative Controls
Manual"
.
t.
AP 1005.02, " Handling of Procedures"
u.
AP 1005.03, " Document Control"
v.
AP 1005.04, " Control and Use of Bypasses and Jumpers"
w.
AP 1005.09, " Plant Records Management"
x.
AP 1005.10, " Document Retention and Disposition"
.
.
-
-
_
. _ _ _ _ . _ _ . _ ._ __.
__ __ _
_ _ _ _
_
_ _ _ _ .
- _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _
_
-
.,
.
)
.
.,
'
r,
,
,
,
.
l
R0 Rpt. No. 50-313/74-9
IV-3
.
!
n
4.
During the inspection of April 24-25, 1974, certain deficiencies
l
were still found to exist. These were brought to the attention of
l
and discussed with AP&L personnel. These remaining deficiencies
were subsequently reinspected and closed out during the May 13 - 17
i
1974, inspection. Unresolved items 74-5/1 through 18 are
considered to be resolved.
t
.
!
l
-
,
e
l
b
,
-
.
(
!
q
}
i
.
.
!
l
i
e
\\
-
i
l
l
l
l
l
..
f
~ - - '
- ~ ^ *
.
-
.
. . .
. .
w'-dF
.
__,,_,p
, _ _ _ __
.-
.
'
.
-,
G
-
>
.
.
R0 Rpt. No. 50-313/74-9
V-1
.
,
DETAILS V
Prepared by: A. M b A
4/s-/74/
D. J.pBurk'e, Reactor Inspector
/Date
Facilities Test and Startup Branch
Dates of Inspection: May 13-17, 1974
Reviewed by: 8. 6.
M
R. C. Lewis, Acting Branch Chief
Date
Paci11 ties Test and Startup Branch
i
1.
Individuals Contacted
i
,
Arkansas Power and. Light Company (AP&L)
i
B. A. Terwilliger - Operations Supervisor
J. A. Orlicek - Quality Control Engineer
.
R. R. Culp - Test Administrator
M.. H. Shanbhag - Procedure Administrator
2.
Procedure Audit
The inspector compared the licensen
procedures to those required
,
by Reg. Guide 1.33 and found no discrepancies. The licensee has identi-
i
, fied and written procedures to cover the activities identified in 1.33.
The inspector also audited AP&L's Operating Procedures (OP) by
comparing the master file with the current OP index and by spot-checking
-
1
OP's for the proper reviews and approvals. The spot-checks uncovered
.
no instances of OP's which were improperly reviewed or not approved and
signed-o ff. However, the following procedures, which are listed in the
'
'
procedure index, could not be found in the master file of OP's:
1303_100, " Nuclear Data 4410 Canna Spectrometer Calibration"
l
1304.16, " Steam Line Break Instrumentation and Control"
1304.29, " Calibration Procedures for itydrogen Analyzers"
1104.85, " Pipe Hanger Settings"
1304.86, " Liner Plate Surveillance Test"
1304.89, " Diesel Generator Protective Relaying, Starting, Interlocks
and Circuitry"
1304.90, "Off-Site Power Undervoltage and Protective Relaying Interlocks
and Circuitry"
1401.18, "CRDM Axial Power Shaping Rod Installation"
1405.01, " Station Batteries and Switchyard Batteries"
a.
1303.100,1304.16,1304.29, -1304.85 and 1304.86 are calibration
and/or periodic surveillance type tests which are not required
.
J
.
-
-
~
.
- - - - , - ,
, , ,
- . - , - , - - , , . , ,
.-,..nn,,,,---
,.,
._
_
__
_ _ .
'
.
,
5
--;
.3
.,
^
,
,
.
.
RO Rpt. No. 50-313/74-9
V-2
for the. initial operation of the plant. They are in the process
of being written and approval is expected before criticality.
,
i
b.
1304.89,1304.90, and 1405.01 are surveillance tests that are re-
quired before the issuance of an Operating License. Consequently,
'
these three procedures were completed, properly reviewed, and
approved during the week of the inspection.
c.
1401.18 has been written, but requires the appropriate reviews
~
'
and approvals. It is not needed before refueling.
.
3. .Annuce.iator Alarm Procedures
The. inspector reviewed some of the several hundred annunciator alarm
procedures included in 1203.12 " Annunciator Alarms." There were no
j
unresolved questions on the individual alarms and procedures, but the
inspector stated that the preface for' the entire voluminous procedure
'
appeared weak.
.
,.
The operation of the panels, alarms, and acknowledge buttons is not
described in. this (or any other) procedure. The inspector also
~
The licensee
questioned how the panels are verified to be operational.
j
stated that he will initiate a temporary change to 1203.12 to provide
/
a description of the alarm system. In addition, the licensee will
l
provide a description of the system "off-normal" alarms when their
design and installation is complete.
1
- The inspector noted that there was little mention of administrative
'
controls or guidelines in the procedure for the operators, but the
licensee maintained that most of this would come from the training
!
The licensee stated that he will document additional
program.
guidance and control if it becomes necessary. The inspector had no
further questions.
4.
hit Procedure Review
'
. The_ inspector reviewed several of the Test Proced'ures (TP) with out-
standing deficiencies to assure that these items would not delay the
!
!
issuance of the operating license. The deficiencies were either
resolved or required to be resolved before criticality. The items
l
requiring resolution before initial criticality are listed in the Region II
- .
letter to Licensing recommending issuance of the ANO-1 operating license.
i
The review included the following TP's:
l-
165.01, " Filter Test"
a..
i
'
172.01, " Penetration Room Ventilation"
b. . 200.06, "RC Pump and Motor Initial Operation"
c.
1
,
_ - .
-
--
- ,_
,. _ _ _ _ _
, _ . . . - _ _ -
.
. . .
.
.
.
.
. ~
.
.
m
- ,
<
..
.
,
O
'
RO Rpt. No. 50-313/74-9
V-3
,
d.
200.11, "RC Pump Flow Test"
200.12, "RC Pump Flow Coastdown Test"
e.
210.03, " Chemical Addition and Sampling System Functional Test"
f.
240.14, " Intermediate Cooling Water System Pre-Operational Test"
g.
h.
266.11, " Service Water Pre-op Test"
320.02, "ICS Open Loop Calibration"
1.
351.30, " Computer and Control Room Ventilation Pre-op"
j.
K.
400.02, "DC Power System Pre-op"
600.13, " Pressurizer Operational and Spray Test"
,
1.
600.17, " Control Rod Drive Operational Test"
m.
5.
system Inspection
The inspector " walked-down" one of the safety-related systems with
licensee personnel to verify that the system matches the description
in the FSAR. The emergency core cooling legs of the high and low
pressure injaction systems were inspected. The inspector verified that
the FSAR does, in fact, describe the system as built in the field.
.
s
9
4
4
9
6
.
I
.
~ ^ ~ ^
- - " " - - -
.
-
.
_
_.
--
. .
__
-- .._ - . - . . . _ _ . _ . _ __. -_.
,
l
1 -
'
r
7m
s
t
-
.
.
.
4
Arkansas Power and Light Company
ANO-1
,
!
RO Inspection Report No. 50-313/74-9
,
cc w/encla
H. D. ThoJnburg, X
,
4
RO HQ (sn
ws
DR Central Files
Regulatory Standards (3)
,
Directorate of Licensing (13)
i
ec enc 1. only:
- PDR
- Local PDR
i
- NSIC
'
J
'
- DTIE, OR
l
- State
,
M
'
- To be dispatched at a later date
.
< :
e
l
[
4
o
4
!
l
'
. . . . . . , , ,
. _ _ _ ,_.
. . . _ - _ - - - - - . -
. _ , _ . - .
...
-
.
..
. - -
.-
. - . .
. . - . .
- .- .-. - . -