IR 05000302/1974014
| ML19319C840 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 11/04/1974 |
| From: | Bryant J, Swan W, Vallish E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19319C827 | List: |
| References | |
| 50-302-74-14, NUDOCS 8003040801 | |
| Download: ML19319C840 (9) | |
Text
(G,)
-
g
-
? -(g'
ATOMIC ENERGY COMMISSION
'
~-
UNITED STATES
)g ', - (',i % '. '.
amd :7 C.T>: OP 3fP11 Atony GPIM TIi.E
~ '.j t
.g,:.
-
G C.+, L,-
nr.c.o.. n - cm re osa r-m.., y,,, o e,, n m. n - r.. ~ > r
,
-
-
Qgg
.... <,,..,......s.
4,... n r 4. ot.onc i a
.p> 30s
.
RO Inspection Report No. 50-302/74-14 Licensee: Florida Power Corporation
-
320134th Street South Post Office Box 14042
'
St. Petersburg, Florida 33733
'
Facility Name: Crystal River 3
Docket No. :
50-302 License No.:
CPPR-51 Category:
A3/B1
'
Location: Crystal River, Florida Type of License: B&W, PWR, 2452 Mwt, 855 Mwe Type of Inspection: Routine, Unannounced, Construction Dates of Inspection: October 10-11, 1974 i
f Dates of Previous Inspection: September 11-13, 1974
,
Principal Inspector:
E. J. Vallish, Reactor Inspector Facilities Section Facilities Construction Branch Accompanying Inspector:
W. B. Swan, Reactor Inspector Engineering Section Facilities Construction Branch Other Accompanying Personnel:
J. C. Bryant, Senior Inspector Facilities Section
,
j lities Construction Branch Fa Principal Inspector:
Q/
// -/ - 7 </
,
~
E. J. Ja11ish, Reacter Inspector Date-Facilities Section Facilities Construetion Branch Reviewed By:
f +s #
~
// M 7g J.
. 'Bryant(Senior' Inspector
' ' Date F ilities Section Facili".'.es Construction Branch
.
8003040[C/
. -
-
..
_-
- - -.
. -.
.
.
-.
-. -. )
-
s
-
.
,
a:
.
-
'~
RO Rpt. No. 50-302/74-14-
-2-
~7 -
.
s SUMMARY OF FINDINGS
'
c.
I.
Enforcement Action
.
A.
Violations None
/
B.
Safety Items
'
None
6-II.
Licensee Action on Previously Identified Enforcement Matters
'"
'
A.
Violations There were ao previously identified enforcement matters.
B.
Safety Items There were no previously identified safety items.
III. New Unresolved Items 74-14/1 Reactor Internals Handling Fixture Design Deficiency -
(10 CFR 50.55(e) Item)
The design of a spherical nut capture device could have allowed the aut to become unthreaded to such an extent that the weight of the internals would have sheared the threads, dropping the internals.
A field chcnge to correct this deficiency (FC-76) has been issued by B&W. FPC will monitor the B&W corrective action.
(Details II, paragraph 3)
.
74-14/2 Seismic Qualification of the Control Rod Drive Control System i
AC Breaker Cabinets.
(10 CFR 50.55(e) Item)
A re-evaluation of the seismic qualification of the control rod drive control system AC breaker cabinets has revealed that the attachment of the cabinets to the floor is inadequate to with-stand the SSE acceleration levels.
A field change will be issued by B&W to modify the attachment of the AC breaker cabinets to the floor. FPC will monitor the B&W corrective action.
(Details II, paragraph 4)
~
.
- -, -,. - - -
,,c
,
-
-.,
,,.
-
-.
_
,
.
.
~
_
,
-
-.
-
,.
RO Rpt. No. 50-302/74-14-3-
.
.
T.
s sl
,
.
IV.
Status of Previously Reported Unresolved Items
~
.
74-10/3 Defective Westinghouse Type-W-2 Control Switch Component (DROB No. 74-6)
.
Recent inspection findings indicate a possible safety problems may exist in the plants using Westinghouse Type W-2 control i
switches. The licensee is. investigating this possibility for.
'
Crystal River 3.
This item remains open.
74-10/4 Deficiency in ITE Molded Case Circuit' Breakers, Type HE-3
,
(DROB No.'74-8)
-
.
e Deficiencies were discovered during preoperational testing of the Trojan Nuclear Plant of magnetic trip elements in these circuit breakers. The licensee is investigating this possi-bility for Crystal River 3.
This item remains open.
~
.
74-10/5 Deficiency in the General Electric Model 4KV Magne-Blast Circuit Breakers (DROB No. 74-9)
The licensee has determined that none of this model breaker
is being used at Crystal River 3.
This item is closed.
i 73-5/3 Valve Wall Thickness Verification i
1-Thickness measurements were not made of valve bonnets, a.
cover plates or nozzles adjacent to the weldend prepa-ration.
This item remains open.
b.
The engineering evaluation of the valves' thin walls was not acceptable.
FPC stated that the valves would be repaired by welding to meet the requirements of the original purchase order. This item remains open.
FPC is waiting for an engineering evaluation from c.
Babcock and Wilcox Company justifying the wall
'
thickness at the weld joint of the pilot operated relief valve. This item remains open.
i 73-5/14 Reactor Internals Modifications
,
Kodification of the internals is being accomplished in the
,
'
vendor shops and verification will be made af ter the internals are delivered to the site. This item remains open.
!
V.
Design Changes None
_.. _.,.... _
__ _,.,.
,,, _, _ -,..
__
.., _
... _...,..,
. _. -... _,. =.,,,,,.,.
_...,m._
.-. _
,q.
,
.
l RO Rpt.'No. 50-302/74-14-4-.
-
i
..
!~
VI.
Unusual Occurrences t
.
None VII. Other-Significant Findings
-
'
None
.t-
.
VI'I.
Management Interview
,.
-
>
An exit interview was held with Messrs. C. E. Jackson, Superintendent, i
Site Construction, and R. W. Slater, Quality Engineer. The inspectors
.;
related the results of their inspection. No new violations were found.,
but two netz unresolved items were identified.
i
i i
.
.
.
,
'
4 i
l I
,
s
!
.;
"
i (
,
'
..
.
...
..
__
.
.
.
. -..
..
..
..
.
m; _
m
.
.
I RO Report No. 50-302/74-14 I-l (
.
A DETAILS I-Prepared by: [..-[. M _
j._,
I/-/- 7f
.
W. B. Swan, Reactor Inspdctor Date Engineering Section Facilities Construction Branch
'
Dates of Inspection: October 10-11, 1974
,-
~
Reviewed by:
.
be.cI_
//-/- 74
.
L. L. Beratan, Senior Inspector Date Engineering Section j
Facilities Construction Branch i
,
-
.-
.
1.
Persons Contacted a.
C. E. Jackson - Construction Superintendant E. E. Froats - Manager, Generation Quality and Standards (Nuclear)
R. W. Slater - Quality Engineer T. L. Baker - Supervisor, Quality Records (\\
R. S. Dorrie - Civil Engineer D. W. Pedrick, IV - Operations Compliance Engineer
,
b.
Contrar :or Organizations Pittsburgh Testing Laboratories (PTL)
G. B. Browne - Chief Inspector
Gilbert Associates, Inc. (GAI)
J. M. Cockroff - QA Site Representative, Post Tensioning System Babcock and Wilcox Construction Co. (B&WCC)
L. M. Watson - Project Manager j
F. J. Richardson - Project Engineer, Procedures
!
R. Shope - Site QA Representative J. R. McGill - QC Supervisor
,
I Conosco Midcontinent, Inc. (Conesco)
A Division of Construction Engineering and Services Company e
G. Clayton - Prescon QA Consultant to Conesco on button heading
'
i
,
I
!
\\
and stressing of tendons.
l
.i
_ _.
_
_
. _. _
__
_
. _ _.
_
_
__ _
_ _,.
_
.
'%
,%
)
.
.
p.(,)
RO Rpt. No. 50-302/74-14 I-2
-
.
'~
2.
Post Tensioning Tendons - Buttonheading 1md Stressing
'All tendons had been pulled into their sheaths and the sheaths had been temporarily capped for interim protection of thg tendons until
'
'the sheaths are filled with grease. On this containment the designers specified the use of conduit for sheaths instead of the more conven-
,
tional spiral wound thin wall sheathing, and these had been pneumatically
-
tested before concrete placement. No concrete grout could intrude and cleaning problems were minimized. As a result, no sheath became plugged
,
so installation of every design specified tendon was accomplished.
,
'
Buttonheading and stressing of the tendons was being done by Conesco crews under the QA surveillance of Prescon and PTL specialists.
'
Implementation of the program had been audited.
Prescon/Conesco Procedure, Field Post Tensioning Procedure, was being
'
used, including Addendum No. 5 approved May 15, 1974, and field ' changes through No. 16 approved September 23, 1974. GAI specification SP-5909 was referenced.
Sections of the procedure and the changes were reviewed prior to observation of work in progress.
/'~'g
.
(
)
Step by step QC inspections were observed on buttonheading of a dome tendon and stressing of four vertical tendons including 34V24 and IV8.
.
Inspection and audit records of the work were reviewed.
Problems had
!
been noted and corrective actions initiated.
For example, instructions had been issued to, " check all stressing records between September 23 and October 2,1974, Ram #5 out of calibration." A master calibration gauge was available for calibration checking of ram gauges such as
.
when one gauge had been accidentally overpressured.
During one stressing observed an excessive elongation was initially l
recorded.
A sliding gauge was found to have slipped from jarring of l
the machine. The gauge was stabilized and the tendon was relaxed (
and restressed.
Elongation was found to be in the acceptable range.
,
No deficiencies were noted ih the procedure, records, or QC of the
work.
b The stressing of dome tendons and horizontal " hoop" tendons will be more complex than that for the vertical tendons.
For these jacks i
(rams) are operated simultaneously. The stressing sequence requires
'
that some pairs of tendons and several sets of three be stressed at the same time. This will require synchronization of operations of a maximum of six crews on suspended scaffolds.
\\
l
- -
-
-.
--.
.__
_.
.
.
y
.
~
.
's
..
(%
.
- /
RO Rpt. No. 50-302/74-14 I-3-
\\~.
,
i I
3.
Reactor Internals
.
j, The major internals were in storage. They are sealed in two large metal transport containers in wood frame canvas shelters. They are
-
p subject to daily inspections.
'
';
Installation will be done by about six B&WCC boilermakers under the surveillance of B&W QA-QC men and a metallurgist NDE specialist of
~
FPC.
Installation is expected to start after December 1, 1974, and to take a maximum of two months -to complete.
,
'
The sealed containers are to be moved into the internals storage area
_
of the containment and be depackaged in a clean room enclosing the containers and reactor vessel top.
B&W procedure 090, Rev. O, dated September 27, 1974, was approved and issued for use on October 17, 1974.
It references supplemental *
documents:
B&W drawings, specifications and field generated procedure FCP-090 and the procedure was reviewed.
It lists 1,370 "seqaences" of installation and QC actions.
\\
The resident B&W representatives are experienced in QC but only one,
,
N the QC supervisor, has been involved (at Duke's Oconee plant) with installation of RV internals for NSS-7.
The project manager stated that the boilermakers and all parties involved will be briefed on j
Procedure 090 and referenced documents before operaticns are started.
,
n
e
_ _ _.
-
,
_. _ _
_
_
.
q
.s
--
-
,
RO Rpt. No. 50-302/74-14 II-1
.
A DETAILS II Prepared by: ~
[
E
'
E. Jg Vallish, Reactor Inspector Date Facilities Section-Facilities Construction Branch
Dates of Inspection:
October 10-11, 1974
,
Reviewed Byi b6,,[
/// /[/y
-
yC'Bryanf/,' Senior-Inspector.
/ Date
.
Facilities Section-
-
Facilities Construction Branch
<-
1.
Persons Contacted
.
.
a.
Florida Power Corporation (FPC)
R. Slater - Quality Engineer
.
x b.
Contractor Organization
,
Pittsburgh Test. ag Laboratory (PTL)
G. B. Browne - Chief Inspector 2.
General Construction A valk-through was made of the general plant construction for surveillance of the overall quality of workmanship. No deficiencies were observed.
3.
Reactor Internals Handling Fixture Design Deficiency (10 CFR 50.55(e) Item)
A design review by B&W engineering has found that the set screw concept
.
of preventing -the spherical nuts from rotating relative to the threaded rods; i.e., become unthreaded is inadequate in the original design.
The spherical nuts could be unthreaded during the adjusting procedure, leaving inadequate thread engagement to support the weight of the internals.
Should the threads shear off, the resulting drop of the internals from as little as 6 inches height could result in fuel damage. Drops from greater heights could also result in damage to the internals and the reactor vessel.
O
'
m
__
-
_- -
q q
,
l-
'
,-~
-
(V
RO Rpt. No. 50-302/74-14 II-2
.
All affected plants have 'been notified of the problem.
There has' been no failure of the internals handling fixture and,.thus, no damage or adverse effect on safety resulting from this deficiency.
'
'
A redesign of the capture device has been implemented on other contracts.
,
FPC will monitor B&W's' corrective action on this item.
'
,
4.
Seismic Qualification of the Control Rod Drive Control Systen AC Breaker
.
Cabinets (10 CFR 50.55(e) Item)
A reevaluation of the seismic qualification of the control rod drive
,
control system AC breaker cabinets has found that the attachment of the cabinet to the floor is inadequate to withstand the SSE acceleration
levels.
.
.
.
The control rod drives are tripped by loss of power to the drive stators.
'
Power is interrupted by disengaging either the AC or DC CRD breakers.
i Should the breaker function be disabled in some manner, the control rods would have to be driven into the core to shut the reactor down.
Analyses cannot show that the failure will always be fail safe. Therefore, the deficiency in the design of the structural attachment of the AC
'
breaker cabinets to the floor could adversely affect safety.
FPC will follow the corrective redesign and field correction of this item.
5.
Nonconforming Materials, Parts or Components Inspection was made of the QC program for the containment building
tendon installation and stressing operations. Nonconformance reports were reviewed for analysis of nonconformity, prompt action and ultimate disposition. No deficiencias were evident.
.
i o
-
,
.
,.
..,.
.-
-
.
-
-
_., _ - -
-
-...-