ML13336A257
| ML13336A257 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 05/14/1990 |
| From: | Tatum J Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9005230221 | |
| Download: ML13336A257 (34) | |
Text
May 14, 1990 Docket No. 50-206 FACILITY: SAN ONOFRE NUCLEAR GENERATING STATION, UNIT NO. 1 LICENSEE: SOUTHERN CALIFORNIA EDISON COMPANY
SUBJECT:
SUMMARY
OF MEETING HELD AT ONE WHITE FLINT NORTH ON MAY 7, 1990, RE:
THERMAL SHIELD REPAIR On May 7, 1990, the NRC staff met with representatives of Southern California Edison Company (SCE) to discuss SCE's thermal shield repair plans. Persons attending the meeting are identified in Enclosure 1.
The meeting was held to discuss the details of SCE's Amendment Application No.
181 dated April 20, 1990, on this subject to facilitate NRC review efforts.
SCE has requested NRC approval of the thermal shield repair plans by July 15, 1990, to support the Cycle XI outage schedule.
As a result of the meeting, SCE has agreed to provide additional information to support NRC review efforts. The information required is included in.
Viewgraphs used at the meeting are included in Enclosure 3.
original signed by James E. Tatum James E. Tatum, Project Manager Project Directorate V Division of Reactor Projects -
- III, IV, V and Special Projects Office of Nuclear Reactor Regulation
Enclosures:
- 1. List of Attendees
- 2. Additional Information Required
- 3. Viewgraphs DISTRIBUTION Docket File NRC & Local PDRs FMiraglia JPartlow PDV Reading JLarkins JTatum OGC EJordan NRC Participants ACRS (10)
JRogge (17G21)
,D P/PDV (A) P PDV tum:sg ins 5//.5/90 5/4 90 b FO
'005230221 90)514 PDR ADOCK 05000206 P
pg REG 0
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 May 14, 1990 Docket No. 50-206 FACILITY: SAN ONOFRE NUCLEAR GENERATING STATION, UNIT NO. 1 LICENSEE: SOUTHERN CALIFORNIA EDISON COMPANY
SUBJECT:
SUMMARY
OF MEETING HELD AT ONE WHITE FLINT NORTH ON MAY 7, 1990, RE:
THERMAL SHIELD REPAIR On May 7, 1990, the NRC staff met with representatives of Southern California Edison Company (SCE) to discuss SCE's thermal shield repair plans. Persons attending the meeting are identified in Enclosure 1.
The meeting was held to discuss the details of SCE's Amendment Application No.
181 dated April 20, 1990, on this subject to facilitate NRC review efforts.
SCE has requested NRC approval of the thermal shield repair plans by July 15, 1990, to support the Cycle XI outage schedule.
As a result of the meeting, SCE has agreed to provide additional information to support NRC review efforts. The information required is included in.
Viewgraphs used at the meeting are included in Enclosure 3.
ames E. Tatum, Project Manager Project Directorate V Division of Reactor Projects - III, IV, V and Special Projects Office of Nuclear Reactor Regulation
Enclosures:
- 1. List of Attendees
- 2. Additional Information Required
- 3. Viewgraphs
Mr. Harold B. Ray San Onofre Nuclear Generating Southern California Edison Company Station, Unit No. 1 cc David R. Pigott Mr. Richard J. Kosiba, Project Manager Orrick, Herrington & Sutcliffe Bechtel Power Corporation 600 Montgomery Street 12440 E. Imperial Highway San Francisco, California 94111 Norwalk, California 90650 Mr. Robert G. Lacy Mr. Phil Johnson Manager, Nuclear U.S. Nuclear Regulatory Commission San Diego Gas & Electric Company Region V P. 0. Box 1831 1450 Maria Lane, Suite 210 San Diego, California 92112 Walnut Creek, California 94596 Resident Inspector/San Onofre NPS U.S. NRC P. 0. Box 4329 San Clemente, California 92672 Mayor City of San Clemente San Clemente, California 92672 Chairman Board of Supervisors County of San Diego 1600 Pacific Highway Room 335 San Diego, California 92101 Regional Administrator, Region V U.S. Nuclear Regulatory Commission 1450 Maria Lane, Suite 210 Walnut Creek, California 94596 Mr. John Hickman Senior Health Physicist Environmental Radioactive Management Unit Environmental Management Branch State Department of Health Services 714 P Street, Room 616 Sacramento, California 95814 Mr. Don Womeldorf Chief, Environmental Management California Department of Health 714 P Street, Room 616 Sacramento, California 95814
ENCLOSURE 1 LIST OF ATTENDEES NRC SCE J. Larkins F. Nandy D. Terao J. Reilly C. Trammell R. Ashe-Everest J. Tatum R. Ornelas C. Hinson M. Motamed L. Wharton G. Stawniczy J. Rajan R. Hermann A. Wang M. Hum C. Sellers WESTINGHOUSE OTHER C. Boyd J. Porowski J. Goossen (O'Donnell & Assoc)
B. Bevilacqua T. Landgraf D. Dominicis (Bechtel-KWU-Alliance)
D. Forsyth M
M. Mundt (SDG&E)
ENCLOSURE 2 Additional Information Required to Facilitate NRC Review
- 1. Details of ALARA considerations, including dose assessments and time-motion studies.
Approximate Submittal Date: 5/20/90
- 2. Details and results of qualification testing for fasteners (torque vs.
preload).
Approximate Submittal Date: 5/30/90
- 3. Final drawing details.
Approximate Submittal Date:
5/30/90
- 4. Material Specifications for KWU and Westinghouse fasteners.
Approximate Submittal Date: Not Established
- 5. Details and justification for exception to ASME fatigue design requirements.
Approximate Submittal Date:
5/30/90
- 6. Final design criteria.
Approximate Submittal Date:
6/15/90
- 7. Description of measures being taken to assure RCS cleanliness during repair evolution and prior to returning the unit to service.
(This item was discussed with SCE following the meeting).
Approximate Submittal Date:
5/20/90
- 8. Description of followup inspections planned following repair of the thermal shield, and frequency for conducting inspections.
Approximate Submittal Date: Not Established
ENCLOSURE 3 TECHNICAL PRESENTATION ON THERMAL SHIELD SUPPORT SYSTEM REPLACEMENT AND CYCLE 11 MONITORING PROGRAM FOR SAN ONOFRE UNIT 1 SOUTHERN CALIFORNIA EDISON MAY 7, 1990
SONGS 1 REACTOR VESSEL THERMAL SHIELD SUPPORT REPLACEMENT AGENDA May 7, 1990 I. INTRODUCTION R. ORNELAS II. TECHNICAL OVERVIEW R. ASHE-EVEREST III. DESIGN OF NEW SUPPORT SYSTEM R. ASHE-EVEREST A. Comparison of SONGS I and J. GOOSSEN Haddam Neck Design B. Implementation C. Design Criteria D. Response to NRC Questions IV. FUTURE INSPECTIONS R. ASHE-EVEREST V. THERMAL SHIELD MONITORING SYSTEM R. ASHE-EVEREST (AS REQUESTED)
VI. LICENSING ACTIONS R. ORNELAS VII. QUESTIONS AND
SUMMARY
ALL
0 THERMAL SHIELD SUPPORT REPLACEMENT SAN ONOFRE NUCLEAR STATION - UNIT 1 OVERVIEW o
Commercial Operation -
January 1, 1968 o
Effective Full Power Years -
12 Years o
Prior to Cycle X refueling, several thermal shield inspections performed with no unacceptable degradation observed o
Cycle X refueling thermal shield inspection revealed limited support degradation Five of six flexures broken (four were known to be broken from prior inspections)
Three of thirty support block bolts broken One dowel pin in lower support block had a broken tack weld o
Analysis performed and monitoring systems added to allow operation through Cycle X o
NRC required inspection of thermal shield support system o
SCE has elected to replace the support system during the 1990 actual outage 0
Thermal shield material and core barrel material are 304SS SA240
SONGS Reactor Internals Reactor Vessel Reactor Vessel Head Upper Reactor Internals Thermal Shield Core Support Barrel Thermal Shield Lower Supprt
Specimen Tubes Core Support Spcm nTb Barrel Specimen Basket Expansion Joint Thermal Shield Flexure Fixture (Typ 6 Places)
Thermal Shield Umiter Key Support Block (Typ 4 Places)
(Typ 6 Places)
Existing Design A
mThermal Shield Thermal Shield Support Block (Typ 6 Places) eB Core Barrel Thermal Shield 2.18 Min Top Bolt (Hidden)
.7500 Top Bolt
.8750 Dowel Pin Dowel Pin Intermediate
.7500
.750 Diffuser Plate Bottom Bolt Support Block
.5000 Section A-A Section B-B
Mtl.
Bolts and Dowel Pins New Design 316 Strain Harden Stainless Steel Thermal Shield SCHCAP..R ShownOWith OocLing Plug Three P.7acesn875 - 9 x 2.75 Long SCH CAP SCR AHC SCical 2 Places Typicalx5.00 Long SCHCAPSCR T ial3 Places 1.250 x 5.31 Lon Dowel Typical 3 P.ce Core Barrel
.500 - 13 x 2.75 Long'
.820 x 4.480 Long Dowel SCH CAP SCR Typical 2 PlacesSuprBlc Typical 2 PlacesSuprBlc
Section A-A Thermal Shield Core Barrel Support Block Mtl.
304 Stainless Steel 2.000 2,945 1.560 Support Block
.0
Section B-B Thermal Shield Core Barrel 1.900 3.29 1.560 Support Block
LOWER SUPPORT BLOCK NEW DESIGN OLD DESIGN o
11" wide; more than three 8" wide times contact area thermal shield and support blocks o
Seven bolts Five bolts 75% increase in area o
Five dowel pins Four dowel pins 150% increase in area o
Dowel pins designed for Shrink fit interference fit o
Bolts & dowel pins will be Bolts - welded lock bar restrained by a locking nut Dowels welded or crimped head
Coo IJL
New Design
.005/.010 o(2l50 thermal shield)
- 2. 46 thermacl shield radial (4.255) 0.820a.sowel Therrnal shield 1.000-8 x 1.25 min radial support 300yp effective threads (typ) key
.100
New Thermal Shield Flexure Assembly Mtl. XM-19 High Alloy Stainless Steel Flexure Dowel Core Barrel Bolt Thermal Shield
New Thermal Shield Flexure Assembly Dowel Goo Bolt Flexure Core Barrel Thermal Dowel Original design Bolt previously welded 00)
FLEXURE DESIGN COMPARISON NEW DESIGN OLD DESIGN o
Material SA-479 XM-19 high ASTM A-276 type 304SS alloy stainless steel (annealed)
Tensile 100 KSI Tensile 75 KSI YieldYield 30 KSI Hardness 293 Brinell Hardness 202 Brinell o
Flexure fabricated in one Welded in web area piece (no welds) o Preloaded flexure web area No preload o
Web geometry improved to reduce stress o
One flexure will be moved to reduce loading
MATERIAL DATA COMPARSION 316ss vs XM-19 CHEMICAL REQUIRMENTS : 1986 EDITION ASME B&PVC
% COMPOSITION SA-479 ELEMENT 316ss XM-19 C
.03
.06 MN 2.0 4/6 P
.045
.04 S
.03
.03 Sz
- 1.
- 1.
CR 16/18 20.5/23.5 Ni 10/14 11.5/13.5 Mo 2/3 1.5/3.0 N
.1
.2/.4 CB
.1/.3 V
.1/.3
MATERIAL DATA COMPARSION 316ss vs XM-19 CORROSION RESISTANCE PROPERTIES LABORATORY TEST DATA (SUPPLIED BY ARMCO STEEL)
TEST MEDIUM CORROSION RATE INCHES PER YEAR (UNLESS SPECIFICALLY INDICATED)
XM-19 316ss (ANNEALED)
(ANNEALED) 10%FECL3
<.001 G/IN2 0.011 G/IN2 (25 C PLAIN) 10%FECL3
<.001 G/IN2 0.186 G/1N2 (25 C CREVICED) 1%H2SO4
<.001 0.002 (80c
)
2%H2SO4
<.001 0.011 (80c
)
5%H2S04
<.001 0.060 (80c
)
5%H2SO4
<.194 0.26 (BOILING )
1%HCL
<.001 0.012 (35c
)
2%HCL 0.024 0.021 (35c
)
65%HNO3 0.010 0.012 (BOILING )
70%H3PO4 0.203 0.202 (BOILING )
33%ACETIC ACID
<.001
<.001 (BOILING )
TESTING PERFORMED ON 5/8" DIA X 5/8" LONG MACHINED CLINDERS.
RESULTS AN AVE. OF FIVE 48 HR. PERIODS.
S S
MATERIAL DATA COMPARSION 316ss vs XM-19 MECHANICAL REQUIRMENTS : SA-479 1986 EDITION ASME MECHANICAL PROPERTIES MECHANICAL TENSILE YIELD ELONGATION REDUCTION BRINELL PROPERTIES STRENGTH MIN 2 IN MIN AREA MIN HARDNESS (MAT TYPE)
(KSI)
(KSI)
MAX 316ss 70 25 30 40 XM-19 100 55 35 55 293
THERMAL SHIELD SUPPORT SONGS 1 HADDAM NECK o
Replace lower support with Original blocks retained.
larger design.
o Interference fit lower sup-No adjustments made.
port between thermal shield and core barrel.
o Bolting and dowel pin system New bolts installed*
has been redesigned.
only backed out dowel pins replaced.
o Dowel pins installed with Clearance fit installation interference fit, of new dowels.
o Four new limiter keys.
No refurbishment of original keys.
o Six new, improved flexures Six new limiter key and (no wear surfaces).
keyways.
o New support system incorporates lessons learned from Haddam Neck.
San Onofre Unit 1 Reactor Cavity With Core Barrel and Upper Internals Removed Work Platform Water Level -41' Elevation -42' 122.00 Dia.
I'...
PERFORMANCE OF THE QUALIFICATION TESTS INTERFERENCE FIT Test No.
Diameter 1
2 3
4 Dl
.8125" 0.0002" 0.0005" 0.0008" 0.0011" D2 1.25" 0.0003" 0.0007" 0.001" 0.0013" D3 1.75" 0.0004" 0.0008" 0.0011" 0.0014"
LOCKING SYSTEM QUALIFICATION TEST DETORQUING METALLURGICAL TEST PRESS IN TEST BOLT 1 DOWEL PIN DIMENSIONAL VISUAL OF LOCKED SECTION AND ORIGINAL DIMENSION EXAMf,~ SPEC~IEN A-A B-B PUSHLQ MATEIAL 81 (4) 1.00-8x2 long CH CAP SCR (4)
(4)
(3)
(1) x 82 C) 0.500-13x2.75 long SHCAP SCR (4)
(4)
(3)
(1) x D4 (2) 1.75"-4.' long dowel pin (2)
(2)
(2)
D5 (2) 1.25N-5.2190 long dowel pin (2)
(2)
(2)
D6 (2).8125'-4.045" long dowel pin (2)
(2)
(2) 83 (2) 1.00-8x2 long Detorque measurement on unlocked SCH CAP SCR (2)
(2) locked bolts 1)
D7 (2) 1.25-5.215 long dowel pin (2)
(2)
Cyclic loading test 2)
(2) x NOTES: 1) Measurement of effect of crimping on bolt preload by detorquing the nuts instead of bolt head.
- 2) Test parameters see paragraph 3.2e.
QUALIFICATION TEST PRELIMINARY TOOLING TESTS
- 1. Removal of the old parts:
EDM of locking on fasteners (bars, washers, dowel pins).
Pull out of dowel pins.
Detorque of bolts.
Removal of remaining parts of broken fasteners.
- 2. Preparation for installation of new design:
EDM pilot holes and counterboring.
EDM recess for mechanical locking of fasteners.
EDM recess for flexure and key.
EDM thread 1.5".
Thread cutting 1".
Reaming.
- 3. Final installation:
Press in dowels.
Torque bolts.
Locking of fasteners.
Tool testing in addition with original tools and equipment on mock UPS.
SCE THERMAL SHIELD SUPPORT SYSTEM REPLACEMENT DESIGN CRITERIA o
DESIGN LIFE Goal of 15 operating years o
ASME CODE The ASME Boiler and Pressure Vessel Code,Section III, Nuclear Power Plant Components,Section III, Subsection NG, does not apply to the thermal shield.
However, for conservatism the 1986 edition of the ASME Code,Section III, Subsection NG, will be used.
o LOAD COMBINATIONS Normal and Upset Conditions o Deadweight + pressure + FIV o Deadweight + pressure + seismic + FIV o Deadweight + pressure + thermal + FIV o
Deadweight + pressure + seismic + thermal + FIV Faulted Conditions o
Deadweight + pressure + seismic 0
NSSS OPERATING CONDITIONS Flexure preload will be established based on 100% full power with reduced Tavg conditions.
Evaluations will also be performed to show acceptability at 92% full power at reduced Tavg conditions.
O SEISMIC seismic condition loads will be considered in -the design of the Thermal Shield Support System.
The San Onofre Design Basis Earthquake Spectra used at the Reactor Vessel Supports is commonly referred to as the.67 g Modified Hausner Earthquake.
14L, Fdition APPENDIX I Fig. 1-9.2.2 4P#.1 99 8Qvs.
Fr44/m co Argoat o
Lo4 CeA7oL Curve A 24 IL Curve 8 Curve C to 107 1
109 1010 101l NOTE:
Numbert of cycles. N E - 283 X 106 psi Criteria for the Use of the Curves in This Figuret. 2.3.4 Elastic Analysis of Material Elastic Analysis of Other Than Welds and Welds and Heat Curve Heat Affected Zones Affected Zones A
(P L b
O)Range CI tjtfksi C
(PL +
1 b
OlRange (PL
- b # 0 Range
> JV9 ksi NOTES:
(1) Rtange applies to the individual quantities P,. P. and 0. and applies to the sat of cycles under consideration.
(2) Thermal bending stresses resulting from axial and radial gradients are excluded from 0.
(3) Curve A is also to be used with inelastic analysis with S. * % 6.*. where A.
is the total effective strain range.
(41 The maximum effect of retained mean stress is included in curve C.
FIG. 1-9.2.2 DESIGN FATIGUE CURVE FOR AUSTENITIC STEELS, NICKEL-CHROMIUM-IRON ALLOY, NICKEL-IRON-CHROMIUM ALLOY, AND NICKEL-COPPER ALLOY FOR S, s 28.2 ksi, FOR TEMPERATURES NOT EXCEEDING 800F (For S. > 28.2 ksi, use Fig. 1-9.2.1.)
Table 1-9.2.2 Contains Tabulated Values for Accurate Interpolation of This Curve 175
DRAFT PROPOSED CODE CASE Alternate limit on primary plus secondary stress intensity range permitted for austenitic stainless steel to use Design Fatigue Curve A in Figure 1-9.2.2,Section III, Division 1.
Inquiry:
When using the Design Fatigue Curve on Figure 1-9.2.2 for austenitic stainless steel, is it permissible to limit the primary plus secondary stress intensity range excluding thermal bending to the elastic portion of the applicable cyclic stress-strain curve in lieu of meeting the specified value of 27.2 ksi?
Reply:
It is permitted to use the minimum elastic limit of 44 ksi for austenitic stainless steel in the fatigue analysis in lieu of the specified value of 27.2 ksi to determine the acceptability of using Design Fatigue Curve A on Figure 1-9.2.2.
Keeping the primary plus secondary stress intensity range below this value of 44 ksi assures the fatigue analysis is independent of whether the applied loading is strain or load controlled.
NRC QUESTIONS Q
What is the material for the fastener locking devices?
A Bolts & dowel pins 316 strain harden SS. 304 SS threaded locking plug for the 1.25"1 diameter dowel pins.
Q Have these fasteners been corrosion tested after locking process?
Has some surveillance or procedure documented results with respect to corrosion properties?
Is there a qualification program?
A No corrosion testing is performed because of extensive use in Europe. Approximately 4500 bolts have been installed in higher temperature and higher radiation fields than seen in this design.
No failures observed to date.
Q What is the material for the limiter keys, support blocks,
- flexures, and the thermal shield?
Have they been corrosion tested after peening?
A Limiter keys 304SS, support block 304SS, flexure XM-19, thermal shield SA-240 304 SS -
no peening will be used.
Q What is the method for ISI of new material?
Have they been proven by industry? What confidence do we have that they won't break?
The lower support block fasteners will be ultrasonically tested next refueling.
Future inspections will be incorporated into the ISI Program (VT).
Q Baffles - Westinghouse had problems with lateral flow, cross-flow through mechanical joints.
Westinghouse did some peening to impede flow. Have these peened baffles exhibited cracking? Has anyone (Westinghouse) looked at this?
A
- 1. No peening will be performed to lock the fasteners.
- 2. No baffle cross flow fuel damage observed at SONGS 1.
- 3. SONGS is an up-flow, not a down-flow design.
- 4.
SONGS does not have a high delta pressure drop across the baffle plates.
Q Is there a contingency plan to remove the thermal shield?
A Yes, Westinghouse is doing a feasibility study.
Also, Pacific Nuclear is doing preliminary engineering on cask preparations.
Q Why are neck down bolts used as fasteners in the support blocks?
A Higher fatigue strength - more flexibility in the joint lower stresses in the bolt (1" bolts and.881" shank).
Q Preloading will minimize stress in the flexure web area at operating temperatures and power levels.
Is the operating condition used 92% power and reduced Tavg, or rated conditions?
Will this impact the licensed power rating of SONGS 1 ?
A Reduced temperature conditions were used for preload of the flexure.
Preload will be established at 100% power with reduced Tavg.
The design stress criteria will. be met for 92% power and reduced Tavg.
Q How does moving the location of one flexure reduce loading?
A Analysis showed that moving the 21 degree flexure to 352 degrees reduced the stresses more than 10%.
Q If water from the spent fuel pool drained to the cavity, could water level drop below the top of fuel assemblies in the pool?
A The reactor cavity water level will be higher than the spent fuel pool water level.
Both the fuel transfer gate valve and the transfer gate will be closed. Hence, a double isolation is established.
Upon a postulated failure of the double isolation, and if the cavity water level should drop, approximately 31 of water will be maintained above the top of the fuel assemblies due to the presence of the wier wall.
CYCLE 12 REFUELING THERMAL SHIELD INSPECTIONS During the Cycle 11 support replacement, additional inspection ports will be added through the core barrel flange.
These inspection ports will allow camera access through ports in the flange when the core barrel is in the reactor vessel.
The planned inspections during the Cycle 12 refueling are as follows:
A)
Visually inspect the six upper flexures.
B)
Visually inspect the four upper limiter keys.
C) Visually inspect the six lower supports.
D)
Ultrasonically examine the lower support fasteners at all six support blocks.
E)
Visually inspect a representative sample of the surveillance capsules and baskets.
F)
Visually inspect the lower vessel head region for debris.
This inspection plan, in conjunction with the continued monitoring system, will enable the condition of the thermal shield to be assessed.