IR 05000278/1974004
| ML20086L483 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 04/11/1974 |
| From: | Allentuck J, Haynes R, Heishman R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20086L448 | List: |
| References | |
| 50-278-74-04, 50-278-74-4, NUDOCS 8402080258 | |
| Download: ML20086L483 (21) | |
Text
__
_ _ _ _ _ - _ _ - _ _ _ - - - -
O
.
O
~
~
.
.
.
.
/
U.S. ATOMIC ENERGY COW!ISSION
,
DIRECTORATE OF REGULATORY OPERATIONS
'
REGION I
'
N-
,
.
RO Inspection Report No:
50-278/74-04 Docket No: 50-278 Licensee:
Philadelphia Electric Company
'
License No:
-
2301 Market ' Street Priority:
Philadelphia, Pennsylvania 19101 Category:
B Location:
Delta, Pennsylvania (Peach Bottom 3)
Typ,e of Licensee: _ BWR 1100 MWe (GE)
.
Type of Inspection: _ Routine. Unannounced
.-
Dates of Inspection:
March 12-15, 1974 Dates of Previous Inspection:
December 26-28, 1973 Reporting Inspect'or:
h./
M /[
V
- '"/I' I[
f
- k. A c, Reactor Inspector Date n
Accompanying Inspect' ors:
A49 [e h "/[ Y P
'
'
' '.
R.Ha'yn/s,Reacto'rInspector
"'
.
Date Date
.
.
Date
<-
-
Date Other Accompanying Pers'o,nnel:
None Date Reviewed By:
.c
- !
'
-
R. F. Heishman, Senior Reactor Inspector
'
D' ate
.
'
i 8402080258 740605
'
l
.
'
PDR ADOCK 05000277
,
S PDR
'
.
i,
,
, - g.,
,y...--.-
i
-.,.w.
..-.~.,.-.y....
o.. e o j
.
..
-
-
- -.
a 2,g, _ _ g a..,,a ;, ; g 1.,
-
_&
_
l O
O
'
~
.
.
.
.
i i
-
~
SUMMARY OF FINDINGS
!
!
l
i
!
Enforcement Action t
i
!
Insulation was applied to a valve prior to the receipt of a Quality Control Release for such application, contrary to Bechtel site proce-
'
dures. This is a violation of 10 CFR 50, Appendix B, Criterion V.
(Details, Paragraph 3)
Licensee Action on previously Identified Enforcement Items
A.
Deviations in the design and manufacture of the wire rope assem-j blies for the Main Steam Line Restraints have been reviewed and
!
approved by the original design group.
i
'
This matter is resolved.
(Details, Paragraph 4)
,
.
B.
'
Prestressing the wire rope assabbly of the Main Steam Line
.
Restraints was apparently not excessive.
This matter is resolved.
(Details, Paragraph 5)
'
C.
The suspect "over-ground" condition at the attachment weld of a
-
pipe support to a pressure pipe has been reinspected and there is no apparent deficiency.
(Details, Paragraph 6)
,
t j
D.
Documentation exists at the site indicating control of the Reactor
- " ',
Vessel Stabilizers as Q-listed components. This matter is resolved.
!
.
(Details, Paragraph 7)
-
E.
Documentation exists at the site indicating re-evaluation by the NSSS of vendor radiographic reader sheets for the RHR heat ex-changer. These sheets indicate that repairs were acceptable. This
!
matter is resolved.
(Details, Paragraph 8)
'
"^ F. ' Required material certification and documentation of prestressing for the wire rope assemblies of the Main Steam Line Restraints
!
exist at the site.
In addition the Constructor has taken action to assure proper receipt inspection in the future.
(Details, Para-
.
graph 9)
-
-
.
,
C.
Weld procedures are entered on hanger sketches.
(Details, Para-graph 10)
-
,
.
Other Significant Findings
'
,
Current Findirgs a.
i
.
e
--r w
- ,--
--,,
w
,y
,
m
,
-.
.
.
o l
O-2-
.
.
.
1.
On the basis of the current inspection of the work and docu-mentation as well as interviews with licensee a-2 constructor personnel, the following items are outstanding:
.
.
a.
The problem associated with the Grinnell snubbers which has led to a decision to replace the orifice blocks, should be reported as a 10 CFR 50.55(c), item.
(Details, Paragraph 11)
v b.
The location on the instrument rack of the IIPCI steam line differential pressure switches apparently does not meet the requirements of the GE recommended specification
'
NEDO 10139, dated June 1970.
(Details, Paragraph 12)
c.
Pipe hangers for 2" and under piping appear not to be
]
double nutted. (Details, Paragraph 13)
d.
The RV drain line contacts the reactor pedestal wall
,
raising the possibility of imposing undesirable stresses
-
on the RV nozzle.
(Details, Paragraph 40)
e.
A vertical portion of the reactor vent and drain system is out of plumb which may lead to a failure to meet the i
requirements of GE spec 22A1295, F2.1.5.2.
(Details, Paragraph 14)
l I
f.
A pipe which shall operate at 5450F and is presently un-insulated is within 12" of CRD control cables.
This item
-
'
,,
will be outstanding.until insulation is applied.
(De-
.
tails, Paragraph 15)
g.
A sampling point in the reactor drain line was installed not in accordance with the Constructor's P&ID.
(Details, Paragraph 16)
h.
The Reactor Water Clean-Up Pump has a plastic plug in a
vent port.
(Details, Paragraph 17)
1.
Attached to the end of the reactor drain pipe is a tem-porary pipe. The inspector was informed as to the system in place to assure that it is properly dispositioned. (De-
"
-
tails, paragraph 18)
-
j.
Two snubbers on the main steam relief valve discharge system had empty fluid accumulators.
(Details, Paragraph
'
19)
,
,
-
,
.
e E
I
..
_,_,._ _
_ - _ _ _ _ _ _ _ _ _
_______
- _ _ _ _ -
_
_ _ _ _ _ _ _ _.
_
O O
.
.
.
_
-3-
.
k.
The standby liquid control relief valves do not have set point shown on nameplates.
(Details, Paragraph 20)
2.
On the basis of the current inspection of the work and docu-mentation as well as interviews with licensde and contructor personnel, the following items indicate no apparent deficiencies:
a.-
The status of the licensee's NCR's.
(Details, Paragraph 21)
b.
The resolution by the licensee of NCR's. (Details 3 Para-graph 22)
Welding on the reactor head spray removable piping.
(De-c.
tails, Paragraph 23)
,
.
B.
StatusofPreviouslyReportedUhiresolvedItems 1.
The following previously outstanding items are now re solved:
a.
The respsnsibilities of QC inspectors relative to
,
pipe support welds to structural steel has been defined.
(Details, Paragraph 24)
b.
An NSSS FDDR exists at the site authorizing apparent
~_
ly excessive gaps thatween the shroud wall and the core spray spargers.
(Details, Paragraph 25)
c.
An NSSS FDDR exists at the site authorizing modifi-cation of the core spray sparger support brackets.
(Details, Paragraph 26)
d.
The NSSS has measured the ID of the sparger pipe.
-
(Details, Paragraph 26)
e.
The weld on the core spray sparger bracket was con-
~
tinuous.
(Details, Parec.raph 27)
.
f.
Tests for ferritic content of core spray sparger welds were made.
(Details, Paragraph 28)
g.
A document detailing the core spray sparger rework
-
is available at the' site.
(Details, Paragraph 29)
h.
QC documentation of the repairs to the damaged top fuel guide is available at the site.
(Details, Para-graph 30)
-
L----
_----------1---------------------
_ _,. _. _.... -.-
-..
. _...
.,..
..
-
. _.
.....
.. _ _
-.. _.,
.. - _.
.
.-
-... _
.,
..
..............
-'-
O O
.
.
i.
Maintaining the cable spreading room at 250C need not be relied upon to justify apparent cable tray over-loa'd. (Details, Paragraph 31)
j.
The correct maximum allowable cable pulling tension has been recorded on the pull card for cable.
(Details, Paragraph 32)
.
2.
The following previously outstanding items continue unresolved:
'
~
a.
The completion of the valve wall thickness measure-ment program has not been adequately documented.
(De-tails, Paragraph 33)
b.
The Dresser valves have not yet been reworked and re-calibrated.
(Details, Paragraph 34)
There has been no functional testing of the core c.
-
spray sparger. (Details, Paragraph 35)
.
.
d.
Access to the secondary containment is not in accord-ance with eithec GE Plant Specification 22A1202 or 22A1220.
(Details, Paragraph 36)
2 y g,i ~lk{
The cable tray barrier installation is incomplete.
7 T.lL r.
e.
(Details,. Paragraph 37)
s. y[& p
'N~
'-
..
t,-
,i r g f.
Data forms on MOV's remain incomplete.
(Details,
- ~;. " - -.
."
Paragraph 38)
,.k[Id
<
.
[.:.C $N:yy
.
g.
Acceptance criteria for pipe support attachment
$ [i' ',(.;;;
. 7..,,;
y,.
welds are not available to QC inspectors.
(Details, Mu
.
.
Paragraph 39)
M#.;W
~n v cy.W:.c 7" b f.. ).
Exit Interview e
- C,'
. 6;.(;f, f f
<w c
,
g;.ss..,. p' :
The. inspector held an exit interview at the site with the following
=
individuals:
,).
Philadelphia Electric Co.
F. M.,Valentino, QA Site Supervisor
..
F. W. Iloelzle, Jr., Site QA Engineering P. A. Tutton, Site QA Engineering D. A. Marsacio, Site QA Engineering
-
s
~
R. Costagliola, Operating QA
.
T. P. Gotzis, Construction Supervisor, Site
..
.
.
':-
.:
--
..
_ _ - -
" '
'
_.
a_,
. l
.
.
&
&
M %p..-.
-5-
.,y>
.., O.c-y.
- i. 9 ;, ? ; -
,b} f-l
'
=
,
Bechtel 4hE I."'
m.v.
.,
.
- 't A..F
';: "
E. R. Klossin, Quality Assurance Engineer yf.f..
R. L. Rich, Project Field Quality Control Enginner y.].) i '.
M. Henry, Field Engineer tX (M -.~i A. Langanke, Field Superintendent
$ f ?...
.
Items discussed are suwarized below:
.
$l5$:.D-W:.. '. 'i * e
.w.
ym.
sfef.., J.:
A.
The inspector stated that he had reviewed the licensee's action on f(..;.p'.g R
.
'g;J: O q previously identified enforcement items listed below and noted no
.g<..,
deficiencies.
hh;g., p,%
4"
'
- [tg 1.
Deviations in the design and manufacture of the wire rope
.
' g}g.[. ' h'
assemblics for the Main Steam Line Restraints have been re-
.
,g
.p.
'
viewed and approved by the original design group. This matter is resolved.
(Details, Paragraph 4)
.;
7.y, :;.__
.
..
.;
.
y ~....
2.
Pre-stressing the wire rope assembly of the Main Steam Line f.,g..
.
Restraints was apparently not excessive. This matter is q.g.l-g p (
resolved.
(Details, Paragraph 5)
yf 7 w
..; y e.. -
3.
The apparently "over-ground" condition at the attachment weld h.kh of a pipe support to a pressure pipe has been inspected and
- gl(.]h-Q there is no apparent deficiency.
(Details, Parag::aph 6)
Q. p.;.
.
g. :;. +a 4.
Documentation exists at the site indicating control of the
'/M.OI'
Reactor Vessel Stabilizers as Q-listed components.
This d.%.' U e
-
,
matter is resolved.
(Details, Paragraph 7)
$ h Tjf
.
):;f^'::
-
.h p, ' [Q 5.
Documentation exists at the site indicating re-evaluation by
.k.Sg the NSSS of vendor radiographic reader sheets for the RHR heat
'S!.:.rq exchanger showing that repairs were acceptable. This matter is resolved.
(Details, Paragraph 8)
.sgy lJfr 6.
Required material certification and documentation of pre-stressing for the wire rope assemblies of the Main Steam Line
'
Restraints exist at the site.
In addition, the Constructor has taken action to assure proper receipt inspection in the future.
(Details, Paragraph 9)
.-
.
7.
Weld procedures are entered on hanger sketches.
(Details, Paragraph 10)
B.
The inspector' stated that on the basis 'f current findings the fol-o lowing items are outstanding-
'
s e
e e
___-__
_ -___ __,___.__________ _ _
_
__
__,
,
'
-'-
O O
,
.
.
1.
The problem associated witih the Grinnell snubbers which has
-
led to a decision to replace the orifice blocks should be reported as a 10 CFR 50.55(c) item.
(Details, Paragraph 11)
2.
The location on the instrument rack of the HPCI line pressure switches apparently does not meet the requirements of the GE recommended specification NEDO 10139, dated June 1970.
(Details, Paragraph 12)
.
3.
Pipe hangers for 2" and under piping appear not to be double nutted. (Details, Paragraph 13)
4.
The RV drain line contacts the reactor pedestal wall raising the possibility of imposing undesirable stresses on RV drain nozzle.
(Details, Paragraph 40)
5.
A vertical portion of the reactor vent and drain system is out of plumb at a socket weld joint which may lead to a failure to meet the requirements of GE Spec. 22A1295, F2.1.5.2.
(Details.
Paragraph 14)
-
C.
The inspector stated that his current findings indicated no ap-parent deficiencies relative to the items listed below:
1.
The status of the licensee's NCR's.
(Details, Paragraph 21)
2.
The resolution'by the licensee of NCR's.
(Details, Paragraph 22)
-
.
.
-
3.
Welding on the reactor head spray removable piping.
(Details,
-
Paragraph 23)
D.
The inspector stated that on the basis of his current findings the following previously outstanding items are resolved:
1.
The responsibilities of QC inspectors relative to pipe support welds to structural steel has been defined.
(Details, Para-
,
graph 24)
2.
An NSSS FDDR exists at the site authorizing apparently ex-
, cessive gaps between the shroud wall and the core spray
,,
sparger.
(Details, Paragraph 25)
,
- k 3.
An NSSS FDDR exists at the site authorizing modification of
{
core spray sparger support brackets.
(Details, Paragraph 25)
'
i
'
4.
The NSSS has measured the ID of 'the sparger pipe.
(Details, ~
Paragraph 26)
,
.
i'
.
'
O O
-'-
.
.
5.
The weld on the core spray sparger bracket was continuous.
(Details, Paragraph 27)
6.
Tests for ferritic content of core spray sparger welds were made.
(Details, Paragraph 28)
7.
A document detailing the core spray sparger rework is avail-abic at the site.
(Details, Paragraph 29) -
8.
QC documentation of the repairs to the damaged top fuel guide is available at the site.
(Details, Paragraph 30)
l 9.
Maintaining the cable spreading room at 250C need not be relied upon to justify apparent cabic tray overload.
(D e-tails, Paragraph 31)
-
10.
The correct maximum allowable cable pulling tension has been
,
recorded on the pull card for cable.
(Details, Paragraph 32)
.-
11.
A pipe which shall operate at 5450F and is presently uninsul-ated is within 12" of CRD control cabic. This item will be outstanding until insulation is applied.
(Details, Paragraph 15)
12.
A sampling point in the reactor drain line was installed not in accordance 61th the Constructor's P&ID.
(Details, Para-graph 16)
.
13'.
The Reactor Water Clean-Up Pump has a plastic plug in a vent
port.
(Details, Paragraph 7)
14.
Attached to the end of the reactor drain pipe is a temporary pipe. The inspector was informed as to the system in place to assure that the temporary modification is properly dispositioned.
(Details, Paragraph 18)
15.
Two snubbers on the main steam relief valve discharge system
--
had empty fluid accumulators.
(Details, Paragraph 19)
16.
The standby liquid control relief valves do not have set point
,shown on nameplate.
(Details, Paragraph 20)
,.
For each of the items above the licensee acknowledged the inspector's remarks. In the case of item 1 above, the licensee stated that he would file the; required report. For each of items 2 to 16 above, the licensee stated that he would take the necessary action to resolve the matters indicated.
.
___
-
-
-
-8-
-
.
.
>
.
-
.
E.
The inspector stated that on the basis of his current inspection the following previously outstanding items continue unresolved.
1.
The completion of the valve wall thickness measurement program has not been adequately documented.
(Details, Paragraph 33)
2.
The Dresser valves have not yet been reworked and recali-brated.
(Details, Paragraph 34)
3.
There has been no functional testing of the core spray spargers.
(Details, Paragraph 35)
4.
Access to the secondary containment is not in accordance with either GE Plant Specification 22A1202 or 22A1220. (Details,
Paragraph 36)
5.
The cable tray barrier installation is incomplete.
(Details, Paragraph 37)
.
,
6.
Data forms on MOV's remain incomplete.
(Details, Paragraph 38)
.
7.
Acceptance criteria for pipe support attachment welds are not l
available to QC inspectors.
(Details, Paragraph 39)
In each of the above cases the licensee stated that he would fur-nish the required information and/or documentation to resolve these outstanding matters.
_
.
Y'
.-
.
_
.
O d
s
.
_ _ _ _ _ _ _ _ _. __ _ _ _ _ _. - _ ___
_ _ - - _ _ _,,, -
.-__
-
- - - - - - - - - - - - - - - - - - - - - - - -
r---
- ~ - - - - -
I O
O
'
DET ILS
1.
Persons Contacted The following persons were contacted:
Philadelphia Electric Co.
'
F. M. Valentino, QA Site Supervisor F. W. Hoelzle, Jr., Site QA Engineering P. A. Tutton, Site QA Engineering D. A. Marsacio, Site QA Engineering R. Costagliola, Operating QA T. P. Gotzis, Construction Supervisor, Site Bechtel E. R. Klossin, Quality Assurancq Engineer
.
-
W. Hindle, Project Field Engineer R. L. Rich, Project Field Quality Control Engineer R. Bowren, Assistant Project Field Quality Control Engineer M. Henry, Field Engineer A. Langanke, Field Superintendent L. Kohlbus, Quality Control Welding Engineer B. Raymond, Quality, Control Engineer CE
~
~
,
C. 0. Nelson, Site Superintendent
-
S. J. Bellows, Site QA Engineer 2.
Status of Construction The licensee reported that construction was 97% complete and that the estimated fuel loading date was May 15, 1974.
^ 3.
Insulation Applied to Valve Prior to Quality Control Release On March 14 the inspector observed two arc strikes on the reactor water clean-up system inboard isolation valve, MO 3-12-15.
The arc
strikpswereonthebodyofthevalve. The valve material is an austenitic stainless steel.
This condition was brought to the attention of site QA/QC personnel who stated there was an are strike removal procedure in effect for piping and equipment pressure parts within the reactor coolant system boundary. This procedure reportedly included inspection of system components for are strikes and removal of same. For com-ponents scheduled for insulation, release of the component by QC personnel was reported as required prior to installation of in-sulation that no are strikes are covered up.
g
_
_
_ _ - - - - -
(])
-10-()
.
On March 15, 1974 the inspector observed that valve no. 3-12-15 had been insulated since the arc strike observatien of the previous day. Upon inquiry, the inspector found that QA/QC personnel had no evidence that the are strikes had been removed nor that the com-ponent had been released by QC personnel for the installation of insulation. The insulation was then removed and'it was confirmed the are strikes had not been removed.
Subsequently the arc strikes were removed before the insulation was reinstalled.
This failure to implement procedures is a violation of Criterion V, Appendix B, 10 CFR 50 which states in part, " Activities affecting quality shall be prescribed by documented instruction, procedures, or drawings... and shall be accomplished in accordance with these instructions, procedures or drawings.
.
..
4.
Approval of Main Steam Line Res'traint Design Changen The inspector examined the letter from Bechtel to the licensee dated December 31, 1973 which referred to Bethlehem Steel Co.
Proposal #135, Becht'el Purchase Order 6280-F-30721 and stated in part, "...we have reviewed the above referenced documents regarding the main steam line' restraint wire rope design and find them acceptabic without any comment."
.
5.
Wire Rope Assembly Pre-stressing
.-
On the basis of the licensee's technical evaluation included in Attachment I to his letter to the AEC dated February 11, 1974 the pre-stressing of the wire rope to 75 Kips does not apparently affect the breaking strength of the wire rope assembly.
6.
"Over-ground" Condition at Attachment Weld On the basis of the licensee's technical evaluation included in Attachment I to his letter to the AEC of February 11, 1974 and on
,,
the results of NDE recorded on PC # H-132A which the inspector examined at the site, it may be concluded that the attachment weld f rom Pipe Support No. 3-10-DE-H-132A to spool No. 161 is without apparent deficiencies.
.
-
~
e
__ _ _ - - - - -
m n
--
_
l O
O-11-
,
.
.
.
7.
Reactor Vessel Stabilizer The QC 101 file for the RPV stabilizer dated 10-10-73 was reviewed by the inspector.
Included, was the GE Source Inspection Tag dated March 29, 1970. There were no apparent deficiencies.
.
8.
RHR Heat Exchanger Radiographs The inspector examined f or RHR Heat Exchangers S/N 07131 and 07131-1 X-Ray Indication Report Evaluation Sheet 897-Y-022 which was reevaluated by the NSSS on 9/25/73 9.
QC Documentation'of Main Steam Line Restraints (MSL)
The Inspector examined the following documents which relate to MSL Restraints and noted no deficiencies:
.
Material Certification by Clif ton Steel Co., Cleveland, O.
a.
providing chemical analysis.
b.
TWX from Bethlehem Steel dated June 17, 1974 which indicated
,
prestressing to 75 Kips.
Record of QC Inspector's Meeting on November 20, 1973 in which c.
the' topic "MCR-128C-need for assuring proper documentation available at the time of receipt of material" was, discussed.
.
_,
d'.
Waiver of requirement for mill certifications for nuts, etc.
-
included in TWX dated 2/8/74 approving washer material for MSL restraint cable anchorage.
TWX was prepared by SFHO design group.
10.
Weld Procedures on Hanger Sketches The inspector examined hanger sketches for System 16, Feedwater and
,.
System 14, Core Spray System.
In all cases the weld procedures were noted on the hanger sketches. There were no apparent defi-ciencies.
' ' ull. Orifice Blocks on Grinnell Snubbers The inspector er.amined the following documents which were generated by the necessity for replacing the orif. ice blocks in the Grinnell Snubbers:
.
Portion of Presentation to PORC, January 1974
'
a.
"By using electronic transducers on selected piping systems, the following test problem areas were observed to exist:
.
--
---
,
a
-
'
_in O
I-12-
,
- '
- 4
.
,
-
(a) RV dishcarges to torus
!
'
(b) Main Steam by-pass to condensers
...it was noted that only during the
!
1000 psi RV test the combined themal and dynamic be-havior was sufficient to close the snubber check valve.
,
During the subsequent thermal growth the snubber remained
-
locked causing thermal stress to be built up in the pipe.
'
...it was noted that the snubbers were locked either
' -
[
during themal expansion or contraction causing stresses t--
I
-
to exist when pipe is either hot or cold.
-
To resolve these problem areas it was decided to replace
-
135 snubber valves installed in Unit #2. The modifica-
tions consist of installing new valve assemblies having
E check valves and orifices with respectively large flow
,
and bleed rates."
'. = '
'
,
.
b.
Minutes of site cceting (this subject) dated January 17,
~~[
-
1974 stated in part "... Initial operating experience
'p
-
indicates that the RV-71A (Main Steam Relief) has been
-
locking on RV test operation, as it should, but then J_
remaining locked and restraining the themal growth due
'
to the steam discharge. This restrained growth would
-
impose str'ess levels beyrad the design intent, into the
-
~
piping nystem
i
,
.
,
During test operation the snubbers on the by-pass lines
=
'.
have been erratic, sometimes allowing themal expansion ll
but not contraction or sometimes not allowing expansion.
='
Dynamic operations appears adequate."
D.-
The conditions reported in the above documents meet the cri-
-[
'
teria for reporting under the requirements of 10 CFR 50.55(e).
-
m
'^12.
HCPI Line Pressure Switches
-
The close proximity of the two HPCI steam line break differential pressure switches on their instrument rack is in apparent dis-
=i
~
-~
agreement with GE Specification NEDO 10139, June 1970 edition as
regards separations. It was noted that the separations criteria of this publication are observed for the analagous RClO switches. The licensee stated that a rationale for this condition would be
-
provided.
-
,
,
-
.
.
13. Pipe Hangers for 2" and Under Piping
$
!
'
.
The inspector observed that RV drain pipe load bearing bolts were fastened with unsecured single nuts. The inspector was informed by
>
>
,
,
.
T I
.
__
-
'
(])
()
-13-
.
site QA/QC personnel that engineering instructions had not yet been received relative to this detail for pipe hangers supporting field
)
run safety related piping. The inspector was informed that pipe
-
hangers for non-ficid run safety related piping would have their load bearing bolts secured by double nuts or secured single nuts.
This item is unresolved.
,
14. Misaligned Socket Weld Joint - Line out of Plumb The inspector observed a misaligned socket weld joint on a vertical run of two-inch austenitic steel pipe located between Valve No. 3-4-91 and the connection to the reactor water clean-up system. The misalignment appeared to be about 3 degrees.
In response to the inspector's inquiry if the minimum engngement requirements of the socket weld joint were met, the inspector was advised by QA/QC personnel that fit-up procedures required a minimum gap of 1/16 inch between the bottom of the socket and the end of the pipe prior to' welding but that there was no requirement for mialmum engagement
-
i of the pipe within the socket a'fter welding. The practice of lack of socket weld minimum engagement requirements appears to be con-trary to the nuclear steam suppliers specification for the nuclear boiler system wherein for pipe sizes larger than lh inches, a minimum engagement of h inch'is required.
This requirement appears in schedule F1 of NSS specification titled " Pressure Integrity of Piping and Equipment Pressure' Parts."
This item is unresolved.
. Pipe Insulation The inspector observed that the drain piping was uninsulated and the piping was routed within about twelve inches of control rod
'
drive electrical signal cables.
The inspector was informed that the piping was scheduled for insulation.
Ins.tallation of the insulation will be confirmed at a subsequent inspection.
16. As-Built Drawing Status - Installation of Sampling Point The inspector observed that the outlet of a one-inch grab sample connection did not match the current piping and instrumentation drawing nor the applicable as-built isometric drawing M-296, Sheet 77, Revision F6.
The installed outlet was an open-ended section of h inch tubing welded in a one inch nipple instead of a capped one inch nipple.
.
_
_ _. _ _. -
l-14-
.
'
O O
-
.
.
l
.
The inspector was informed that a site non-conformance report (NCR)
was active for as built drawings of field run piping whereby QC re-view was required of final as built drawings to assure the drawing correctly depicts the as built condition.
This item is unresolved.
17. Reactor Water Clean-up System Pumps _
.
The inspector identified that the vent port on the inboard pump seals has plastic dust covers installed rather than the properly rated pressure components. These pumps were turned over for operation and were in service at the time of this inspection.
It appeared that the vent port would be subjected to reactor water at a pressure of about 1200 psig.
This item was reported as not having been previously identified by site personnel.
The inspector observed that the pump installation drawing available
~
to site personnel was illegibic on this detail.
The drawing had been marked previously for replacement with a more legible copy.
The licensee agreed to ob*tain a cicarer drawing, determine and provide the required installation on the pump seal.
This item is unresolved.
18.
Temporary Piping Modifications The inspector observed that the reactor drain piping downstream of the second isolation valve in the branch line to the drywell equip-m'ent sump was welded to a galvanized pipe for overboarding drain The inspector was informed this was a temporary arrangement
,.
water.
-
to support testing activities. The inspector understood that prior to turnover of equipment for operation, the responsible field engineer in each discipline reviews the equipment or system to assure the installation conforms with approved design and any exceptions documented. Exceptions are then resolved prior to final acceptance of the equipment of system.
"'
Removal of the temporary piping and disposition of the affected coupling to which the temporary weld is made will be inspected at a fut;ure date.
This* item is unresolved.
19. Accumulators on Hydraulic Snubbers The inspector observed the fluid accumulator was empty on a hor-izontal plane seismic hydraulic snubber (serial number -12DCN52)
installed on a reactor safety / relief valve discharge pipe as well
.
_ _ _
p,
,, +..
ge..
- "+A=>
'e
" *g'
F
- "'
'
b
'
.
L.
4 ;
-
.......
.
.
-
-
..
-..
..
.
..
._
f
%
= - -
.
O P
O
'
-15-
.
.
w
,
m-L
-
-=
--
as on an adjacent snubber. The inspector observed residual fluid s
around the outside of the accumulator connection to the snubber
asaembly. The licensee reported the snubbers were manufactured by i
Grinnell Company, Inc. and that all snubbers presently installed Q
were scheduled to be removed, reworked and reinstalled in accord-g_
ance with engineering instructions which would correct the fluid
.
loss problem.
t
.
$-
This item is unresolved.
j
-
M 20.
Standby Liquid Control System Relief Valves
^
_.W-The inspector identified that the subject liquid relief valves M
(RV39A and 39B) did not have the set point pressure shown on the
"-
valve or its nameplate. The licensee agreed to correct this condition. The item had not been previously identified by site
_
personnel.
_b
.
"-
This item is unresolved pending cc,nfirmation during a subsequent inspection that valve set point pressure is properly indicated.
Y g_
Status of Liceroce NCR's
---
ww As of March 6, 1974 there were some 20 open licensee NCR's.
'A
-
'
Resolution of Licensee NCR's y
.
The inspector audited the following recently closed licensee hCR's
%
~
'
Welding Electrical 129 E-155 130 E-164 Q
132
--[
143 144 Q
,
145 5-r
-__
He noted no apparent deficiencies.
m
~
Reactor Pressure Vessel Head Spray Piping
_ O___
23.
T-The removable reactor head spray piping was selected by the inspector
{
for audit of field welding practices.
The audit included visual
=
inspection of-welds, verification of material identification markings E
l on omponents with quality assurance records in site vault, review di ik
-
of base material and filler rod chemical composition and physical 5f
-
.
b
-
--
~{
c
-
b E
p i
-
'
o o
-
-16-
-
.
.
g
.
l properties test results, verification of welder symbols at pipe l
weld joints with welding fit-up and weld rod withdrawal record.,,
!
review of welder qualifications, review of NDT inspector quali-fications and review of objective evidence of implementation of
,
quality control procedures.
During the visual inspection of welds, the inspector observed that
,
,
I maximum weld reinforcement was on weld number 3-10-29-9 which was the last weld made on this piping and was performed in place with the reactor vessel head installed on January 11, 1974. The weld joint is an " ell" to weld neck flange.
The weld reinforcement f-varied up to 1/8 inch as measured which is the maximum permissible for.432 inch wall thickness pipe (6 inch, schedule 80) as des-
!
cribed in FSAR, Appendix A, paragraph A.l.l.
I e
The inspector selected spool pieces 2096, heat nuober 27388 and
l 2691, heat number 18435 as identified by stamping on the base
-
)
material and compared with the vault records.
No deficiencies
-
'
were identified.
-
s
!
The material certification records for heat number 18435 was selected for review and the test report from Flowline Corporation
,
dated 8/12/71 showed that the material met the requirements for j
ASTM A-312, TP 304 which agreed with the material identification marked on the base. material.
The welder symbols on field weld joints were compared with the associated WR-5 and WR-6 forms for audited weld joints and no
.,
deficiencies were found.
.
.
The qualifications for the above welders were reviewed by the inspector as well as the qualifications for the level 2 Bechtel j
and X-ray Engineering NDT personnel performing liquid penetrant and radiographic test interpretations. No deficiencies were identified.
i The inspector audited the field weld check-off list and filler
metal withdrawal records, forms WR-5 and WR-6, for weld number 3-10-29-9 and found no deficiencies.
24.
Res onsibilities of QC Inspectors on Hanger Welds to Structural Steel The inspector examined a Bechtel letter to all Mechanical QC Field Engineers which stated in part,"... Ensure that hanger structural attachment is as the sketch calls for. Measure beams and plates to
-
ensure that they are as large and heavy as called for.
Closely inspect all welding and ensure that all welds are complete (weld bead carried all around beam, or as shown.)
Inspect weld bead
_
appearance, verify weld is of sufficient size as required by
.a
,
k
_ _, _ _ _. _ _ _ _ _ - - - _
-17-
'
.
.
O O
-
.
t
.
.-
"
sketch,andthatnomorethan1/32"undercuttingofwe'.d'cNists.
"
Ensure that anchor bolts are fully engaged, (i.e.) no space cyists between plate and bolt head." '
The inspector examined Form 129 for System 6, Reactor Feed Pump Discharge, entitled " Pipe Support / Restraint Installation Review" and noted that a welding deficiency was properly noted therein by the appropriate QC inspector.
-
25.
FDDR's Relating to Core Spray Sparger i
The inspector examined the following closed NSSS FDDR's and noted no deficiencies.
FDDR HE-L-124 Fermits brackets to be modified so as to facilitate installation FDDR HE-3-126 Fermits a change in nozzle to bracket clearance
.
FDDR HE-3-127 Permits.a change in the sparger ahroud clearance 26.
Core Spray Sparger ID The vendor has entered the core spray sparger ID discuss. ion on the RD1 document attached to the licensees NCR A-6 27.. Continuous Uelds on One Side of Irackets
-
T[e inspecter determined by exami:tation of the corrective action to
'
,
the licensee's NCR-A-7 that a continuous weld was indeed provided.
-
28. Magnetic Test for Ferrite
_
The inspector determined by cuminatica of the Magnetic Test (for Ferrite) Document attached to NCR A-8 that the requirements for ferrite testing were met.
f-
.
29.
Core Spray Sparger Rework Procedures The inspector examined the following RDM/RCl procedures and noted no deficiencies:
,.
_ Procedure Title 10.26 Rev 2 - Removal of Core Spray Sparger 10.31
- Ascembly of Core Spray Sparger Systems in RCl shop
,
.
/
)
,
32y. v; w - -
-__--__-_r-------
--
r. ;_ :-- -
,
....
...
..
.-
.m
--
--=-
- ~=
-18-
'
.
.
.
O O
-
.
10.32
- Assembly of-Core Spray Sparger Into Shroud
,
10.33
- Aiming Specification for Adjustment...of Core Spray Nozzles in Core Spray Headers In connection with Procedure 10.26, the inspector verified the NDE requirement of Paragraph 9.0 " Grind attachment welds flush to shroud inner wall and P.T. inspect per RCl procedure, approved by GE and record." The above was satisfactorily ac,complished on 6/18/73.
i I
30.
Repairs to Top Fuel Guide The inspector reviewed the QA/QC documentation of the repairs by Bingham-Williamette and noted no deficiencies.
This review together with the licensee's final report dated October 25, 1973 resolves this matter.
31.
Cable Tray Overload - Spreading Room Temperature
'
The inspector examined NCR No.;A-21 which stated in part..."The
.
review determined that cabic trays and conduit do not require cable spreading room air conditioning to keep the cable temperatures at 250C or less."
This matter is resolved.
l l
,
32. Maximum Cable Pulling Tension
'
Referring to cable EA 3B1014A the inspector determined that the maximum allowable cable tension is 12,000 lbs as indicated on the pull card.
.
-
33. Valve Wall Thickness Program.
The inspector examined the Bechtel letter to the licensee dated March 13, 1974 which stated in part "...All data sheets are complete and in QC files on jobsite..."
The inspector stated that only a licensee or Bechtel statement attesting to the fact that all valve wall thickness data had been rN eivad, reviewed and approved as meeting minimum requirements A cil suffice to resolve this out-
-
st.anding item.
,. 34.
Dresser Valves
.
There is no evidence available at the site to indicate that the Dresser valves have been reworked, tested and reset in accordance with the licensce's statement reported in R0 Report 50-277/73-07.
This work had been accomplished for the Unit 2 n lves.
'
.
.
-r
,-- -
,
- 7,W W.,,
, Mw 9 -**** emeD -
'4 m Jms** N 9*'84*99f9*4ep
.___
, _ _ _ _ _ _ _ _ _ _ _ _ _
_ _ _ _ _ _
,_
_
____
__
,
_
_
,
.
-19-O O
..
.
.
}
,
35.
Functional Testing of Core Spray Sparger Nozzles The inspector examined NSSS FDDR llE-3-176 which would waive the requirement for functional testing on reinstallation of the core spray sparger not les. This proposed resolution of the above FDDR has not yet been approved by GE - San Jose and the matter remains outstanding.
36.
Entry Into Secondary Containment
.
The inspector examined a letter f rom GE to the licensee dated January 28, 1974 which stated in part"...GE Plant Spec 22Al220 does not apply at Peach Bottom. The comparable General Electric spec-ification which does apply to Peach Bottom is 22A1202, Revision 0."
The inspector referred the licensee to paragraph 4.4.2 of 22A1202 which states:
Seconda y Containment Personnel and $quipment Openings
.
"A personnel and equipment airlock shall be provided for access to the secondary containment from the outside. The airlock shall consist of two (2) gasketed doors in series. The doors shall be. interlocked so that one door cannot be opened unless the second door is sealed.
The airlock shall be pro-vided with connections for leak testing."
The licensee agreed with the inspector that the Peach Bottom faci-lity did not meet the requirements of the above paragraph. The i
licensee further stated that he would present his contention at a
-
subsequent inspection that it was never his intention to meet the above condition.
37.
Cable Tray Barrier Installation The inspector examined Bechtel's report dated March 8,1974 en-titled " Fire Barriers and Waterproof Seals." This report indicated the following status of fire barrier installation:
-
Reactor Building Radwaste Turbine Bldg 99%
99%
70%
,.
.
This item remains unresolved.
38. Data Forms on MOV's
~~
MOV data sheets have not yet been revised to standardize the nota-tion for required operating times. This condition was first identi-fied in RO Inspection Report No. 50-278/73-06.
.
f I
_
.
,
---
__ --
.
,
..
,
'
o o
-
-20-
.
.
.
.
-
.
39. Acceptance Criteria for Pipe Support Attachment Welds Acceptance criteria for pipe support attachment welds have not been furnished to QC inspectors.
This matter was initially identified in RO Inspection Report 50-278/73-07 dated 11/29/73.
40.
Drain Line in Contact with Reactor Pedestal Wall.
The inspector observed that the piping came in contact with the reactor pedestal wall such that during reactor operations the thermal growth of the piping would be biased in the direction toward the reactor pressure vessel drain nozzle connection.
The inspector noted that during reactor operation the piping temper-ature vill be about 5000F.
.
The inspector requested information relative to the as built condition which would show that excessive stresses would not be developed at the reactor vessel drain nozzle or piping. This information was
,
not available but the inspector was informed the procedure is for
-
representatives of the Bechtel home office stress analysis group to review the as built condition of this and other critical piping prior to fuel loading. This inspection is scheduled to begin around May 1 according to site QA/QC personnel.
l This item is unresolved pending documentation of engineering review that excessive stresses will not develop in the drain piping or reactor vessel drain nozzle.
.
.
O e '
=
-
_
.
.
.
e
.
O
%
.
.
.