IR 05000313/1973013
| ML19319E559 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 09/20/1973 |
| From: | Browlee V, Crossman W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19319E552 | List: |
| References | |
| 50-313-73-13, NUDOCS 8004110816 | |
| Download: ML19319E559 (9) | |
Text
'O
~
-
.
.
o
,
v
%
UNITED STATES k
ATOMIC ENERGY COMMISSION
o l
17C<*<
DIRECTORATE OF REGULATORY CPERATICNS
-
'
j g$ f
- - '
meosou n - suite ate
)
g 23018E ACHT m E E ST m E ET, NomT HwCSP
- farts O'
AT t. A NT A. GE emo s A 30303
,
,
g,
.
.
RO Inspection Report No. 50-313/73-13 Licensee: Arkansas Power and Light Company Sixth and Pine Streets
,
Pine Bluff, Arkansas 71601
Facility Name: Arkansas Nuclear One, Ur.it 1 Docket No.:
50-313
,
License No. :
CPPR-57 i
Category:
A3/B1
incation: Russellville, Arkansas p\\
Type of Licensee: B&W, PWR-2568 Mwt, 880 Mwe
- O
.
Type of Inspection: Special, Unannounced, Construction Dates of Inspection: September 5-7, 1973
'
,-.
Dates of Previous Inspection: August 7-10, 1973
.
Principal Inspector: I
- i ' "I.'
3/
V.'L. Brownlee, Reactor Inspector
'Date
.
Facilities Section, Facilities Construction Branch
.
i j
Accompanying Inspectors: lbne i
Other Accompanying Personnel: None Reviewed by:
.3
% __
W. A. Crossman, Senior Inspector, Facilities Section Date
.
l Facilities Construction Branch i
n 0h$
800411 e
l i
..
.
..
_
. _ _
- _ - ___
_ _ _ _ _ _ _
.
.
-
,
__
._
_.m.
_
_-_
_
_ _
.
,,,
\\
2-RO Rpt. No. 50-313/73-13
-
-
SUMMARY OF FINDINGS
.
I.
Enforcement Action
A.
Violations None j
B.
Safety Items None II.
Licensee Action on Previously Identified Enforcement Matters
l A.
Violations 73/11-A1(II) Control of Special Processes AP&L has been responsive to this item. This item is closed.
(Details I, paragraph 5)
B.
Safety Items
,
None
.
'
III. New Unresolved Items None
.
'
IV.. Status of Previously Reported Unresolved Items
.
72-4/2 Cable Routing (Region II Letter, July 27, 1972, Item 2)
The routing of, cables under the control room removable
,
floor may not meet safeguards and reactor protection systems separations and loading criteria.
AP&L has been responsive to this area of concern in their j
letters of May 11, 1973, and August 1, 1973, to the Directorate of Licensing. This item is closed.
(Details I, paragraph 7)
.
J.
I k
,-
-
. -.
,
,...,
- 1,
. -.
.
__ _ _- _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _
-
'
.
___s
.
_.
.u
- - - -
_
.
,,
%
%
(v)
RO Rpt. No. 50-313/73-13-3-
-
72-4/3 Cable Routing (Region II Letter, July 27, 1972, Item 2)
.I This item was resolved in RO Report No. 50-313/73-9, Details I, paragraph 9, in conjunction with item 72-11/2; however, it was not removed from this section of the report at that time. RO:II has no further questions regarding this item.
72-7/1 Pressurizer Saf**y and Relief Valve Mounting and Connecting Piping (Region 1: Tetter, August 21, 1972, Item 1)
Bechtel has completed a stress analysis of the discharge piping and concluded that additional restraints are required.
This item is closed.
(Details I, paragraph 6)
72-11/3 Cable Installation in control Room and Computer Room False Floor and Floor of Main Control Panels (Region II Letter, December 19, 1972, Item 3) (See Item 72-4/2)
It is our understanding that the licensee'will require gsg (
)
Bechtel to evaluate cable installation under the control
\\- '
room and computer room false floor regarding separations, loading and quality workmanship criteria.
,
j AP&L has been responsive to this area of concern in their letters of May 11, 1973 and August 1, 1973 to Directorate of Licensing. This item is closed.
(Details I, paragraph 7)
j 72-12/2 Valve Wall Thickness Verification (Region II Letters, June 30, 1972, and February 16, 1973)
,
The wall thie'gness verificati' n program should be concluded o
by October 15, 1973.
(Details I, paragraph 8)
73-11/1 Malfunction of. Containment Purge Supply Valve Switch
,
ROB No. 73-2 AP&L has been responsive to this item.
(Details I, paragraph 4) This item is closed.
g V.
Design Changes None
.
.
(
.
-
. __ _
_ _ _ _ _ _ _ _ _ _ _ _ _
. _ - - _ _ _
- -
.
,
- - - -
_ _ _ -
.
,
%
O RO Rpt. No. 50-313/73-13 4-
-
-
_
VI.
Unusual Occurrences
.
None VII. Other Significant Findings
A.
Projec.t Status None B.
Personnel or Organizational Changes None
.
C.
Inquiry Reports None
?
VIII. Management Interview Our findings were discussed at the conclusion of the inspection Le with Mr. Moore, Chief QA Coordinator.
.
.
.
'
.
sh b.
m l
I l
_
_.
.
.
..
. _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _
.
,.
.
--- _
...
.
s
,
l I
RO Rpt. No. 50-313/73-13 L-1
'
DETAILS I Prepared by: f nt Yli /'
k 7_?
'
V. L. Brownica, Reactor Inspector
' Date Facilities Section, Facilities Constrpetion Branch
,
Dates of Inspection: September 5-7, 1973
.
Reviewed by:
%
W. A. Crossman, Senior Inspector Date Facilities Section, Facilities Construction Branch 1.
Individuals Contacted
.
a.
Arkansas Power and Light Company (AP&L)
N. A. Moore - Chief QA Coordinator A. C. Bland - QC Inspector (Civil)
C. L. Bean - QC Inspector (Mechanical)
.
E. Quattlebaum - QC Inspector (Electrical)
b.
Contractor Organizations Bechtel Engineering Corporation (Bechtel)
.
W. J. Stubblefield - Project Superintendent C. G. Beckham - Lead QA Engineer Sabcock and Wilcox Company (B&W)
t
R. M. Moore - Mechanical Engineer (Site)
'
Babcock and Wilcox Construction Company (B &WCC)
E. F. Berdyck - Supervisor, Baseline Inspection Team
'
D. Witt - NDT Technician, PT M. Harley - NDT Tech.dcian, PT T. F. McDermott - NDT Technicien, UT C. L. Mason - NDT Terhnician, UT 2.
Work Summary and Schedule Bechtel has' 244 manual and 25 nonmanual construction personnel working on Unit 1.
The only work remaining on original scope of contract work is the security system. Other work primarily cansists of modification /
j g
rework, assist B&W in baseline inspection, primarry coolant pump repair, painting, insulation, hangar setting and assist AP&L operations as
'
required.
'l
.. _ - -.
.-.
-
_
.
.
,
_... -
-
-
_
--
-
-.
.
(,,,
.R0 Rpt. No. 50-313/73-13 I-2
.
-
Reactor coolant loop cold hydro was completed August 11, 1973, contain-ment air test is scheduled for October 13, 1973, hot functional testing is scheduled for October 15 - November 15, 1973, and reactor vessel
'
baseline inspection is scheduled for November 15 - December 1, 1973.
Emergency core cooling crossover modification is scheduled for completion by September 21, 1973. Reactor coolant pressure boundary valve wall thickness verification is scheduled'to be completed by October 15, 1973.
3.
Insarvice Inspection - Baseline Inspection The reactor coolant pressure boundary inservice inspection program has
'been developed to comply with Section XI of the ASME. B&PV Code, " Inservice Inspection of Nuclear Reactor Coolant Systems," 1971, including 1971 Winter Addenda edition.
The B&WCC NDT personnel were onsite examining shop and field welds within reactor coolant pressure boundary. Emergency core cooling crossover piping modification will be examined during the latter
,
part of October 1973 and the reactor vessel is scheduled to be i
'
examined during late November 1973.
.
,
'
The B&W inspection group consists of a supervisor and five two-man
,
inspection teams. Work commenced during the week of August 26, 1973,
,
'
and was completed during the week of September 10, 1973.
Two Factory Mutual Engineering code inspectors were onsite (E. W. Resell and J. A. Magruder). Magruder was to remain in residence to witness the program implementation.
AP&L has performed an audit of personnel qualifications and continuous j
surveillance of program implementation.
'
,
i
l Several discrepancies were identified by AP&L and immediate corrective measures implemented.
Review of program and discussion with several AP&L and B&W personnel'
identified no areas of concern or significant departures from program
'
requirements which were not being resolved by the applicant.
i 4.
Malfunction of Containment Purge Supply Valve Switch - ROB No. 73-2 AP&L responded in their letter of August 16, 1973. Becht'el has reviewed the design of the control circuit for the containment ventilation system isolation valves and has determined that all reactor building isolation valves in the reactor building purge
'
and hydrogen purge systems are equipped with individual hand f
rwitches. They conclude that single failure of one of these switches
\\s_,
does not prevent isolation of the reactor building, using the redundant
~
valve. Examination of control panels in the control room confirms installation of dual switches.
l
.
I
_ _ _ _ _, - - - _
.
-
.
L
.: -
,-
.-
_ _ _ _ _ _ _ _ _ _ _ _ _ -
_ _ - _ _ _
___
. - - _
,.
-
-
~~-
-
.
. -
-m
..
M i
\\
V RO Rpt. No. 50-313/73-13 I-3 a
~
5.
Control of Special Processes - Valve Wall Thickness Verification Program
'
AP&L's letters of response, August 6,1973, and September 10, 1973, are considered adequate.
RO:II has reviewed the AP&L letter of August 6,1973, " Valve Wall Thickness Verification Program," and concludes that the program adequately addresses the corrective action relating to valve nozzle or weld end preparation measurements.
6.
Pressurizer Safety and Relief Valve Mounting and Connecting Piping Bechtel has completed a stress analysis on the pressurizer discharge piping. The analysis confirmed that the following additional piping restraints were required on the three lines analyzed:
7 hydraulic snubbers, 2 rigid restraints and 1 revision from spring support to rigid support.
7.
Cable Installation in Control Room and Computer Room False Floor and Floor of Main Control Panels
,,_{
AP&L has responded to AEC's concern in their letters to Directorate (
}
of Licensing, May 11, 1973, and August 1,1973.
\\s_/
j Physical installation of safeguards cable in the control room false floor is as shown in Drawing SKE-690, " Control and Computer Rooms S ub-Floors." Cable rerouting was not required to accomplish separa-tion by methods identified in the letters and drawing.
,
Safeguards cabling within the computer room is run in rigid conduit.
8.
Valve Wall Thickness Verification
,
Valve profile drawings have been completed by the field a'nd forssrded to Bechtel, San Francisco. The drawings provide both side and bottom views, identify 0.D. dimensions of both body and run, low spots and lugs. Bechtel engineers have reviewed the drawings, marked locations of measurements to be made and returned marked up drawings to the fie'ld.
Wall thickness verification program is expected to be completed by October 15, 1973.
'
RO:II has made arrangements to review program and data, onsite, during a subsequent inspection.
9.
Significant. Deficiency Report (10 CFR 50.55(e)) Reactor Coolant Pressure Boundary Valves (AF&L Deficiency Report of July 2,1973)
Based on measurements initially performed by Bechtel on valves three G,l inches and larger in the reactor coolant pressure boundary, the following valves were deficient when compared with minimum wall valves in Table
,
_._ _
,.
-
_
--.
.. - _
_.
_ __.-.
-
.
m
-
,
,
V
'
.
.
'
-
RO Rpt. No. 50-313/73-13 I-4 452.1 of draft ASME Code for pumps and valves for nuclear power for
~
1500-pound rated stainless steel valves.
Valve No.
Size (in.)
Type.
Service CF-1A & IB
CK.
Core Flood q
DE-13A & 13B
CK.
Decay Heat DH-14A & 14B
CK.
Decay Heat DE-14A and 14B are the first valves installed on the reactor coolant
,
l pressure boundary where as CF-1A and 1B and DH-13A and 13B are the l
second valves.
'
l
.
j Based on FSAR Table 6-3, the following design conditions apply to l
these valves:
CF-1A and 1B (item 3-b of table 6-3) - 2500 psig at 300 f DE-13A and 13B (item 2-d of table 6-3) - 2500 psig at 3000f
DH-14A and 14B (item 3-a of table 6-3) - 2500 psig at 650 f Utilizing the rules of draf t ASME Code for pumps and valves, Paragraph 452.lb, "Non-Standard Pressure Rated Valve," and using design condi-tions of 2500 psig and 300 f, Bechtel calculated the new pressure rating and the minimum wall required. The pressure rating works out to 1100-pound rating and the minimum wall required is 1.50 inches for 14-inch nominal diameter valve and 1.46 incheh for 12-inch nomi.ial diameter valve. The actual wall thickness measured was 1.60 inches and over. Bechtel concludes that valves CF-1A and 13 and DH-13A and 13B are acceptable.
,
Weld repair is required on DH-14A and 14B. Velan personn'el were onsite during the inspection to restore the valves to design wall thickness.
Velan's onsite QC instructions for welding repair, VEL-QCI-335 included the welding procedure proper, VEL-P-595, " Pressure Code Welding Procedure
-
Stainless Steel P-8 to P-8 Manual Arc" and the necessary procedure qualification tests.
.
.
'
W
.
l t
.
__ -.-...
.-__..
.
_
_ __ -
_
_ _ _ _ _
- - _ _ _ _ _ _ _ _ - _ _ _ _.
-
~
..
<.
'N
.
.
.
Ltr to Arkansas Power and Light Congany fm N. C. Moseley d** SEP 211973 i
.-
DISTRIBUTION:
m s
D. Thornburg, RO a
era.
RO:HQ (4)
Directorate of Licensing (4)
DR Central Files
- PDR l
- Local PDR
- NSIC
- DTIE, OR
- State i
l
- To be dispatched at a later date.
l r
i'
t I
l l
i I
i i
i
,
i
}
!-
._.. _ _, -_... _. _
..__,.,_,_ _ _., _.,. _. _..,., _._
. _ _. _...,,.. _. _ _ _ _.... _ _
_ _, _ _ _ _.. _ _.. _. _
,,._.__..-______J
'