ML20210E487
| ML20210E487 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 12/31/1985 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | |
| Shared Package | |
| ML17299B113 | List: |
| References | |
| CEN-321(V)-NP, NUDOCS 8603270361 | |
| Download: ML20210E487 (20) | |
Text
'
CEN 321(V)-NP A COMPREHENSIVE t
VIBRATION ASSESSMENT PROGRAM FOR PALO VERDE NUCLEAR GENERATING STATION UNIT 2
- SYSTEM 80 NON PROTOTYPE - CATEGORY I; EVALUATION OF PRE-CORE HOT FUNCTIONAL INSPECTION PROGRAM PRELIMINARY REPORT DECEMBER 1985
[e?"I88e!!8=jp comausTION ENGINEERING
LEGAL NOTICE TH13 REPORT WAS PREPARED AS AN ACCOUNT OF WORK SPONSORED BY COMBUSTION ENGINEERING, INC. NEITHER COMBUSTION ENGINEERING NOR ANY PERSON ACTING ON ITS BEHALF:
A.
MAKES ANY WARRANTY OR REPRESENTATION, EXPRESS OR IMPUED INCLUDING THE WARRANTIES OF FITNESS FOR A PARTICULAR PURPOSE OR MERCHANTABluTY, WITH RESPECT TO THE ACCURACY, COMPLETENESS, OR USEFULNESS OF THE INFORMATION CONTAINED IN THIS REPORT, OR THAT THE USE OF ANY INFORMATION, APPARATUS, METHOD, OR PROCESS DISCLOSED IN THIS REPORT MAY NOT INFRINGE PRIVATELY OWNED RIGHTS;OR B. ASSUMES ANY UABluTIES WITH RESPECT TO THE USE OF OR FOR DAMAGES RESULTING FROM THE USE OF, ANY INFORMATION, APPARATUS, METHOD OR PROCESS DISCLOSED IN THIS REPORT.
e
- - - - - - - - _ _ _ _ ~ _ _ _ _ _ _, _ _.__
A COMPREHENSIVE VIBRATION ASSESSMENT PROGRAM FOR PALO VERDE NUCLEAR GENERATING STATION UNIT 2 (SYSTEM 80 NON PROTOTYPE - CATEGORY I)
~
PRELIMINARY EVALUATION OF PRE-CORE HOT FUNCTIONAL INSPECTION PROGRAM
- ~:
TABLE OF CONTENTS SECTION PAGE TABLE OF CONTENTS i
LIST OF TABLES if LIST OF FIGURES 11
.l
SUMMARY
iii.
1.0 INSPECTION-1 1.1 Introduction 1
1.2 Discussion 2
1.3-Inspection 3'
1.3.1 Reactor Vessel 3-1.3.2 Core Support Barrel Exterior 4
1.3.3 Core Support Barrel Interior 4
1.3.4 Upper Guide Structure 5
1.3.5 Holddewn Ring 6
1.3.6 Reactor Vessel Closure Head 6
1.4 Conclusions 6
Tables 8
Figures 9
References 14 i
TABLES PAGE TABLE I SNUBBER CONTACT
SUMMARY
FOR 8
BOTH REACTOR VESSEL AND CORE SUPPORT BARREL SURFACES 1
FIGURES PAGE FIGURE I REACTOR VESSEL AND INTERNALS-9 I
ARRANGEMENT SHOWING COMPONENT CONFIGURATION FIGURE II REACTOR VESSEL 10 FIGURE III CORE SUPPORT BARREL ASSEMBLY 11 FIGURE IV UPPER GUIDE STRUCTURE ASSEMBLY 12 FIGURE V UPPER GUIDE STRUCTURE 13
""3 ii
c-6
SUMMARY
In accordance with the United States Nuclear Regulatory Commission, Regulatory Guide 1.20 (Ref. 1), a Comprehensive Vibration Assessment Program (CVAP) has been developed for Palo Verde Nuclear Generating Station Unit 2.
This plant is a Combustion Engineering 3800 Mwt System 80 pressurized water Reactor (Figure I) and is desicnated Non-prototype Category I.
The purpose of the CVAP is -to verify the structural. integrity of the reactor internals to flow induced loads prior to commercial operation.
The dynamic flow related loads considered are associated with normal steady state operation and anticipated operating transients.
The applicable reference prototype for Palo Verde Unit 2 is Palo Verde Unit 1.*
A complete CVAP for the prototype is found in Reference 2.
Since Unit 2 is identical to Unit 1, the required analysis for Unit 2 is the same as that found in Reference 3.
A complete inspection of the Reactor Internals for Palo Verde Unit 2 has been performed to fulfill the requirements of regulatory Guide 1.20.
This report contains the preliminary findings of this inspection program.
- (Hence forth the terms Unit 1 and prototype will designate Palo Verde Nuclear Generating Station Unit.1, and the terms non-prototype and Unit 2 will refer to Palo Verde Nuclear Generating Station Unit 2.)
iii
.A pre-core baseline and post hot functional inspection of the internals was performed.
In both cases, the internals were positioned to permit' l
visual inspection of specified. location's (Ref. 5). Major load bearing surfaces, contact surfaces,' welds'and' maximum stress locations,~as l
identified by analysis, were examined.
Comparison of results of the visual baseline inspection post hot-functional and pre-core inspection indicate no signs of abnormal wear or contact on any of the structures.
iv
1.0 INSPECTION PROGRAM 1.1 Introduction In accordance with the United States Regulatory Commission Regulatory Guide 1.20 (Rev. 2), Reference 1, a Comprehensive Vibration Assessment Program (CVAP) has been developed for Palo Verde Nuclear Generating Station Unit 2 (Ref.5). This-plant is designated a Non-prototype Category I per Regulatory Guide 1.20 (Ref.1) requiring " Analysis and either extensive measurements or full inspection". The required analysis is identical to that of the prototype Unit i reactor and is found in Reference 2.
The full inspection, in lieu of " extensive measurement", is the subject of this report.
It should be noted that Unit 2 has included identical upper guide structure modifications as completed in Unit 1.
The details of these modification and the demonstration of structural adequacy of the upper guide structure for the prototype is found in Reference 4.
The Comprehensive Vibration Assessment Program (CVAP) visual inspection for the Arizona Public Service Company Palo Verde Nuclear Generating Station (PVNGS) Unit 2 was performed in two stages; the Baseline (pre hot functional testing) stage and the Post (post hot functional testing) stage. The Baseline stage inspection is used to establish a foundation which is compared to the results of the Post stage inspection, after the reactor vessel internals have been subjected to the reouisite number of cycles of vibration. The net structural difference between the two stages is an indication of the performance of the reactor vessel internals and this performance will be compared to the same visual and
~
instrumented testing as completed on the prototype reactor for verification of performance.
l' m
1.2 DISCUSSION The PVNGS Unit 2 visual inspection program is based on regulatory position C 3.13 of the NRC Regulatory Guide 1.20, Rev. 2 (Ref.1).
The inspection of the reactor vessel internals prior to hot functional testing is denoted as the Baseline stage and the inspection following hot functional testing is denoted as the Post stage. The hot functional testing includes all steady-state and transient modes of operation. The Baseline stage inspection for PVNGS Unit 2 was completed on October 26, 1984 and the Post stage inspection was completed on August 16, 1985. Both were completed successfully without deviation from the specified conditions. NRC Regulatory Guide 1.20, Revision 2 (Reference 1) requires that the duration of hot functional testing be sufficient to ensure the critical reactor vessel internal components be subjected to at g
6 least 10 cycles of vibration. The C-E CVAP Visual Inspection Specification (Reference 5) increased that requirement to 10 cycles of vibration to establish a meaningful indication of component function. The critical reactor vessel internals component with the lowest natural frequency and thus requiring the longest hot functional testing duration is the core support barrel, which has a natural frequency of [
] Hz (Reference 2). The hot functional testing had a duration of 14400 minutes of cold (less than 350 F) coolant flow and 55440 minutes of hot (above 350 F) coolant flow for a total duration of 64840 minutes of steady-state and transient operation. Thus, all critical reactor vessel internals components were subjected to a minimum of [
]
cycles of vibration, exceeding requirements of Reference 1 and Reference 5.
The visual inspection program consisted of the inspection and photographing of all elements of the reactor vessel internals, both before and after hot functional testing. Throughout the visual inspection program, all surface orientations left (L) and right (R) are as viewed from the reactor centerline. The Baseline stage 2
inspection of the reactor vessel internals-revealed the presence of various manufacturing grinding marks and tooling marks along with assembly and disassembly' scratches over most of the-contact. surfaces. These marks are of negligible depth _,-
are' considered normal and do not impair the function or integrity of the reactor vessel internals. All welded joints appeared secure ~
The Post stage inspection revealed a general discoloration of all surfaces due'to coolant residue buildup and high temperature operation.
In general-throughout the internals, light staining and residue build-up occurs in areas of high flow wh_ile heavier residue deposits occur in areas of low flow and stagnation.
In areas where a'small gap or contact occurs, residue build-up is kneaded between the mating surfaces during relative motion of 'ifferential thermal d
expansion. High and low spots on the contacting surfaces can be observed as variations in discoloration due to this ~
relative motion. Specific observations are summarized below.
1.3 INSPECTION 1.3.1 Reactor Vessel (Figure II)
The core support barrel flange seating surface has a circumferential series of uniform bands of radial scratches at the location of core support barrel contact. These scratches are the-result of the relative thermal growth between the two ccmponents. The alignment key slot sides have indications of light contact with the alignment keys at the following locations: 0' (R & L), 90* (R & L), 180 (R) and 270' (L).
This corresponds to indications of contact on the alignment keys. Both outlet nozzle faces show evidence of minor localized contact and disassembly scratches. The snubber inserts have indications of contact with;no appreciable wear as given in Table I.
This corresponds to contact indications on the core support barrel snubbers. The attachment cap 3
screws and locking pins all appeared intact and unchanged from the baseline. The surveillance capsule holder assemblies, flow skirt attachment points and in-core instrumentation nozzles appeared sound and secure.
1.3.2 Core Support Barrel Exterior (Figure III)
All circumferential girth welds appeared secure. The snubber surfaces have indications of contact corresponding to those on the reactor snubbers given in Table I.
The outlet nozzles both showed evidence of very light localized contact with the j
reactor vessel.
I The flexure weld at the core support barrel and lower support structure interface appeared sound and secure. The lower support structure bottom plates, flow hole sleeving, in-core instrumentation support plate, columns and nozzles all appear scund and secure. The upper flange and reactor vessel interface surface has indications of contact corresponding to those on the reactor vessel. The alignment keys have indications of contact corresponding to those on the reactor vessel, holddown ring, upper guide structure and reactor vessel closure head key slots. The upper flange and holddown ring interface surface has a circumferential band of radial scratches from contact and differential thennal growth with the holddown ring. The transition radii under the core support barrel upper flange appear sound.
1.3.3 Core Support Barrel Interior Both sides of all the guide lug inserts have indications of contact with the upper guide structure guide lug slots. This corresponds to contact indications on the upper guide structure guide lug slots. The dowel pins, attachment cap screws, locking pins and lockwelds all appear sound and secure.
4
The annulus between-the core support barrel and the core shroud top plate is uniform with a gap of [
] at various circumferential locations and is onsistent with Baseline stage measurements. The insert pins, locking bars and lockwelds all appear sound and secure. The circumferential girth welds and core shroud joints, all appear sound and secu're.
1.3.4 Upper Guide Structure (Figures IV & V)
The tube sheet tubes extending below the fuel alignment plate, all appear sound and secure.
The guide lug slots have indications of contact corresponding to those on the guide lug. inserts. The alignment key slots have indications of 100% contact with the alignment keys at the 90'(L) and 180*(R) locations with localized centact at the remaining surfaces. This corresponds to indications of contact on the alignment keys. The holddown ring has uniform gaps and concentricity with the alignment key slots. The stack-up assembly dimensions taken of the core support barrel, holddown ring and upper guide structure; relative to the reactor vessel is uniform, with a height of [
] at various circumferential locations. This is somewhat less than measurements taken during the Baseline stage, due to the seating-in of the surfaces during hot functional testing, but is well within design guidelines. The CEA shroud assembly tie rod locking strap welds, upper and lower shroud tube / web' welds, and snubbers appeared sound and secure. The upper flange circumferential girth weld and the tube sheet tubes ap-peared sound and secure. The upper flange and holddown ring
. interface has a circumferential band of radial scratches from the holddown ring due to differential thermal growth and a ro-tation of the holddown ring during reactor vessel closure head tensioning. The upper flange and reactor vessel closure head interface has indications of contact with the closure head.
5
1.3.5 Holddown Ring The holddown ring and core support barrel flange interface surface has indications of. contact corresponding to those on the core support barrel flange. The holddown ring and upper-guide structure flange interface surfone has indications of contact corrt.sponding to those on the upt.er guide structure flanga. The alignment key slots have no indications of contact with the alignment keys. This corresponds to the alignment findings.
1.3.6 Reactor Vessel Closure Head i
The alignment key slots have indications of contact ~ ith the w
alignment keys at the 0*(R), 90*(L) and 270 (R) locations.
This corresponds to indications of contact on the alignment keys. The reactor vessel closure head and upper guide structure interface has indications of contact corresponding to those on the upper guide structure.
1.4 CONCLUSION
S The visual inspection program was performed to meet the requirements of regulatory position C 3.1.3 of NRC Regulatory Guide 1.20, Revision 2 (Reference 1). The Baseline stage was completed on October 26, 1984 and established a foundation on which to compare the results of the Post stage, which was completed on August 16, 1985 after the reactor vessel internals were subjected to approximately [
] cycles of hot functional testing vibration which exceeds the required 6
1 X 10 cycles of vibration.
6
The visual inspection program consisted of the inspection of all major load bearing elements, restraint elements, locking components and contact surfaces within the reactor vessel internals.. Also the reactor vessel internals were inspected for the presence of loose parts or foreign matter. ' A l
comparison of the Baseline stage surface conditions with the Post stage surface conditions indicated that no abnormal flow' induced vibration had occurred and that no reduction in the structural integrity of the reactor vessel internals, closure head or reactor vessel had occurred. There were indications of nomal amounts of relative thermal growth between the' reactor vessel internals and the reactor vessel at the flange surfaces and there were indications of contact between surfaces at the snubbers, outlet nozzles, guide lugs and alignment keys.
This contact between surface is considered normal and as designed, with little or no wear indicated. All welded joints, threaded fasteners and locking mechanisms appeared secure.
The results of a Post stage visual inspection revealed no evidence of unacceptable motion, excessive or undue wear or deviation from the predicted results of the analysis program.
The stack-up assembly dimensions taken of the core support barrel, holddown ring and upper guide structure; relative to the. reactor vessel indicate that the holddown ring exerts sufficient force on the reactor vessel internals ~to retain them during operation.
7
m TABLE I SNUBBER CONTACT
SUMMARY
FOR BOTH R.V. & CSB SURFACES I
l SNUBBER 100% CONTACT LOCALIZED CONTACT NO CONTACT 0
R X
O' L
X 60 R X
60' L X
120*R X
120*L X
180'R X
180*L X
240*R X
240*L X
300*R X
300*L X
l 8
!Q
'!:j.
h..h 3
igiaagoa es t,,,;
..q,,r
(
/
\\
iM E!""S
%s lty W" fl^s&"iY U{g?
CORE trau61uas g;;&ar 3
o.
} O&
{&
I 1
iM!uo l::
h Es?EMBLY i
e I
l g gg
$ggir/gm'!ar s
e--
p3.- sg REACTOR VESSEL AtlD INTERNALS ARRANGEMENT SHOWING COMP 0flENT CONFIGURATION FIGURE I 9
l REACTOR VESSEL gqger SURFACES 4
1 W
i
@t g
y mm%
l N
k E
-suussEn h
GF j
j0$@[.
s s
- CORE ST6'P' l
s l
Q t
x gg--
- 0
,j 4
h REACTOR VESSEL FIGURE II 10
/
g N
/
/
/
\\
l
\\
/
/
\\
mw t t
fN$d[{ N*I f
- )
1@ g
{
l h
=[>##~p'WSQ; T
alli,,d,.
j4" :
f l
/
- d b'0'p'l;
&...ss iji);
' \\;
e=
c (
[dy, a"I~
l
\\
a.
.~
.s f,Y~
"~,
e: e f.,
ll l
e""#,
8 l
ye.v'g G 5 <#. e.
l
.e 2
?>
? j J a33 l
CORE SUPPORT BARREL ASSEMBLY FIGURE III 11
s f
C Ck dl lN]
.Cl
'j F -
r
$1 j
Cl
]
jC l
i l C
C3 y
83 r
l 3
i j
- r. ' C Wi i[
1 q7 T*
- ,1 l
aug L w,g e l 3 r,
ga i
- e i
i h,
,sll m
i...,,..,,...g e
/
a h
a n
a.
a..
J J.
e NY UPPER GUIDE STRUCTURE ASSEMBLY FIGURE IV i
12
Mange Blocle p
Pinned and Bolted g,
M c M q5
/5 Ah
/
)
vf
%)
MW l
g u
t 5
MMM)
C (M y
r-l
~
) 3 ( M M F S C' l
M>-1Hb1HM s
CMMMM
~
gl[
T I'
I UPPER GUIDE STRUCTURE FIGURE V I
13
REFERENCES (1) Regulatory Guide 1.20, Rev. 2. " Comprehensive Vibration Assessment Program for Reactor Internals During Pre-Operational and Initial Startup Testing."
(2)
"A Comprehensive Vibration Assessment Program for Palo Verde Nuclear Generating Station Unit 1 (System 80 Prototype) Evaluation of Predictions and Pre-Core Hot Functional Measurement and Inspection Programs Final Report", CEN-263(V)-P, Rev. 1-P, January 1985.
(3)
"A Comprehensive Vibration Assessment Program for the Prototype System 80 Reactor Internals Palo Verde Nuclear Generating Station.
Unit 1", CEN-202(V)-P.
(4) " Final Report on the Performance of the Palo Verde Control Element Assembly Shroud". CEN-267(V)-NP, Rev.1-NP, August,1984.
(5) " Comprehensive Vibration Assessment Project Procedure for Visual Inspection of Reactor Vessel Internals for Arizona Nuclear Power Plant Unit #2", Specification No. 14373-RCE-457, Rev. 00.
14
1 h
COMBUSTION ENGINEERING, INC.
I h
_.... _.