ML13311B149
| ML13311B149 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 03/21/1989 |
| From: | Trammel C Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8904050378 | |
| Download: ML13311B149 (80) | |
Text
0 tREG(&
UNID STATES Oa NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 March 21, 1989 LICENSEE:
SDUTHERN CALIFORNIA EDISON COMPANY FACILITY:
SAN ONOFRE UNIT NO. 1
SUBJECT:
SUMMARY
OF MEETING HELD ON MARCH 14, 1989 REGARDING THE BROKEN FASTENERS ON THE REACTOR VESSEL THERMAL SHIELD On March 14, 1989, the NRC staff and its consultant met with representatives of Southern California Edison (SCE) and its consultants to discuss SCE's recent visual inspection of the reactor vessel thermal shield.
Attendees are shown in Attachment 1. Viewgraphs shown at the meeting are contained in Attachment 2.
SCE's recent visual examination of the thermal shield disclosed three broken bolts on two of the six lower support blocks. SCE has proposed a license amendment that would provide for monitoring the condition of the thermal shield by neutron noise measurements and a loose parts monitor. It is SCE's position that repairs are not necessary at this time and the unit can be safely operated for the next 18 month fuel cycle.
The meeting was requested by the staff in a letter to the licensee dated March 3, 1989 which transmitted staff concerns regarding the proposal to defer repair until the next refueling outage. A earlier meeting was held on January 27, 1989.
No conclusion as to the acceptability of the licensee's proposal was reached at the meeting.
/Z Charles M. Trammell, Senior Project Manager Project Directorate V Division of Reactor Projects -
- III, IV, V and Special Projects Attachments:
- 1. Attendees
- 2. Viewgraphs CONTACT: C. Trammell, NRR 492-3121
Mr. Kenneth P. Baskin San Onofre Nuclear Generating Southern California Edison Company Station, Unit No. 1 cc Charles R. Kocher, Assistant Mr. Paul Szalinski, Chief General Counsel Radiological Health Branch James Beoletto, Esquire State Department of Health Southern California Edison Company Services Post Office Box 800 714 P Street, Office Bldg. 8 Rosemead, California 91770 Sacramento, California 95814 David R. Pigott Orrick, Herrington & Sutcliffe 600 Montgomery Street San Francisco, California 94111 Mr. Robert G. Lacy Manager, Nuclear San Diego Gas & Electric Company P. 0. Box 1831 San Diego, California 92112 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 Director Energy Facilities Siting Division Energy Resources Conservation &
Development Commission 1516 -
9th Street Sacramento, California 95814 Regional Administrator, Region V U.S. Nuclear Regulatory Commission 1450 Maria Lane, Suite 210 Walnut Creek, California 94596 RailgclHelhBac
ATTACHMENT 1 MEETING ATTENDEES March 14, 1989 NRC SCE C. Trammell K. Baskin M. Virgilio M. Medford C. Hinson R. Ornelas R. Lipinski R. Ashe-Everest L. Lois M. Motamed M. Hum C. Chiu R. Hermann A. Hernandez K. Wichman J. Murray H. Gray G. Knighton Other F. Huey R. Erickson, SDG&E T. Collins W. Hodges R. Kryter (Consultant, ORNL)
Westinghouse U. Dominicis N. Singleton D. Bhandari J. Goossen R. Perez R. Sterdis R. Schwirian K. Graham
TECHNICAL PRESENTATION ON ISSUES RELATED TO DEGRADED FASTENERS FOR THERMAL SHIELD SUPPORT BLOCKS AT SAN ONOFRE UNIT 1 Southern Callfornia Edison Westinghouse Electric Corporation MARCH 14, 1989
SAN ONOFRE UNIT 1 THERMAL SHIELD AGENDA MARCH 14, 1989
MORNING SESSION ----
INTRODUCTION KEN BASKIN/SCE (9:00 a.m.)
TECHNICAL OVERVIEW CHONG CHIU/SCE (9:05 a.m.)
ENGINEERING ANALYSIS CHUCK BOYD/-W JOHN GOOSSEN/-W DEV BHANDARI (9:10 a.m.)
INSPECTIONS BOB ASHE-EVEREST/SCE (11:15 a.m.)
LUNCH 12:00 Noon
SAN ONOFRE UNIT 1 THERMAL SHIELD AGENDA MARCH 14, 1989
AFTERNOON SESSION ----
ACCIDENT ANALYSIS DAVE DOMINICIS/-W (1:00 p.m.)
NEUTRON MONITORING NORM SINGLETON/-W (1:45 p.m.)
REPAIR CHONG CHIU/SCE (2:30 p.m.)
NUCLEAR INSTRUMENTATION SYSTEM ARMANDO HERNANDES/SCE (2:45 p.m.)
CONCLUSIONS 3:15 p.m.
SAN ONOFRE UNIT 1 THERMAL SHIELD AGENDA MARCH 14, 1989 Specific Questions & Answers Speaker Time Concerns 7, 1, 3, 4, Boyd 9:10 a.m.
Questions 4, 8 Questions 1, 6, 7 Goossen 9:10 a.m.
Concern 6 Bhandari 9:10 a.m.
Concerns 2, Questions 2, 3 Ashe-Everest 11:15 a.m.
Concern 1 Dominicis 1:00 p.m.
Concern 5, Question 9 Singleton 1:45 p.m.
Question 10, 5 Boyd/C. Chiu 2:30 p.m.
0)
OVERALL CONCERN NO. 7 The Westinghouse letter of February 12, 1988 to the licensee discusses three possible and progressively worse scenarios.
Given the lack of definitive data on the current condition of the bolts and pins, and the relatively qualitative monitoring system, the staff is not convinced that the continued operation of SONGS1 can be justified without repair.
RESPONSE
o The referenced letter was issued to summarize the POTENTIAL implications of the Haddam Neck thermal shield issue.
o At the time the letter was issued, no specific evaluation and analyses had been performed for SONGS1.
o Since that time, a number of proactive steps have been taken to address the potential concerns:
Analysis was performed to estimate the extent of degradation.
Inspections were performed to confirm the extent of degradation.
Analyses were performed to determine if significant further degradation is expected within one fuel cycle.
Stability limits were determined and compared to current and further degraded condition. Results confirm that shield will remain dynamically stable.
Analyses were performed to justify continued safe operation for an additional cycle.
Vibration and loose parts monitoring programs are being implemented.
Both Westinghouse and SCE agree that a repair can be deferred for one cycle without affecting safety.
OVERALL CONCERN NO. 1 The overall concerns are that the thermal shield may (a) vibrate and cause damage to the core barrel, (2) create sizable and numerous loose parts, (c) obstruct flow and cause flow maldistribution, or (d) cause fuel failure due-to deformation of the core barrel from thermal shield dislocation. The licensee has not analyzed transient or accident conditions with the thermal shield in either its current configuration or with progressively more damage.
RESPONSE
Approach taken was to demonstrate that the thermal shield will remain in place considering:
Current Condition Worst Credible Condition Design Basis Seismic Event Loose parts analysis was performed for reasonable expected parts (i.e., remnants of broken support block bolts).
Since thermal shield will remain in-place even for worst case conditions, no effect on flow distribution and hence no effect on accident analysis.
Analysis shows that even if worst credible degradation occurs, there is no impact on structural integrity of core barrel.
Have shown that even if shield were to implausibly drop, motion is limited by lower radial support keys.
Keys have been shown capable of withstanding impact.
As another level of conservatism, vibration and loose parts monitoring will be used to obtain "early warning" of any significant changes in support conditions.
WORST EXPECTED DEGRADED CASE (18 MONTH OPERATION)
BASED ON INSPECTION PLUS ANALYSIS o
124 DEGREE FLEXURE REMAINS INTACT o
0 DEGREE, 240 DEGREE AND 300 DEGREE BLOCKS DEGRADED Top BOLTS BROKEN DOWEL PIN IN PLACE LOWER BOLTS INTACT o
REMAINING BLOCKS INTACT
WORST CREDIBLE DEGRADED CASE ASSUME:
All Flexures Broken All Support BLock Bolts Broken All Dowel Pins "Loose' Upper Displacement Limiter Keys Worn Away All Blocks Still in Place BASIS:
Support Blocks Cannot Dislodge:
Geometrical Constraints Preclude Dislodging Support Blocks Will Not Break "Self Locking" Features in Shield/Blocks Relative Motion Between Shield and Block Much Less Than Width of Support Ledge Shield Does Not Go Unstable Even If All Bolts Broken
EVALUAT
- VERVIEW CASES CONDITIONS ANALYSIS CONCLUSION One flexture intact, flow instability Acceptable WORST EXPECTED three blocks degraded seismic analysis Consequences review all FSAR analyses WORST CREDIBLE All flextures and bolts flow instability Acceptable are broken seismic analysis Consequences*
review all FSAR analyses WORST CONCEIVABLE Thermal shield drop drop impact analysis Acceptable normal operation Consequences*
review all FSAR, DNB, and non-DNB events LOCA, SLB (preliminary)
Monitoring System will detect degradation prior to achieving this condition.
OVERALL CONCERN NO.
3 Many bolts and pins of the support blocks are already degraded and the rate of further degradation is based on engineering judgement rather than on facts.
RESPONSE
Predictions relative to further degradation are based on an analytical methodology which was developed to evaluate Haddam Neck situation.
Methodology successfully predicted no degradation at Zorita Plant.
Methodology successfully predicted extent of degradation at SONGS before inspection.
Hence, predic ted future behavior is based on a proven engineering analysis method rather than on engineering judgement. Furthermore, evaluations have been performed toldemonstrate continued safe operation if degradation occurs faster than expected.
OVERALL CONCERN NO. 4 In view of uncertainty regarding the bolts of the support blocks there is a possibility that as a result of vibration of the thermal shield the support blocks may slip out from their positions.
In the worst case the thermal shield could drop to the bottom of the reactor vessel.
RESPONSE
As explained in the response to Concern No. 2, the uncertainty has effectively been eliminated by the analysis, monitoring and consequences of failure evaluations.
The blocks cannot credibly slip out of position because of the safety mechanisms inherent in the design of the support system.
Analyses demonstrate that shield remains stable even for worst credible degraded conditions so loads would not increase in an unpredictable manner. Therefore, support block failure is not plausible.
It is not credible to postulate that thermal shield could fall.
Nevertheless, further analyses confirm that if shield were to fall, its downward motion would be restrained by the lower radial keys.
Hence shield cannot drop to the bottom of the reactor vessel.
Core Barrel Thermal Shield Core Barrel Thermal Shield Lower Lower Core Fuel Plate Fuel Plate
+ 110.85 3.05
++ ++
+ +
+
Support Block Support Block SONGS 1 HADDAM NECK DISTANCE FROM BOTTOM OF LOWER CORE PLATE TO TOP OF SUPPORT BLOCK
005
THERMAL SHIELD 8.00 HIGH 8.00 WIDE BEARING SURFACE 3.35
- 62. 134 RADIUS 0.1 BLOCK DEGRADED BLOCK ANALYSIS CRITICAL DIMENSIONS
FH 1.42 FH 62.134 RADIUS 0.616 Fv BLOCK DEGRADED BLOCK ANALYSIS APPLIED LOADS
- SPECIFIC CONCERN NO. 4 Will one cycle of operation with the "Worst Credible Degraded Case" cause or increase the probability of fatigue cracking of the core barrel to lower support weld?
RESPONSE
No, a fatigue analysis has been performed which shows a cumulative usage factor of less than 1.0 for operation in this condition.
In addition:
It should be noted that SONGS1 uses a "Secondary core support assembly" as an additional safeguard against catastrophic failure of the core barrel.
The potential downward motion of the core barrel is limited to a fraction of an inch in order to ensure that control rods maintain engagement.
- SPECIFIC CONCERN NO. 8 ASSUMING ALL BOLTS AND PINS IN SUPPORT BLOCKS ARE FAILED WHAT IS THE EFFECT OF THIS ON THERMAL SHIELD LIFTOFF DURING A LOCA?
RESPONSE
o FOR THE WORST CREDIBLE CONDITION WHERE ALL BOLTS ARE FAILED, THE SIX FLEXURE BLOCKS ABOVE THE SHIELD (2.5" ABOVE THE TOP OF THE THERMAL SHIELD) CAN RESIST THE UPLIFT DUE TO A DOUBLE ENDED BREAK AT THE VESSEL INLET NOZZLE.
THE ADDITIONAL DROP HEIGHT DUE TO THIS POSTULATED LIFT DOES NOT ALTER THE CONCLUSIONS OF THE THERMAL SHIELD DROP ANALYSIS.
o FOR THE UNEXPECTED CASE WHERE THE DOWEL PINS ARE ALSO FAILED, THE SAME RESULT IS OBTAINED.
FIGURE REACTOR VESSEL ANO INTERNALS "CAD vtaT SyTTE
- STRUMENIT TUGE oaIvt S14ArT SHROUs TUsc r
seCAD LIFTING LUGS UPP(Rt SUPPORT PLATC C
A*
OUSA SEAL y*TELESCOPmNC SUPPORT COLUMN MCAD-VtSSCL MATING SURFACC t0-Amos MEACTQA VESSEL FLANGE ftCAD-vCSSEL*Ce4[C SAaaCL AussNIEtT 00Me PLATE COR(
gARREL fLbe4(
SUPPORT 4--.COLUMN CONTROL R00
)u o
w e
-t L E T O Z Z UPPER CORE PLATE A TAXIMAL SNICSLD AAiA UPP0R COME PLATCOEAa sSCl FurL ASSEuggy TIMERMAL CONTROL A400EL AXIAL SAFFLES TiESSEL LOWER CORE PLATE ACCESS Pear
_TNERMAL S"4CLD COME SUPPORT CASTING I
EA UPR LOWER COME RADIAL SUPPOor mITERaNDIATE DIFrUSta PLATE CLAD
J I
K I
SPECIFIC CONCERN NO. 1
- 1.
BOLT FATIGUE EVALUATION THE TOP BOLT FATIGUE EVALUATION WAS PERFORMED ON A BEST ESTIMATE BASIS TO EXPLAIN THE CURRENT OBSERVED CONDITION OF THE THERMAL SHIELD LOWER SUPPORT BLOCKS.
THESE ANALYSES WERE EXTENDED TO PREDICT THE EXPECTED PROGRESSION OF THE THERMAL SHIELD DEGRADATION OVER THE NEXT FUEL CYCLE (18 MONTHS)
SPECIFIC CONCERN NO. lA WHY WAS LOSS OF BOLT PRELOAD FROM RADIATION-ASSISTED RELAXATION NOT CONSIDERED?
RESPONSE
o IT WAS CONCLUDED THAT THE SONGS 1 LOWER SUPPORT BOLTS HAVE BEEN AFFECTED VERY LITTLE BY IRRADIATION AND BECAUSE OF THIS, RADIATION-ASSISTED RELAXATION IS NOT CONSIDERED.
o RADIATION-ASSISTED RELAXATION WAS NOT CONSIDERED TO BE A CONCERN BASED ON THE INFORMATION OBTAINED ON THE HADDAM NECK THERMAL SHIELD LOWER SUPPORT BOLTS.
o ONE BOLT REMOVED FROM THE HADDAM NECK LOWER SUPPORTS WAS SUBJECTED TO A TENSILE TEST.
o RESULTS SHOWED THE YIELD STRESS TO BE 77 KSI.
o THE SPECIFIED DRAWING YIELD STRESS FOR THIS BOLT WAS 65 TO 80 KSI.
o THE YIELD STRESS OF STAINLESS STEEL TENDS TO INCREASE WITH EXPOSURE TO IRRADIATION.
o THIS WOULD INDICATE THAT THE BOLT WAS AFFECTED VERY LITTLE BY IRRADIATION.
o THE LOWER SUPPORT BLOCKS AT HADDAM NECK ARE APPROXIMATELY 3 INCHES BELOW THE LOWER CORE PLATE AND BLOCKS AT SONGS 1 ARE APPROXIMATELY 11 INCHES BELOW THE LOWER CORE PLATE.
o THE HADDAM NECK PLANT HAS BEEN IN OPERATION LONGER THAN SONGS 1 (15 EFFECTIVE FULL POWER YEARS vs. 11 EFFECTIVE FULL POWER YEARS.
o CONSIDERING THE ABOVE, IT WAS CONCLUDED TO BE UNNECESSARY TO ASSUME RADIATION-ASSISTED RELAXATION.
Core Barrel Thermal Shield Core Barrel Thermal Shield LowerLower CoreCore Fuel Plate FulPlate FuelPlt 10.853.05
++
+ + ++
Supportt loer SupSuppor BBlock SONGS 1 HADDAM NECK DISTANCE FROM BOTTOM OF LOWER CORE PLATE TO TOP OF SUPPORT BLOCK
SPECIFIC CONCERN NO. lB WHY WAS A SECTION III FATIGUE ANALYSIS NOT PERFORMED?
RESPONSE
o THE FATIGUE EVALUATION PERFORMED WAS NOT INTENDED TO CONFIRM THE DESIGN OF THE LOWER SUPPORT BOLTS, BUT TO PREDICT THE LOCATIONS OF EXPECTED BOLT FAILURES.
o A SECTION III FATIGUE ANALYSIS IS FORMULATED TO ENSURE THAT CRACKS WILL NOT INITIATE IN A STRUCTURE OVER THE DESIGN LIFE OF THE STRUCTURE.
o THIS TYPE OF ANALYSIS WOULD NOT NECESSARILY PREDICT THE FAILURE OF A COMPONENT EVEN IF THE FATIGUE USAGE OBTAINED WAS GREATER THAN 1.0.
o THE PURPOSE OF THE FATIGUE EVALUATION PERFORMED WAS TO PREDICT AT WHAT SUPPORT BLOCKS BOLT FAILURE WAS EXPECTED TO OCCUR.
o BOLT FAILURE WAS PREDICTED AT THE 0 DEGREE AND 240 DEGREE BLOCKS AND POSSIBLY AT THE 300 DEGREE BLOCK.
o THIS MATCHES QUITE WELL WITH THE OBSERVED DEGRADATION AT THE 0 DEGREE AND 240 DEGREE BLOCKS.
o THEREFORE, THE USE OF A BEST ESTIMATE FATIGUE ANALYSIS WAS CONSIDERED APPROPRIATE.
SPECIFIC CONCERN NO. 1C JUSTIFY THE USE OF FAILURE CURVES IN LIEU OF LOWER BOUND (OR "DESIGN") CURVES FOR THE FATIGUE ANALYSIS.
RESPONSE
o SINCE THE PURPOSE OF THE FATIGUE ANALYSIS WAS TO PREDICT THE LOCATIONS OF EXPECTED BOLT FAILURE, IT WAS APPROPRIATE TO USE THE FAILURE CURVE RATHER THAN THE DESIGN CURVE.
o THE FAILURE CURVE IS USED TO DEFINE THE POINT OF FAILURE OF THE BOLTS IF A CUF OF 1.0 IS REACHED.
o THE USE OF THIS CURVE IS BASED ON CORRELATING THE FATIGUE EVALUATION OF THE TOP BOLTS DURING HOT FUNCTIONAL TESTING WITH THE POST HOT FUNCTIONAL BOLT INSPECTION RESULTS.
o A FATIGUE EVALUATION, USING THE FAILURE CURVE, OF THE BOLTS WAS ADJUSTED SO A CUF OF JUST BELOW 1.0 (.99) WAS OBTAINED FOR THE BOLTS AT THE HIGHEST LOADED BLOCKS (00 AND 1800) CONSIDERING 21 DAYS OF HOT FUNCTIONAL TESTING.
0 THE ASSUMPTION OF CUF JUST BELOW 1.0 WAS MADE SINCE THE BOLTS WERE REMOVED AND INSPECTED FOLLOWING THIS TEST AND NO EVIDENCE OF CRACKING WAS DISCOVERED.
o ONCE THIS CORRELATION WAS MADE, FATIGUE ANALYSES WERE PERFORMED TO DETERMINE THE EXPECTED BLOCK LOCATIONS OF BOLT FAILURES AT VARIOUS TIMES IN THE OPERATIONAL HISTORY OF THE PLANT.
'SPECIFIC CONCERN NO. 1D EXPLAIN THE BASIS FOR THE USE OF A STRAIN CONCENTRATION FACTOR IN THE CONTEXT OF ESTABLISHED PROCEDURES FOR CUF CALCULATIONS.
RESPONSE
o WHEN THE ALTERNATING STRESS IS ABOVE THE MATERIAL CYCLIC YIELD STRESS, A STRAIN CONCENTRATION FACTOR SHOULD BE APPLIED RATHER THAN A STRESS CONCENTRATION FACTOR BECAUSE THE ELASTIC STRESS CONCENTRATION FACTORS ARE NO LONGER VALID.
o SINCE THE FAILURE CURVE WAS USED IN LIEU OF THE DESIGN CURVE, IT WAS APPROPRIATE TO USE AN INCREASED CONCENTRATION FACTOR WHEN THE MATERIAL CYCLIC YIELD WAS EXCEEDED.
o A STRAIN CONCENTRATION FACTOR IS APPLICABLE BECUASE THE FAILURE CURVE IS BASED ON BEST ESTIMATE STRAINS. THE DESIGN CURVE HAS A FACTOR OF SAFETY BUILT IN.
o THE FOLLOWING CURVE SHOWS HOW STRESS AND STRAIN CONCENTRATION FACTORS VARY AS A FUNCTION OF Sn/Sy.
o THIS CURVE WAS USED TO ACCOUNT FOR THE INCREASE IN LOCAL STRAIN WHEN THE MATERIAL CYCLIC YIELD WAS EXCEEDED.
I.
e 14 12 10 Fil.
- Stress and strain concentrade facters fO blunt astched CT specime*ns. Curves censtructed from plane strain finite element 0**
ltiones using cyclIc strese*trate curve of M0 stee.
REFERENCE -"ELASTIC-PLASTIC ANALYSIS OF BLUNT NOTCHED CT SPECIMENS AND APPLICATIONS", BY W. K. WILSON (JOURNAL OF PRESSURE VESSEL TECHNOLOGY).
SPECIFIC CONCERN NO. 1E NB-3232.3(C) RECOMMENDS USE OF A FATIGUE STRENGTH REDUCTION VALUE OF NOT LESS THAN 4.0.
JUSTIFY USE OF A SMALLER VALUE.
RESPONSE
0 THE USE OF A SMALLER VALUE IS BASED ON CURRENT LITERATURE AND TESTS.
o AN ELASTIC CONCENTRATION FACTOR OF 4.0 WAS USED AT THE THREADS FOR STRESS DUE TO A TENSION LOAD AND THIS VALUE IS BASED ON THE ASME BOILER & PRESSURE VESSEL CODE SECTION III PARAGRAPH NG-3232.3(C).
o OTHER SOURCES SHOW STRESS CONCENTRATION FACTORS IN THE THREADS TO BE APPROXIMATELY 4.0 FOR TENSION (M. HETEHYI).
o FOR STRESSES DUE TO A BENDING LOAD, THE ELASTIC CONCENTRATION FACTOR USED WAS 3.5. BASED ON A COMPARISON OF CONCENTRATION FACTORS FOR A GROOVED SHAFT
("PETERSON", FIGURES 31 AND 49) WHERE A 4.0 IN TENSION CORRESPONDS TO A 3..5 IN BENDING FOR.THE SAME GEOMETRY.
o WHEN THE MATERIAL IS CYCLING IN THE ELASTIC RANGE, A NOTCH SENSITIVITY FACTOR CAN BE USED TO LOWER THE ELASTIC STRESS CONCENTRATION FACTOR.
o IN THE CASE OF THE 316 SS BOLTS, A MATERIAL NOTCH SENSITIVITY FACTOR OF q =.6 (BASED ON TESTS) IS APPLICABLE.
o THE RESULTING FATIGUE NOTCHED REDUCTION FACTOR Kf, IS DETERMINED BY:
Kf = 1 + q (Kt -
- 1)
= 2.8 FOR TENSION
= 2.5 FOR BENDING WHERE:
Kf = FATIGUE NOTCHED REDUCTION FACTOR Kt = ELASTIC STRESS CONCENTRATION FACTOR
Top of nut Conventionot nut
.Kma,,
- 3.8 5 Bottom of nut 0
1 2
3 4 Kt Fig. 6.41. Variation in Thread. Root Stress Concentration Factors Through out the Engaged Threads of a Conventional Nut and Bolt" Ref. "TMheory and Design of modein Pressure Vessels", J.F. Harvey (Ref. 74, M. Hetenyi, "A Photoetastic Study of Bolts and Nut Fasters#, AME Jornal of Applied Mechanis,June, 1943)
WHEN DID BOLT AND PIN DEGRADATION OCCUR?
RESPONSE
0 IT IS ESTIMATED BASED ON THE FATIGUE EVALUATION OF THE LOWER SUPPORT TOP BOLTS THAT THE BOLT DEGRADATION INITIATED WHEN:
- 1)
ONLY ONE FLEXURE WAS STILL INTACT (AT 1240).
- 2)
SUFFICIENT WEAR HAD OCCURRED AT THE LIMITER KEYS.
o THIS WOULD HAVE BEEN SOMETIME DURING THE LAST 5-6 YEARS OF EFFECTIVE FULL POWER OPERATION BASED ON THE FACT THAT TWO FLEXURES WERE INTACT PRIOR TO THAT POINT IN TIME.
CIFIC CONCERN NO.
7 WHAT PRELOAD IS CURRENTLY ASSUMED?
RESPONSE
o ASSUMED PRELOAD PRELOAD/BOLT @ 70oF 10,970 LB.
PRELOAD 2 BOLTS @ 600 0 F
= 19,690 LB.
PRELOAD REDUCTION FOR THERMAL
= 17,330 LB.
STRESS RELAXATION (12%)
ESTIMATED FORCE REQUIRED TO
= 4,980 LB.
CLOSE.008 INCH GAP EFFECTIVE PRELOAD 2 BOLTS
= 12,350 LB. (-1S)
= 10,150 LB. (-2S) 7/7
.005
.008
OVERALL CONCERN NO. 6 The licensee did not provide sufficient information regarding seismic evaluation to enable the staff to formulate a meaningful opinion. More specific information regarding methodology used in computing impact loads and the techniques used for computing seismic loads at various locations would be required.
RESPONSE
o A detailed discussion follows, but in summary:
Impact loads during a seismic event were determined by performing a non-linear time history seismic analysis.
Nonlinearities due to structural gaps were included in the model.
The analysis was performed using standard Westinghouse seismic analysis methods and computer codes.
Spectrum of time history seismic input to the model (at the RPV supports) enveloped spectrum of SONGSI Modified Hausner spectrum.
Linear response spectrum seismic analysis was performed to develop loads in remaining intact flexure and support blocks.
METHODOLOGY IMPACT LOAD EVALUATIONS - PERFORM NONLINEAR TIME HISTORY ANALYSIS FOR THE FULLY DEGRADED CONDITION (ALL SIX BLOCKS ASSUMED DEGRADED AND NO FLEXURE INTACT)
- SIMPLIFIED NONLINEAR SYSTEM MODEL
- PERFORM MODAL ANALYSIS ON THE SYSTEM MODEL (FUNDAMENTAL BEAM MODE FREQUENCIES OF THERMAL SHIELD AND CORE BARREL BE CONSISTENT WITH THOSE OF 3-D MODEL)
-OBTAIN SYNTHESIZED TIME HISTORY ACCLERATIONS (SPECTRA OF SYNTHESIZED TIME HISTORY ACCLERATIONS SHOULD ENVELOPE 0.67g MODIFIED HAUSNER SPECTRA FOR SONGS 1 AT VESSEL SUPPORTS)
-PERFORM TIME HISTORY ANALYSIS USING MODAL SUPERPOSITION TECHNIQUE OF 'WECAN' CODE (USE 4 PERCENT STRUCTRAL DAMPING)
-RESULTS INCLUDE IMPACT LOADS AND NODAL DISPLACEMENTS.
Vessel Thermal Shield Core Barrel 5 Mating Flange Elevation 1A 03 Limiter Key Elevation 4
(Ic) it Lower Support Elevation Dynamic Impact Element Rotary Spring Hydrodynamic Mass Simplified System Model Figure 1
ILD l.a 5
I
C, I
I a
I U
Table 1 Frequency Canparison Conponent Beam Mode Frequency Siplified Model 3-D Detail Model Thernal Shield 2.59 Hz 2.43 z
Core Barrel 7.4 oz
.30 Hz
ACCELERATION TIME HISTORY
.6
.4 A
C C
E L2 E
R A
T
- 6.
I 0
N G G-.4
-.6
-.8 1
2 3
4 S
6 7
8 9
1 TIME (SECS)
SONGS 1 4% HORIZONTAL SPECTRA
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- -A Figure 7.4-5.
Spectra Primary Horiz-East East (E1.
10.93')
2ss2-a131an10 7-25
SAN ONOFRE SEISMIC ANALYSIS 3.2 3
2.8 Analysis Spectra (4% Damping) 2.6 2.4 2.2 San Onofre Spectra (4% Damping) 2 1.8 1.6 1.4 1.2 1
0.8 O. 6 0
4 8
12 16 20 24 28 FREQUENCY, Hz.
Figure 3a
C Mass LGap Mass (Optional) 1J K
0 Direction of Uniaxial Behavior Three-dimensional dynamic element Force Force Force Gap Gap K
U -U U -U U -U K
K Gap> 0.0 Gap =0.0 Gap <0.0 Spring force in dynamic element as a function of the initial gap and the relative displacement
Relative displacement Unit vector along Separation
=vector of node J
.the direction of
+Initial gap with respect to node I uniaxial behavior
= (UX(J) -
UX(I)) (X(J) -
X(I))
L
+ (UY(J).- UY()
(
-.Y(I))
L
+ (UZ(J) - UZ(I)) (Z(J) -
ZMI))
+ (Initial gap)
Impact Force = K x Separation For Limiter Key K = 0.33 x 106 lb/in CD = 1000.0 lb.sec/in (5% of critical damping)
Gap = 0.250 in.
Relative displacement Unit vector along Separation
=
vector of node J the direction of
+ Initial gap with respect to node I uniaxial behavior
= (UX(J) -
UX(I)) (X(J) -
X(I))
L
+ (UY(J) -
UY(I))
Y())
+ (UZ(J) -
UZ(I)) (Z(J) -
Z(I))
+ (Initial gap)
METHODOLOGY FLEXURE AND SUPPORT BLOCK LOADS - PERFORM LINEAR ELASTIC RESPONSE SPECTRUM ANALYSIS
-CASES INVESTIGATED o ONE FLEXURE WITH THREE SUPPORT BLOCKS (WORST EXPECTED CASE) o NO FL.EXURE WITH THREE SUPPORT BLOCKS o FULLY DEGRADED CASE
-LINEARIZED SYSTEM MODEL
-SIMULTANEOUS APPLICATION OF SEISMIC EXCITATIONS IN BOTH HORIZONTAL DIRECTIONS
-RESULTS INCLUDED FLEXURE LOADS AND BEAM MODE DISPLACEMENTS.
-FOR SUPPORT BLOCK LOADS (RAD. & VERT.), RESULTS OF 3-D FIV BEAM MODE ANALYSES WERE USED TO CALCULATE SEISMIC LOADS. (RATIOING METHOD)
Vessel Thermal Shield Core Barrel
- 2.
Mating Flange Elevation 7
6 Thermal Shield Flexure 4
(ic01 I' Lower Support Elevation Linear Spring Rotary Spring Hydrodynamic Mass Linearized Simplified System Model
SUMMARY
OF MAXIMUM SEISMIC LOADS Component Value Description Limiter Keys Ft = 54,515 (lbf)
Impact Load per key Thermal Shield Flexures Fr = 40,940 (lbf)
-Radial Direction Ft = 34,400 (lbf)
Tangential Direction Thermal Shield Support Block Loads Case a) All Six Blocks Degraded, No Flexure Block Location Fr (lbs)
Fvert. (lbf)
(Degrees)
.0 26,840 300 53,640 240 26,812 180 26,820 120 53,640 60 26,800 Case b) Three Blocks Assumed Degraded One Flexure Intact No Flexure Block Location Fr (lbs)i Fvert. (lbs)
Fr (lbs)
Fvert (lbs) 0
.0 3,507 0
21,448 300 0
0 0
31,517 240 0
3,507 0
21,448 180 8,925 56,350 2,035 22,549 120 0
0 8,048 122,530 60 8,925 56,350 2,035 22,549 The maximum radial loads during horizontal excitations is 52,760 lbs. which is conservatively assumed to be resisted by one block in compression.
Consequently, the core barrel shear stress is less than 2000 psi.
vertical seismic excitations, 1g loads should be considered.
3582s/0577s/020389: 10
OVERALL CONCERN NO. 2 THE LACK OF UT INCREASES UNCERTAINTY REGARDING THE NUMBER AND STATE OF THE DAMAGED BOLTS AND PINS AND THE STRUCTURAL INTEGRITY OF THE CORE BARREL.
RESPONSE
o UT WOULD GIVE MORE DEFINITIVE INFORMATION REGARDING SOME OF THE FASTENERS.
NOT ALL OF THE FASTENERS ARE ACCESSIBLE WITH AVAILABLE UT TOOLS.
o INTERMEDIATE DIFFUSER PLATE o
ANALYSES HAVE BEEN PERFORMED TO DEMONSTRATE SAFE OPERATION EVEN IF ALL BOLTS ARE FAILED AND PINS ARE LOOSE.
SPECIFIC CONCERN NO. 2 DURING THE 1988 REFUELING OUTAGE A LIMITED VISUAL INSPECTION WAS PERFORMED BASED ON THE AVAILABLE ACCESS.
BY USING A DIFFERENT CAMERA SYSTEM OR REMOVING MORE FUEL ASSEMBLIES COULD:
(A) ADDITIONAL SURFACES OF THE COMPONENTS BE INSPECTED OR (B) ADDITIONAL ACCESS BE PROVIDED TO THE LOWER THERMAL SHIELD SUPPORT BLOCKS, LIMITER KEYS, OR THE BARREL RADIAL SUPPORT KEYS?
RESPONSE
o THE OPTIMUM CAMERA SYSTEM WAS UTILIZED o
REMOVING ADDITIONAL FUEL WOULD NOT HAVE PROVIDED ADDITIONAL ACCESS
SPECIFIC CONCERN NO. 2 (continued)
AREAS INSPECTED o
AREAS INSPECTED BETWEEN REACTOR VESSEL AND CORE BARREL TOP AND BOTTOM A.
FLEXURES -
5 OF 6 B.
IRRADIATION SPECIMEN TUBES -
7 OF 8 C.
LOWER THERMAL SHIELD SUPPORT BLOCKS -
4 OF 6 o
INSPECTED INSIDE 3 IRRADIATION SPECIMEN GUIDE TUBES o
WE REMOVED SUFFICIENT FUEL ASSEMBLIES TO INSPECT ALL AREAS A.
INSPECTED THE BACK SIDE OF ALL SIX LOWER THERMAL SHIELD SUPPORT BLOCKS B.
INSPECT THE BACK SIDE OF ALL SIX FLEXURES o
INSPECTED THE BOTTOM OF THE REACTOR VESSEL
SPECIFIC CONCERN NO. 3 WAS THE CORE BARREL WELD INSPECTED DURING THE 1988 REFUELING OUTAGE?
IF SO, EXTENT OF EXAMINATIONS.
RESPONSE
o NO INSPECTION WAS PERFORMED o
INSPECTION NOT NEEDED BASED ON CY EXPERIENCE o
INSPECTION PLANNED FOR CYCLE XI REFUELING AS PART OF THE INSERVICE INSPECTION PROGRAM o
FATIGUE ANALYSIS HAS BEEN PERFORMED WHICH SHOWS A CUMULATIVE USAGE FACTOR OF LESS THAN 1.0.
-. "-8 TAPPED LI FTING HOLES (4 LOCATIONS)
SPECIMEN TUBE SPECIMEN TUBE SPECIMEN BASKET EXPANSION JOINT FLEXURE (TYP 6 PLACES)
LIMITER KEY (TYP 4 PLACES)
THERMAL SHIELD SUPPORT BLOCK (TYP 6 PLACES)
L I-z 0.00 0
0,
c-_
C A,
I~. p~
~UC)L ri91I.LCCVTO L.
'LI2C71 1-fCA-euKN
EFFECTS ON SAFETY ANALYSIS OF A POSTULATED FAILED THERMAL SHIELD AT SAN ONOFRE UNIT 1 D. P. DOMINICIS WESTINGHOUSE
BASIS o
WESTINGHOUSE ENGINEERING ANALYSIS HAS SHOWN THAT THERMAL SHIELD WILL REMAIN IN PLACE EVEN WITH ALL BOLTS FAILED o
POSTULATED ACCIDENTS WITH DROPPED THERMAL SHIELD IS NOT CREDIBLE o
LOW PROBABILITY OF SHIELD DROPPING COMBINED WITH LOW PROBABILITY OF ACCIDENT OCCURRING o
FALLEN SHIELD DOES NOT INITIATE A TRANSIENT o
NEUTRON NOISE AND LOOSE PARTS MONITORING WILL DETECT THERMAL SHIELD DEGRADATION PRIOR TO DROPPING o
DESIGN BASIS ASSUMPTIONS USED WITH CREDIT TAKEN FOR CYCLE SPECIFIC CONDITIONS o
USE OF BEST ESTIMATE ASSUMPTIONS COULD BE USED TO SHOW ADDITIONAL MARGIN
ANALYSIS OF DROPPED OR TILTED THERMAL SHIELD NORMAL OPERATION ANALYSIS HAS BEEN PERFORMED OF HYDRAULIC EFFECTS OF SHIELD TILTING OR DROPPING o
TILTED CASE o
0.2 PSI INCREASE IN DOWNCOMER PRESSURE DROP o
NO CHANGE IN THERMAL DESIGN FLOW o
DROPPED CASE o
SHIELD DROPS ONTO RADIAL KEYS o
1.25 PSI INCREASE IN LOWER PLENUM PRESSURE DROP o
NO CHANGE IN THERMAL DESIGN FLOW o
SOME FLOW REDISTRIBUTION IN CORE INLET 7% LOWER FLOW IN PERIPHERAL ASSEMBLIES 6% HIGHER FLOW IN CENTER ASSEMBLIES ADEQUATE CORE FLOW EXISTS EVEN IN THE CASE OF DROPPED SHIELD SUCH THAT CORE LIMITS ARE NOT VIOLATED FOR NORMAL OPERATION
ANALYSIS APPROACH FOR DROPPED OR TILTED SHIELD FOR ACCIDENT CONDITIONS o
IN RESPONSE TO STAFF CONCERNS, EVALUATIONS AND ASSESSSMENTS HAVE BEEN PERFORMED TO ADDRESS THE EFFECTS ON THE ACCIDENT ANALYSES OF THERMAL SHIELD TILTING OR DROPPING o
NON-LOCA TRANSIENTS o
DNB ACCEPTANCE CRITERIA o
NON-DNB ACCEPTANCE CRITERIA o
WORST CASE IAC EVALUATION FOR LARGE BREAK o
SMALL BREAK EVALUATION ON BASIS OF SIGNIFICANT AVAILABLE MARGIN o
DROPPED SHIELD IS THE MOST LIMITING CASE O
EVALUATIONS USING DESIGN BASIS ASSUMPTIONS WITH CREDIT TAKEN FOR CYCLE SPECIFIC CONDITIONS AS NECESSARY TO MEET SAFETY CRITERIA o
CONCLUSION IS THAT SUFFICIENT MARGIN EXISTS EVEN WITH THE DROPPED SHIELD TO MEET THE ACCEPTANCE CRITERIA OF ALL ACCIDENT ANALYSIS
OVERALL EFFECTS OF TILTED OR DROPPED THERMAL SHIELD o
TILTED SHIELD CASE o
NO IMPACT ON CORE FLOW DISTRIBUTION o
NEGLIGIBLE CHANGE TO VESSEL DELTA-P o
NO EFFECT ON LOOP FLOWS, TEMPERATURES o
NEGLIGIBLE IMPACT ON SAFETY ANALYSIS o
BOUNDED BY DROPPED CASE o
DROPPED SHIELD CASE o
CORE FLOW DISTRIBUTION EFFECT o
7% DECREASE IN PERIPHERY o
6% INCREASE IN INNER ASSEMBLIES o
FLOW RECOVERS IN 3-4 FT. CORE ELEVATION o
NO EFFECT ON THERMAL DESIGN FLOW o
NO EFFECT ON RCS FLUID TEMPERATURES o
DELTA-P INCREASE OF 1.25 PSI IN LOWER PLENUM o
NO IMPACT TO REACTOR TRIP OR ESF FUNCTIONS o
NO IMPACT ON ABILITY OF RODS TO SCRAM o
LOCALIZED EFFECT ON CORE
ASSESSMENT OF NON-LOCA TRANSIENTS ANALYSES UNAFFECTED BY CORE INLET FLOW MALDISTRIBUTION o
LOSS OF NORMAL FEEDWATER/STATION BLACKOUT o
FEEDLINE BREAK o
BORON DILUTION (MODES 3-6) o STEAMLINE BREAK MASS/ENERGY o
LOSS OF LOAD/TURBINE TRIP o
RCCA BANK WITHDRAWAL AT POWER ANALYSES AFFECTED BY CORE INLET FLOW MALDISTRIBUTION o
FUEL/CLADDING TEMPERATURE CRITERIA o
RCCA EJECTION
- o.
LOCKED ROTOR 0
DNBR RELATED CRITERIA o
UNCONTROLLED BANK WITHDRAWAL FROM SUBCRITICAL CONDITION o
LOSS OF FLOW o
STEAMLINE BREAK CORE RESPONSE o
'DROPPED ROD o
STARTUP OF AN INACTIVE LOOP o
BORON DILUTION (MODES 1 AND 2) 0 UNCONTROLLED BANK WITHDRAWAL AT POWER o
LOSS OF LOAD/TURBINE TRIP 0
ADDITION OF EXCESSIVE FEEDWATER o
LARGE LOAD INCREASE
NON-LOCA ANALYSES UNAFFECTED BY INLET FLOW MALDISTRIBUTION NO CHANGE IN INITIAL CONDITIONS USED IN ANALYSIS ANALYSIS RESULTS DEPENDENT ON GROSS RESPONSE OF CORE AND RCS o
POINT KINETICS USED TO CALCULATE POWER o
CORE AVERAGE TEMPERATURE IS NOT AFFECTED SINCE TOTAL CORE FLOW IS UNCHANGED o
LOOP FLOW CONDITIONS ARE UNCHANGED, THUS SECONDARY SIDE HEAT TRANSFER IS NOT AFFECTED
.4 FUEL/CLADDING TEMPERATURE CRITERIA ANALYSES o
FLOW MALDISTRIBUTION DUE TO DROPPED SHIELD RESULTS IN SLIGHTLY HIGHER TEMPERATURE IN SOME CORE REGIONS WITH REDUCED FLOW o
METHODOLOGY OF THE ANALYSIS ASSUMES DNB OCCURS AT INITIATION OF THE TRANSIENT o
POST-DNB HEAT TRANSFER COEFFICIENT RELATIVELY INSENSITIVE TO EXPECTED CHANGES IN BULK COOLANT TEMPERATURE AND FLOW o
AVAILABLE SENSITIVITY STUDIES INDICATE AN INSIGNIFICANT INCREASE IN FUEL AND CLADDING TEMPERATURES o
THERE EXISTS MORE THAN SUFFICIENT MARGIN TO ACCOMMODATE SLIGHT EXPECTED INCREASE IN FUEL AND CLADDING TEMPERATURES o
NO SIGNIFICANT IMPACT TO ANALYSIS RESULTS DUE TO DROPPED THERMAL SHIELD
DNBR RELATED ANALYSES o
DNBR AFFECTED BY CORE INLET FLOW DISTRIBUTION o
EXPECTED FLOW DISTRIBUTION o
7% REDUCTION IN PERIPHERY o
6% INCREASE IN CENTER ASSEMBLIES o
NO CHANGE IN THERMAL DESIGN FLOW o
BOUNDING ASSUMPTIONS USED IN EVALUATION o
7%.REDUCTION IN FLOW TO ALL FUEL ASSEMBLIES o
AVAILABLE MARGIN USED TO OFFSET FLOW REDUCTION
LIMITING DNB TRANSIENTS MEET DESIGN LIMITS CORE THERMAL SAFETY LIMITS REMAIN UNCHANGED LIMITING TRANSIENTS NOT PROTECTED BY CORE LIMITS EVALUATED AND SHOWN TO MEET DESIGN LIMITS LOSS OF FLOW LOCKED ROTOR ROD MISALIGNMENT ROD WITHDRAWAL FROM SUBCRITICAL STEAMLINE BREAK EVALUATION ASSUMPTIONS o
THERMAL DESIGN FLOW BASED ON ACTUAL SG TUBE PLUGGING LEVEL (15.2% VS. 20% ANALYSIS VALUE)
- o.
CYCLE 10 LIMITING AXIAL POWER SHAPES o
ENGINEERING HOT CHANNEL FACTORS BASED ON AS BUILT FUEL DATA o
FDHN WITH UNCERTAINTY BASED ON CYCLE 10 PEAK ROD
ACTUAL DNBR PENALTY EXPECTED TO BE SMALL o
NO CHANGE TO THERMAL DESIGN FLOW o
COMPLETE FLOW RECOVERY EXPECTED AT 30% -
40% CORE ELEVATION o
CORRESPONDING DNBR PENALTY EXPECTED TO BE ON THE ORDER OF 2-3%
o CONSISTENT WITH PREVIOUS EXPERIENCE WITH DNBR PENALTIES ASSOCIATED WITH FLOW MALDISTIBUTION o
RCS FLOW ANOMALY o
THINC 4 ANALYSIS (WCAP-8054)
CONCLUSIONS RELATED TO DNBR CRITERIA NON-LOCA TRANSIENT ANALYSES o
DROPPED THERMAL SHIELD CAUSES A CORE INLET FLOW MALDISTRIBUTION o
DNBR DESIGN BASES ARE MET o
CONSERVATIVE CORE INLET FLOW DISTRIBUTION o
CYCLE 10 CORE CONDITIONS o
ACTUAL DNBR PENALTY EXPECTED TO BE SMALL
EFFECT OF DROPPED THERMAL SHIELD ON LOSS OF COOLANT ACCIDENTS o
PRELIMINARY ANALYSIS RESULTS FOR LARGE BREAK o
LARGE BREAK REANALYZED AS PART OF CYCLE 10 RELOAD TO BOUND 20% SG TUBE PLUGGING o
LIMITING BREAK PEAK CLAD TEMPERATURE FOR REDUCED TAVG OPERATION -
2260.3 0 F (DECLG, CD= 0.8 AS COMPARED TO ACCEPTANCE CRITERIA OF 2300 F o
LIMITING BREAK REANALYZED USING SLIGHT INCREASE (1.25 PSI) IN LOWER PLENUM FLOW RESISTANCE o
INSIGNIFICANT INCREASE IN PCT (<100F) FOR LARGE BREAK o
EVALUATION FOR SMALL BREAK o
ASSESSMENT PERFORMED CONSIDERING THE EFFECT OF SLIGHT INCREASE IN FLOW RESISTANCE IN LOWER PLENUM REGION o
FLOW RESISTANCE IN THE LOWER PLENUM AFFECTS DEPTH OF CORE UNCOVERY AND ALSO DELAYS CORE RECOVERY o
ANY POSSIBLE INCREASE IN PCT WILL BE INSIGNIFICANT IN COMPARISON TO MARGIN AVAILABLE (> 1400 F) 0 SUFFICIENT MARGIN EXISTS TO THE LICENSING LIMIT FOR SMALL BREAK LOCA TO ACCOMMODATE THE EFFECTS OF A DROPPED THERMAL SHIELD
OVERALL CONCLUSION FOR EFFECT OF DROPPED OR TILTED THERMAL SHIELD ON ACCIDENT ANALYSES PRELIMINARY ASSESSMENT CONCLUDES THAT SUFFICIENT MARGIN IS AVAILABLE IN ALL SAFETY ANALYSES TO ACCOMMODATE EFFECTS OF DROPPED OR TILTED THERMAL SHIELD FOR CYCLE 10 AND CURRENT ACTUAL LEVEL OF STEAM GENERATOR TUBE PLUGGING ASSESSMENT IS PRELIMINARY PENDING ENGINEERING REVIEW AND VERIFICATION.
MONITORING PROGRAM OVERALL CONCERN NO. 5 The neutron monitoring system proposed is too vague.
It cannot be concluded from the available information that further degradation would be detected.
RESPONSE
o The purpose of the monitoring program is to detect changes from the present thermal shield condition which has been inferred from inspection results. Thus a long term baseline is not required.
o Monitored data obtained during the operation will be compared to a formal acceptance criteria established to identify further degradation of the thermal shield.
o Neutron noise data and analysis clearly indicate that degradation leading to the postulated worst credible case will be detectable.
MONITORING PROGRAM SPECIFIC CONCERNS NO. 9:
BASELINE To establish a baseline data base for the thermal shield:
a)
The thermal shield must have the proper preload on the supports to yield interpretable data, otherwise, b)
It needs data collection from extended period of time.
Neither of these conditions is satisfied in the proposal.
RESPONSE
o Reported thermal shield monitoring results have shown that interpretable data has been acquired subsequent to "loss of the effectiveness" of supports.
o A long.term baseline is not considered necessary for the SONGS 1 case since the program is structured to monitor for changes from the present support conditions.
o Initial conditions have been inferred from inspection and analysis results. Initial operating data will reflect these present plant conditions, and act as baseline for comparison.
o The acceptance criteria will be established from data acquired during the first 90 days of operation.
o Subsequent data will be evaluated to this acceptance criteria.
If exceeded:
The NRC will be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> The data will be evaluated by an expert committee (Westinghouse/CE/SCE)
Report and documented disposition will be issued to NRC within 30 days
MONITORING PROGRAM SPECIFIC CONCERN NO. 9:
DETECTABILITY LIMITS Such limits to be established require:
a)
An adequate data base, and b)
Parametric studies of the anticipated vibration mode Neither of these conditions is satisfied in the proposal.
RESPONSE
o The data collected during the first 90EFPD will be the baseline.
o Acceptance criteria will be established based on evaluation of these baseline data and parametric analytical studies.
o Parametric studies indicate that detectable differences in thermal shield vibration will accompany further degradation the appearance of a thermal shield beam mode decreases in the natural frequency and increases in the amplitude of thermal shield shell modes increase in the core barrel beam mode natural frequency 0
Neutron noise signal levels estimated for further degradation cases are significantly greater than neutron noise levels for the present condition (data acquired on SONGS 1 prior to the present outage).
o Accelerometers mounted on reactor vessel flange will also be monitored. Past experience has shown that upper head accelerometer peak G levels increased with operating time for a plant with a degrading thermal shield.
SUMMARY
OF SONGS 1 THERMAL SHIELD ASSESSMENT Natural Frequency (Hz)
CASE Beam Modes Shell Modes CB CB TS TS TS N=2 TS N=2 104 6.6 6.7 13.8 4.4 6.8 103 6.6 6.7 13.8 4.3 4.9 102 6.6 6.7 13.2 4.1 4.7 003 7.6 8.7 5.2 3.7 3.9 4.7 000 7.3 7.3 2.4 2.5 3.8 4.6 RMS Vibration Amplitudes of Thermal Shield Modes at The Approximate Locations of the Ex-Core Detectors (mils)
Thermal Shield Beam Modes Thermal Shield Shell Modes Upper Detectors Lower Detectors Upper Detectors Lower Detectors 104
< 0.2
< 0.2 24 13 103
< 0.2
< 0.2 32 16 003 14 11 42 21 000 50 24 49 24 number of undamaged support blocks number of unworn keys number of intact flexures CB:
Ccre Barrel TS:
Thermal Shield i0
Detector 1206, lower section, 1354 10-Case 000 10-7 Case 003 malized V
PSD g0 8
1.5Hz 10
.9
- 5. 5Hz 3.5Hz 10-10..
10-11 0
5 10 15 20 Frequency (HZ)
SONGS 1 Neutron Noise Power Spectral Density (11-2-88)
Figure 1
PECIFIC CONCERN NO. 5 THE LICENSEE PROPOSES TO OPERATE UNTIL THE POTENTIAL DAMAGE BECOMES AN ECONOMIC BURDEN.
DESCRIBE THE PLAUSIBLE ADDITIONAL DAMAGE TO THE THERMAL SHIELD, CORE BARREL OR REACTOR VESSEL THAT THE LICENSEE IS PREPARED TO ACCEPT AND REPAIR AFTER SHUTDOWN. ESTIMATE THE ADDITIONAL RADIATION EXPOSURE TO REPAIR PERSONNEL AS A RESULT OF THIS PLAUSIBLE ADDITIONAL DAMAGE.
RESPONSE
o BASED ON THE WORST EXPECTED CONDITION AND THE WORST CREDIBLE CASE ANALYSIS, IT WILL BE NO ECONOMIC RISK TO OPERATE ANOTHER CYCLE.
o HOWEVER, AN OPTIMUM REPAIR PLAN AT NEXT REFUELING OUTAGE DOES AVOID SIGNIFICANT ECONOMIC BURDEN AND PERSONNEL SAFETY HAZARD ALARA-SIGNIFICANTLY LESS RADIATION EXPOSURE (86 MORE MAN-REM)
LOWER PROBABILITY FOR ANOTHER REPAIR SIGNIFICANTLY LESS OUTAGE TIME (5 MONTHS)
SIGNIFICANTLY LOWER REPAIR COST ($5 MILLION)
- SPECIFIC CONCERN NO. 10 DESCRIBE THE DESIGN OF THE REPAIR THAT WOULD BE PERFORMED ON THE THERMAL SHIELD COMPONENTS CONSIDERING THE.KNOWN DAMAGE.
RESPONSE
10 CONCEPTUALLY, A REPAIR SIMILAR TO THE HADDAM NECK REPAIR IS ENVISIONED HOWEVER; IT CANNOT BE DIRECTLY APPLIED TO SONGS BECAUSE:
THINNER THERMAL SHIELD MAY IMPACT DESIGN AND MOUNTING OF SUPPLEMENTAL LIMITER KEYS.
FLEXURES STILL IN PLACE AT SONGS 1 BUT HAD BEEN PREVIOUSLY REMOVED FROM HADDAM NECK.
DIFFERENCE IN SEISMIC LEVELS BETWEEN SONGS AND HADDAM NECK MAY IMPACT LIMITER DESIGN.
DIFFERENCE IN INTERNALS CONFIGURATION BETWEEN 3 &
4 LOOP PLANT MUST BE INVESTIGATED FOR IMPACT ON LIMITER LOCATIONS AND DESIGN.
0 DEFERRAL OF A REPAIR ALLOWS TIME TO ADEQUATELY ADDRESS THESE ISSUES AND PLAN THE REPAIR TO MINIMIZE RADIATION EXPOSURE AND ECONOMIC IMPACT.
SONGS UNIT 1 DESCRIPTION OF NIS REPLACEMENT o
Replace entire NIS o
Replacement is functionally identical to original NIS except for intermediate range channels which include wide range post-accident monitoring capability.
o Consists of four channels:
Two channels include source, intermediate and power ranges The other two channels consist of power range only
NIS CHANNEL INTERFERENCE/CROSSTALK o
Coincidentor Test Circuit caused high count level on source ranges o
Relay Actuation in Coincidentor caused spurious actuations in NIS o
High counts indicated on Intermediate Ranges while in shutdown 0
Operation of Certain Plant Equipment causes spurious actuation of Intermediate Range High Startup Rate Reactor Trip in shutdown o
Operation of NIS Mode Selector Switch at Operators console caused spurious actuations on the Intermediate Ranges
MODIFICATIONS TO RESOLVE INTERFERENCE/CROSSTALK o
Isolation/Regulating Transformers provide filtering, regulation, ground loop path interruption o
Changed power source from Regulated Buses to Vital Buses concurrently with addition of Isolation/Regulating Transformers o
Suppressors to eliminate transients caused by coincidentor relay switching a
System grounding enhancements o
Corrected wiring error at Channel IV isolation amplifier o
Additional modification being considered to eliminate spurious actuation of Intermediate Range High Startup Rate Reactor Trip at lower end of range due to transients caused by operation of plant equipment
POWER RANGE PERFORMANCE EFFECTS o
Power Range measures d.c. current, not pulses as do the Source and Intermediate Ranges. Therefore, the Power Range is inherently much less susceptible to interference.
o Power Ranges in two-Channels are completely independent of Source and Intermediate Ranges.
o Power Ranges in the other two Channels are associated with the Intermediate Range channels but are not adversely affected by interference on the Intermediate Ranges. No Power Range Interference has been observed even when Source or Intermediate Ranges have indicated high levels of interference.
CONCLUSION:
There is no reason to expect problems with Nuclear Noise Measurements.