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| number = ML20056F880 | | number = ML20056F880 | ||
| issue date = 08/31/1993 | | issue date = 08/31/1993 | ||
| title = Nonproprietary VC Summer Nuclear Sttion SG Pulled Tube Exam Results Presentation Matls from Nrc/Sce&G/Westinghouse Meeting on 930727 | | title = Nonproprietary VC Summer Nuclear Sttion SG Pulled Tube Exam Results Presentation Matls from Nrc/Sce&G/Westinghouse Meeting on 930727 | ||
| author name = Mcinerney J | | author name = Mcinerney J | ||
| author affiliation = WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. | | author affiliation = WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. | ||
Latest revision as of 04:02, 13 November 2023
| ML20056F880 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 08/31/1993 |
| From: | Mcinerney J WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML19310D651 | List: |
| References | |
| SG-93-08-009, SG-93-8-9, WCAP-13824, NUDOCS 9308310260 | |
| Download: ML20056F880 (100) | |
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WESTINGHOUSE CLASS 3 (NON-PROPRIETARY)
WCAP-13824 SG-93-08-009 V. C. Summer Nuclear Station Steam Generator Pulled Tube Examinatio,. Results Presentation Materials from NRC/SCE&G/ Westinghouse Meeting on July 27,1993 AUGUST 1993 APPROVED:
J. MCINERNEY NUCLEAR SAFETY LICENSING O
WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR SERVICES DIVISION P.O. BOX 355 PITTSBURGH, PENNSYLVANI A 15230 C 1993 WESTINGHOUSE ELECTRIC CORPORATION ;
ALL RIGHTS RESERVED
a I
i
'l South Carolina Electric & Gas Company l
. V. C. Summer Nuclear Station l
/ \ l t
l i
NRC / SCE&G / Westinghouse :
Meeting '
\ / l Steam Generator Pulled Tubes i
i Examination Results l l
l N / July 27,1993 j T u s s.a.e v s e s I i
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Westinghouse Energy Systems Electric Corporation P*255"ewama
- mas: ;
July 26,1993 ET-NRC-93-3931 Document Control Dest ET-NSL-OPL-II-93-344 US Nuclear Regulatory Commission Washington, DC 20555 Attention: Mr. James T. Wiggins, Acting Director .
Division of Engineering Technology
Subject:
" Steam Generator Tube Integrity Assessment for Virgil C. Summer Nuclear Station" f Presentation Materials
Dear Mr. Wiggins:
Attached is a single copy of presentation materials entitled " Steam Generator Tube Integrity Assessment ,
for Virgil C. Summer Nuclear Station" (Proprietary), which will be used at the July 27,1993 NRC meeting.
This presentation material contains information which is proprietary to Westinghouse Electric Corporation. Accordingly, we request that this information be withheld from public disclosure.
We will comply with the requirements of 10 CFR 2.790 to provide proprietary and non-proprietary versions of the above material together with an affidavit as soon as the proprietary and non-proprietary versions have been prepared. We will submit the total required number of copies of the proprietary and non-proprietary versions of the information and the required affidavit at that time.
in the meactime, we hava provided sufficient copies for your information and use. M. P. Siemien, Esq.
of the NRC Office of the General Counsel, has advised Westinghouse that she concurs with this procedure.
t We expect to be able to fully comply with the requirements for the proprietary and non-proprietary versions of the informat.on and an accompanying affidavit within four weeks.
Very truly yours, ;
Nicholas J. Liparulo, Manager h
Nuclear Safety and Regulatory Activities MRZ/lp Enclosure
V C Summer Nuclear Station
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r Introduction & Purpose Jack Skolds i Recap of April 27 Meeting Ron Clary Update of NDE Results Rollin Kelso Overview of Evaluation & John Frick Remedial Actions .
Detailed Presentations Pulled Tube Examination Summary Tom Pitterle Leak & Burst Tests llorphology i i
Causative Mechanisms John Barkich Mechanisms Comparisons with Another Plant
. Corrective Action Tube Integrity Analysis Tom Pitterle 1
Burst Margins j SLB Leakage Assessment Remedial Actions Summary Jack Skolds Discussion All
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r V C Summer Nuclear Station i l
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. Inform NRC of: !
9 Evaluations performed on Pulled Tubes I
O Results of Leak & Burst Tests l i
9 Identification of Causative Factors ,
9 Tube Integrity Analysis 9 Remedial Measures
l V C Summer Nuclear Station ;
N.... _fMpgil a p@ggy,3p7eMietingNilip4NRO !
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9 VCS Steam Generators Inspection Techniques & Methodology ,
RF-6 Results !
i RF-7 Inspection Plan & Results :
9 Review of Cycle 7 Operation i
9 Tube Pulls t Reason / Tube Selection Basis Tube Pull Process ;
Examinations / Schedule l t
Preliminary Findings l i
i 9 Cycle 8 Operation !
Basis for Start-up/ Initial Operation l Future Activities j t
9 Steam Generators to be Replaced in Fall 1994 i 1
1 1
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J i
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- V C Summer Nuclear Station 1 :
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t 1 i 9 There were Inspection Improvements Between l l RF-6 and RF-7 !
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-t v I i 1) Data Analyzed per TSP-IPC WCAP #13522 l t
- i
- 2) Probe Wear and IPC Transfer Standards used a
in RF-7 f i
- 3) Industry Events Between RF-6 and RF-7 Raised j
- Analyst Sensitivity for Recognition of TSP Flaws l 1
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- G No Flawlike Indications Left in Service j Regardless of Indicated Deptil !
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V C Summer Nuclear Station l
f e We were Searching for Causative Factor (s)
- 1) No Correlation Found Between Plant Operating Parameters and the 5 Atypical Voltage Indications
- 2) Pulled Tube Testing / Analysis Required to Determine Cause e Tubes Damaged During Tube Pull Process
- Expected Results were Conservative Compared to Insitu Tube Behavior e Population of Atypical Voltage Indications was Small - 5 Indications out of 99,592 Hot Leg ,
Intersections
- Interim Operating Period of 111 EFPD Based on Conservative Safety Evaluation e We informed NRC we Would Return to Discuss Results of the Pulled Tube Testing 7 U B S . ? - 81 ( 9 3
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V C Summer Nuclear Station Refuel 7 Results ;
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L 5 Indications Exceeded 3.7 Volts -
Out of 12,449 Tubes or 99,592 Hot Leg TSP /FDB Intersections i
S/G Row Col Volts Location f
B 42 43 22.32 01H. + 0.00 i B 28 41 11.59 01H + 0.00 ,
i B 33 20 9.84 01H + 0.00 '
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B 31 45 7.72 01H + 0.00 B 30 45 6.02 01H + 0.00 i
3 j 01H Is the Flow Distribution Baf fle Plate r
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ECT Probe Uncertainty l .
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- All 5 Atypical Large Voltage Indications Found During RF-7 were Inspected with the Same Probe During RF-6 s
1
- Review of RF-6 Data Indicates that Voltage Response of this Probe was Deficient with Respect to the i
Requirements in WCAP-13522 I
- Actual Voltage Growth less Than Indicated During i l
Cycle 7 l
- Implementation of Probe Wear Standard Prevented 1 i
Reoccurrence in RF-7 I
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w-w4DNL .AT?L lWi* X*M+gion ; * *@=4: 2:!k?M R28 C41 R33 C20 R4 2* C43 Bobbin: 11.6 V l Bobbin: 9.8 V Bobbin: 2 2.3 V 550/130 550/130 550/130 Field 93% 88% 91%
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l ECT RPC: M AI - 7.7 V R PC: M AI - 6.4 V RPC: M AI - 10.8 V 1
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Bobbin: 4 3.6 V 3obbin: 3 2.8 V ' Bobbin: 42.4 V 94% 95% 89%
Lab ECT R PC: M AI - 14.7 Y R PC: M AI - 13.4 V RPC: M AI - 15.0 V 0.75" Long 0.54" Long 0.75" i
M AI - over 40 M AI - over 90 M AI - over 60 Degrees, within Degrees, within Degrees, within Lab crevice crevice crevice 100% 100% 100 %
UT max length 0.76- max length 0.46" max length 0.68" l
V = Vol ts MAI= Multiple Axial Indications
- Lab ECT did not include use of TSP-IPC Transfer Standard nn.,<..v,
V C Summer Nuclear Station
% ,P%;n "%
es Exa~m.inaTion Res
, _ _ _ _ ..-nu w w . wwww:umm:+n s. Mis Crack Morphology -
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- Macrocracks 1 & 2 are interconnected near the centers of [
the macrocracks by a "V shaped" 100% TW crack that is ;
approximately 0.1" long on each s.ide of the "V".
, l 1
Overview of Evaluation and Remedial Actions
V C Summer Nuclear Station
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6 . ,,-j ' s w &m; :. g;;w., : ,. . ~ . L ,;' .;. j c : A j a y f Causative Mechanisms for Cycle 7 Crack Growth Rates 9 Tube /FDB Misalignment
- Contact Region Maintained During Shutdown and Operating Conditions
- Misalignment Configuration Permits more Rapid Concentration of Contaminants than Standard FDB or TSP Intersections l
- Highest Crevice Superheat Conditions at FDB l
Crevice which Enhances Contaminant Concentration
- High Crevice Temperature n o....,,,,
V C Summer Nuclear Station
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. .: .. agg Causative Mechanisms for Cycle 7 Crack Growth Rates 9 Dissolved Copper Transported to Crevice Locations
- Reducible Cu Compounds ' Acting as Oxidants to Accelerate SCC in Alkaline Environment S Caustic S/G Crevice Chemistry Over Most of Cycle 7
- High Cation / Anion Ratios Present to Accelerate ODSCC in Cu Oxidizing Environment
- Alkaline Conditions Associated with Ammonia Breakthrough of S/G-Blowdown Demineralizers 1
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V C Summer Nuclear Station M iW B a# W m :si M e #1 Remedial Actions Chemistry Actions to Reduce Crack Growth 4 Reduced Copper Transport to S/G MSR Tube Bundles Replaced in RF-7 (Initial Reduction of Copper Concentration in FW by a Factor of 3 and Still Declining) 9 Reduced Potential for Caustic Crevice Environment i
- S/G Blowdown /Demineralizer Operation Modified to Reduce Potential for Na Entry to S/G i
- Monitoring Na/Cl Molar Ratio on a Daily Basis
- Goal to Control Molar Ratio to 50.7 in Bulk Water to Reduce Crevice Caustic Corrosion Potential
- Existing Procedures (Response to Ratio Above Goal) to be Modified for Ratio Control
- Method (Chemical Addition or Alternatives) to Control Molar Ratio to be Implemented TUBB8997#93 l
j
V C Summer Nuclear Station f (h IkhlbO ,
Overall Tube Integrity Conclusions -
9 Even Though Large Crack Growths Occurred on a Few l Tubes in Cycle 7, Insitu Tube Integrity was Maintained at End of Cycle 7
- Burst Ca7 ability Satisfying R.G.1.121 Guidelines with FD3 Restraint :
- Potential SLB Leakage less than that Resulting in a Small Fraction of 10 CFR 100 Dose Limits using SRP Methodology
?
i G Remedial Actions Implemented at V.C. Summer for Cycle 8 Can Be Expected to Result In:
- Reduced Crack Growth Comparable to that from Prior Operation and Typical Domestic Plant Experience
- Enhanced Margins Against Radiological Limits even if it is Postulated that a SLB Event Occurs Subsequent to Reoccurrences of Large Growth Rates 9 Full Cycle 8 Operation to Scheduled Refueling l is Acceptable :
n....... 4,,,
V C Summer Nuclear Station pwamm. -
=n -' == ~nypp;qp kkIhhs!! b. b bh h!fhENN Remedial Actions Plant Controls to Enhance Safety Margins 9 Operating Leak Limit Reduced to 150 GPD
- Increase Leak Before Break Capability 9 Increased Operator Training / Sensitivity to S/G Transients
- Further Enhance Likelihood of Appropriate /Tirnely Response to S/G Transients such as Secondary Pipe Breaks 9 Reduction in Coolant Activity Limit by Factor of Two
- Reduce Potential Radiological Consequences of a Secondary Pipe Break I 9 Maintain 1 PORV Available (Unblocked) at all Times During Cycle 8 !
- Increase Confidence that S/G Tube a P in a Transient will be 2335 psid with Associated l Reduction in Radiological Consequences l O Enhanced Plant Response Criteria Based on Radiation Monitors Sensing S/G Tube Leakage 9 All Flaw Indications Found in 1993 were Plugged
- Reduced Likelihood of significant indications lef t in Service Subject to Potential for Large Growth than 1v....,..,,,, Following 1991 Inspection i
l
a V.C. SUMMER NRC MEETING JULY 27,1993 .
i I
I PULLED TUBE EXAMINATION
SUMMARY
i
\
T. A. PITTERLE WESTINGHOUSE NUCLEAR SERVICE DIVISION
I i.
VOL TAGE GROWTH RA TES HIGH GROWTH FOR 5 INDICATIONS e 5 to 21 Volt increases based on nominal 1991 voltages
-large uncertainties on 1991 volts due to probe wear ,
e All at FDB in S/G B e All located in one quadrant of FDB e All had modest indications (< 1.1 volt) by EC reevaluation of prior inspection GROWTH RATES FOR REMAINING INDICATIONS e 533 Total Indications on 487 tubes e Average growth of 29% < 44% found in prior cycle e Maximum growth of 2.2 volts
- same as prior cycle growth rate
- Typical of domestic experience
i i
1991 INSPECTION CONSIDERA TIONS l I
PROBE WEAR STANDARD AND CROSS-CALIBRATION OF ASME STANDARDS NOT APPLIED IN 1991 e Same probe used in 1991 for all 5 atypical 1993 Indications e if probe wear standard used in 1991, probe would have been rejected on 1st data tape ,
^
e Throughwall hole voltage variations of - 60% to 80% over life of probe CONCLUSIONS e if 1993 EC analysis guidelines applied in 1991, very likely atypical 1993 indications would have been plugged in 1991 e Uncertainty of ~ 100% on 1991 volts for 5 atypical indications
t i
l V. C. SUMMER 1991-93 GROWTH STUDY Largest End of Cycle Indications (>2.0 Volts) i 1993 1991 Voltage S/G Tube TSP Volts' Depth (%) Volts Depth (%) Growth !
B R42C43 2
FDB 21.9 86 0.73 69 21.17 I i
B R28C41' FDB 11.9 93 1.09 97 10 81 2
B R33C20 FDB 9.44 89 0.56 93 8.88 B R31C45 FDB 7.63 87 0.50' -
7.13 8
B R30C45 FDB 5.92 85 0 31 -
5.61 j 81 1.06 76 2.07 !
A R36C64 FDB 3.13
^
B R38C63 2 2.81- 72 0.73 78 2.08 A R17C104 2 2.68 83 0.51 63 2.17 [
i B R41C68 8 2.58 66 1.65 73 0.93 f B R42C44 11 2.54 60 1.47 79 1.07 B C42C47 8 2.31 85 2.33 79 -0.02 ,
t I
C RBC21 2 2.22 83 0.95 46 1.27 B R42C73 5 2.11 90 1.15 99 0.96 B R32C52 2 2.10 68 0.82 76 1.28 Notes-1 Volts do not include cross-calibraton factor for ASME standards in growm study.
- 2. Pulled tube.
I
- 3. Supplemental evaluaton performed to identfy flaw mthin abnormally high noise tevel i
I
i VOLTAGE GROWTH PER CYCLE FOR V. C. SUMMER S/Gs Number of -
Voltate Range SIG Indications Avg.V,oe Avg.AV Avg. %AV i
1990 - 1991 (427 EFPDs)
Entire Range A 17 0.60 0.29 48.3 B 43 0.71 0.30 42.3
~
C 27 0.63 0.26 41.3 All 87 0.66 0.29 43.9 V,0c < 0.75 volt All 55 0.47 0.36 76.6 V,cc 2 0.75 volt All 32 1.00 0.16 16.0 1991 - 1993 (446 EFPCs)
Entire Range A 109 0.59 0.11 18.6 B* 207 0.65 0.21 32.3 C 202 0.54 0.15 27.8 All* 518 0.59 0.17 28.8 B 212 0.65 0.46 70.8 V,0c < 0.75 volt All* 390 0.46 0.18 39.1 V,3e 2 0.75 volt All* 128 1.01 0.12 11.9
- &ctudes 5 largest indicatons at FDB elevaton of S/G B
l l
V.C. SUMMER GROWTH DATA ~l All S/G Indications w/o 5 largest at FDB in S/G B in 93-91 25 _
_. - ; j~: 100 .
AIX'"*
I
- 80 20 I d
$ _N
.9 .
3 i O
E0 15 + = - - 60 .O E .
2 o.
_c D
~
j .
10 - 40 - i 8
cs 1/l/' 1 E
a
- c.
- 1 : i o
f'/ ;
5- -
h 5- s - 20 1
- ,/
/'J i
=
y' -
k < '
0 , i i
i i i ; i i i I i i sEni r
- a i i ,
s1 ,0 .
-0.5 0 0.2 0.4 0.6 0.8 1 1.4 2.2
-0.1 0.1 0.3 0.5 0.7 0.9 1.2 1.6 Voltage Grov4h 93-91 : CDF 93-91 91-90 0 CDF 91-90 ,
i PULLED TUBE EDDY CURRENT DA TA FIELD AND LABORATORY ANALYSIS OF FIELD DATA IN EXCELLENT AGREEMENT e Pre-pull 9.8 to 22.9 volts POST-PULL VOLTAGES SUBSTANTIALLY INCREASED ABOVE
- PRE-PULL VALUES e Post-pull 32.8 to 43.6 volts
- Increase likely due to tearing of crack ligaments ;
e Likely result of 3000 lb. pull force
- Applied to test for binding at FDB
- Binding above FDB so fullload applied through degradation at FDB e R28C41
- Increased from 11.8 to 43.6 volts
- Potentially tearing of 5% wall thickness (0.24"Ig.)
ligament between TW cracks within macrocrack.
POST-PULL RPC AND UT LENGTHS & DEPTHS IN GOOD AGREEMENT WITH DESTRUCTIVE EXAM.
a
ORIENTA TION OF INDICA TIONS IN FDB FDB CREVICES PARTIALLY PACKED WITH DEPOSITS
- ~30 to 180 INDICATIONS LOCATED WITHIN DEPOSITS ,
- Area of contact with FDB EDDY CURRENT DATA INDICATES TUBE OFFSET IN FDB AND LIKELY IN CONTACT AT AREA OF INDICATIONS
- RPC amplitude from low frequency response to FDB is highest near indications FDB CONTACT POINT NOT AT MOST PROBABLE RADIALLY OUTWARD POSITION AT HOT CONDITIONS
- Result of tubesheet bow
- Common hot and cold contact points indicate tube misalignment offsetting typical hot to cold movement of tube.
l CONCLUSIONS l
- Tube in contact with FDB within deposits of partially packed
( crevice
- Tube locations likely associated with lower probability tolerances i on FDB and/or tubesheet hole location and/or tube hole 4 straightness.
V. C. Summer Pulled Tubes Orientation of Deposits and Indications R42C43 R28C41 R33C20 180
- 180' 180'
{l$ ///
Contact-Type / 90' .
f~270' 90*. /
90* 270a .
270*
/
Orientation
~
330*
0* O' Deposa - O' Orientation 180' 180* 180*
j 'f f ,
90a
/ 2 270' wa w 270* 90* 270' Exa ination ,, ,
Indications 45' - d5' 2ma -
/ -
Zone wth 40' l * '*"*
0 0 go \ Zone with indicatons FDB "I* "*
P Cen erine o
,, Tubelane Tubelane
_. _ ~ _ _ . - __ _ -___.
~
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MONTE CARLO ANAL YSIS FOR FDB CONTACT CONDITIONS ; OBJECTIVES
- Probabilistically assess contact conditions at FDB METHODS
- Monte carlo analyses of tube alignment conditions
- Sampling of all dimensions (TSPs, TS) affecting tube alignment
- For each sample of alignment conditions, tube analyzed by i nonlinear analyses for FDB contact conditions RESULTS
- No occurrences of significant hoop stress at FDB
- FDB Contact Loads: 93% < 100 lbs.
7 % - 100 to 250 lbs.
- About 6% of tubes maintain significant (> 15 lbs.) contact forces at both hot and cold conditions
-t o
3 LEAK TESTS INITIAL LEAK TESTFACILITY i R28C41 l l I e Tested to ~2650 psi 6P i e Data shown to exceed facility capacity
- i l
- Inability to stabilize secondary pressure !
- Measured 95 liter /hr at 2650 psi underestimates actual leak rate i o Plastic crack opening and hysterisis effect invalidate repeat tests l l
e Crack opening of ~ 10 mils following leak tests j e 5% wall thickness ligament likely torn before or during leak test ! l R33C20 ; e Tested only to 2250 psid and 36 liter /hr found to exceed facility ' capacity e Retests acceptable at > 2250 psid. l l i i i l 9 M , ,v, , - -~
1 LEAK TESTS ' INCREASED CAPACITY FACILITY i NEW LEAK TEST EQUIPMENT CONSTRUCTED WITH ~ 500 LITERS /HR CAPACITY , R33C20 ;
- Tests at four aPs up to 2505 psid '
- Leak rates up to 137 liter /hr R42C43
- Tests at four APs up to 2040 psid
- Leak rates up to 310 liter /hr.
)
J k p
i.l LEAK RA TES A T REFERENCE sPs LEAK RATES AT 2335 AND 2560 PSID DESIRED i APC LEAK RATE ADJUSTMENT PROCEDURE APPLIED t t e Minor changes to accommodate draft NUREG-1477 comments i i LOG LEAK RATE vs. AP USED TO INTERPOLATE / EXTRAPOLATE TO j REFERENCE APs - i i i i i i i
s, : 1000 t 100 -- 5 5 i! . 2m
% 10 --
1
.3 1 : : :
2200 2300 2400 2500 2600 j Pressure Differential, psi Leak rate vs. pressure differential for R33C20 pulled tube of V. C. Summer steam generator (under adjusted reference absolute pressures and temperature) l I
i-L i
~
i L i 1000 100 -- 5 i .M 3 . $ 10 -- I t . 3 l 1 ; ; l 1910 1960 2010 2060 l Pressure Differendal psi i Leak rate vs. pmssum differendal for R42C43 pulled tube l of V. C. Summer steam generator (under adjusted reference absolute i l pmssures and temperature) l
'l MEASURED AND ADJUSTED LEAK RATES FOR R33C20 Measurements Test Temperature Primary Secondary Differential Leak Rate l Condition *F pp, psia ps, psia Dp, psi liter /hr !
NOP 580 2000 490 1510 0.53 l SLB 580 2620 370 2250 56.5 SLB 585 2790 370 2420 118 SLB 585 2860 355 2505 ,137 Adjusted for Reference Absolute Pressure and Temperature NOP 616 2250 740 1510 0.55 l SLB 616 2265 15 2250 36 SLB 616 2435 15 2420 82 SLB 616 - 2520 15 2505 99 Reference Temperature and Pressure Diffeiantials NOP 616 2250 900 1350 0.19 . SLB 616 2350 15 2335 52 SLB 616 2575 15 2560 137
i MEASURED AND ADJUSTED LEAK RATES FOR R42C43 e t
- Measurements Test Temperature Primary Secondary Differential Leak Rate ;
Condition 'F pp, psia ps, psia Dp, psi liter /hr l } , t NOP 550 2025 ~95 1330 21.4 ; SLB 580 2515 580 1935 176 ! SLB 575 2770 770 2000 255 SLB 585 2765 725 2040 310 Adjusted for Reference Absolute Pressure and Temperature : NOP 616 2250 920 1330 22.1 1 SLB 616 1950 15 1935 74 SLB 616 2015 15 2000 109
- SLB 616 - 2055 15 2040 147 Reference Temperature and Pressure Differentials NOP 616 2250 900 1350 23.9 !
SLB 616 2350 15 2335 707 SLB 616 2575 15 2560 2210 i ] r J t a j 6 i f s 4 9 r -
BURST TESTS BURST PRESSURES l e R33C20 - 5082 psi
- Burst length of 0.885" > 0.47" macrocrack length (significant crack tip tearing) ,
e R42C43 - 3618 psi , ,
- Burst length of 0.658" > 0.50" TW length & < 0.75" i macrocrack length (minimal tearing of crack tip) e R28C41 - 2724 psi - No tearing at crack tip: burst length - TW length of 0.69" and less than 0.80" macrocrack length.
J. No Burst occurred ir .sst. i COMPARISON OF MEASURED AND EXPECTED BURST PRESSURES , i e Expected values based on Burst Pressure vs TW length correlation e R33C20 - 7.2% above expected value e R42C43 ~ 7.3% below expected value
- Acceptable data considering complexity of H-shaped burst crack l t
e R28C41 ~ 16.1 % less than expected value
- Test result unacceptably low due to prior damage from tube :
pulling operations and leak testing (10 mil crack opsning)
- Considered invalid data point l
l l
Crack Profile, R28C41 1.00 , f 0.90 '
\ - -
0.00 Ligaments not present during burst testing. 0.70 8 h 0.60 it k 0.50 -- -- 8 l . tim 0.40 0.30 - - 0.20 - - - - 0.10 - - 0.00 - - 0.00 0.10 0.20 0.30 0.40 0.50
- 0.60 0.70 0.00 Crack Length (in.)
[CGE_URST.XLWlR28C41 Profde P.1 RFK: 7/6f)3,6 4 i PM i
it t I i i 2 i
- c. ;
N ! w . o =
- D e-i \
d a b l' l :2 a r t o N ! 6 . i .I i t o ' C 6. - O - N 4 O ; .. co i CD o i i C * ! o O _ e
- w :: C i
8 o c-
. r i 1" o 5 i x v c o
e 63, N 1 w o a- -
- O es 4.,
0 !, 4 o , t M ,
! 6 i 1-r J
o i cv ; o ) i J I i e o d & i e 1 , o N l t, i,
., . , , 8 g
m g i ~ E$$k$E$'$k!$ s 6 6 6 6 6 6 6 6 6 6 N g i ! uogoe>g tpdeO WBO *i w l O ! O t .t a l i
d . Crack Profiles for R42C43, removed from V. C. Summer I 1.00 r- f 0.90 l
" f
\
\
0.80
/ E ,
.\
c 0 70 O j
/ Profile of Crack "A"~
\
B
= 0.60 f, I .
\-
t I ( L 9 0.so f M f
/
$ o.40 - '
] /
e 6 i o.30 f i
.20 Cracks "A" & *B' are similar They Profile of crack "B" g were connected by a 100% through
( 0 10 i - wait erack. \ 7 0.00 J l i 1 I l 1 1 t_i_ . 0.00 0.10 0.20 0.30 0.40 0.50 0.60 0.70 0.80 Crack Length (in.) f P.3 RFK: 7/6f)3,6.42 PM (CGE_BRST.XLW]R42C43 Proide e.,-_, .--m,,. .,,-...,,.re..,%s.e ,e,e., e..w.,v.cw,,,-,.~.- ..,.r,,e ,-..-- w ,,,-,m --w,..wem.-,..,...-,,,,-.-w..mnv.,-, ,.-we..-.-e. ,,...-e. e a%- e. --ni.m.,.e..-...--.-...enw-a-* -eewe-. - + . - - . . . . - - + . . - - - + . . . - - - . - - ..-
l
.U ,
8, ' 6 < I l I 4 Y I h N E ' u ! b t U
> f L
3 va
- 7) l G f 6 i 4 ,
3 ! ClC I I i i i [ I I 12- : E I
.+
i
COMPARISON OF MEASURED AND EXPECTED BURST PRESSURES Tube Total TW Length Effectiv Measured Expected % Length
- e Burst Burst Difference Length Pressure Pressure R33C20 0.470 0.33 0.431 5082 4739 + 7.2%
(96%) R42C43 0.750 0.50 0.643 3618 3906 -7.3 % (95%) R28C41 0.800 0.69 0.725 2724 3249 -16.1 % (95%) Average depth given in parenthesis. 2-.- ._r.. or m + . - , , ... . . , -,-- --.,e--, - + - - - - - . . , ~ ~ . , , , - , , , . . - - . , - , , ---,-,.-~,m. ,..-4. .-r .r. -- - , . .-
2 , J. 6 6 i , I I i I 1
'1 s
t s t
)
i e 1 ( P 4 L i
)
f r a i I i e e t 6 i i 1 i
' kI v
n l I e J b O l P n
CRACK MORPHOL.OGY R33C20 e Single Dominant TW axial crack ; e Small Cellular corrosion patch j R28C41 ' e Single dominant TW axial crack within - 45 cellular band e Short TW microcracks not yet linked .to form macrocrack - 45 from dominant axial crack e Cellular patterns p.esent to 35% to 58% depth R42C43 i e Two dominant TW axial cracks separated by ~ 20 within ~45 cellular band
- Linked by V-shaped TW indication to form a H shaped crack !
network e Cellular patterns present from 12% to 35% deep i h
- 1. 25 i i i 1.0- ,
- S P Top p
e '
.c c
C
. 52 3
% 7 m Yi>Na v .' ,
- 0. 25- -SP liottom I I I 0
. 0 90 180 270 360 Circumferential Position (degrees)
Sketch of the crack distribution found at the flow distribution baffle plate crevice region of Tube R33-C20. Included is the location of the burst test fracture opening. I
,____..__.______________________._____.__m_ _ _ _ _ _ . _ _ _
_ _ _ . , _ ~_ _.
- l. 25 -
i i i
- 1. 0- SP Top
.\ 0T a L -
$ b
.c d\ g 8 /
\\\ ,
= -
g i c ' p' ) ~
. s2 O
C T', l/ F]:}a IJ 1 0.25- .. -SP Bottom O I I I - 0 90 '180 270 360 Circumferential Position (degrees) - r Sketch of the crack distribution found at the flow distribution baffle plate crevice region of Tube R28-C41. Included is the location of the burst test fracture opening. .--.._.--i..m..-~ ,+ = #, w. ..s,,i-- .m e -.~.-.r ....-3 ,,,.i+e*,e ,<+% .-# ..e-w+>+-. - , = ---s -i, e + o. -.~---~w -+e-wm . . mm =. - . - - . . . . - . - - - - -
1.25 - i i I
- 1. 0- - SP Top
. 3
. \
g{rD 8 " E A U 1 )
.E -
.C I l
(
.e I t
! g ( S M U
M)b' -SP 1101toin
- 0. 25 I I I 0
0 90 180 270 360 Circumferential Position (degrees) Sketch of the crack distribution found at the flow distribution baffle plate crevice region of Tube R42-C43. Included-is the location of the burst test fracture opening. _ , . - . . . . _ . . . ~ . . . - _ . _ . _ . . - . _ . . _ . . . . - . . . . _ . . . _ . . . _ . . _ _ . . . , . _ . _ _ _ . . . . ~ . - -. . . _ - .. . . . . . . . . . . _ . . . . . . . - _ - . _ _ - _ _ _ .
p 1 PULLED TUBE CHEMISTRY EVALUA TION COPPER IN THICK OX1DE FILM ON TUBE SURFACE AND CRACK FACE
- Auger Electron Spectrometry (AES)
- Thickest (0.5 pm) oxide film found on an indication at a TSP e Homogeneous distribution of Cu in Oxide film e Cu in crackface oxide film not previously observed on other tubes
- Indicate presence of an aggressive chemical environment at one time. Cu acting as an oxidant to accelerate corrosion NEUTRAL CREVICE ENVIRONMENT AT TIME OF TUBE PULL e Crackface had some Cr and Fe enrichment with Ni depletion suggesting neutral environment
- Presence of mineral Kaolinite which is stable only in neutral environment
- Consistent with hideout return data at 3/93 shutdown and chemistry monitoring since 1/93 shutdown r
l 1
3 ani: ':0-? an': CraC(aCO COm3ariSOnS _ (normalized wt% from AES) g, D E
i V.C. SUMMER PULLED TUBE EXAM j CONCLUSIONS BURST PRESSURE RESULTS e R33C20 (5082 psi) and R42C43 (3618 psi) consistent with ; expected values from burst / length correlation. e R28C41 (2724 psi) resulted in no crack tip tearing and unacceptably low burst pressure.
- Expected value ~3249 psi based on crack morphology ,
LEAK TEST RESULTS , e R33C20: leak rates of 52 and 137 liters /hr at 2335 and 2560 ! psid. j e R42C43: leak rates of 707 and 2210 liters /hr at 2335 and 2560 psid. 4 CRACK MORPHOLOGY e R33C20: dominant axial crack with small cellular patch e R28C41: dominant axial crack within cellular corrosion bend i e R42C43: two dominant axial cracks within cellular corrosion band i 1 1
I l V.C. SUMMER PULLED TUBE EXAM CONCLUSIONS . ALL INDICATIONS WITHIN SLUDGE DEPOSITS OF PARTIALLY PACKED CREVICES l ORIENTATION OF INDICATIONS e Indications at & near FDB contact point (EC and sludge data conclusion) e Very likely common hot and cold contact points
- Indicative of lower probability range of tube alignment conditions: !
nominally expect hot tubesheet bow to change FDB contact point CHEMISTRY EVALUATION ! e Aggressive chemistry environment present at some time i e Thickest oxide film found to date e Homogeneous Cu distribution in oxide film on crackface and tube surface not previously observed in pulled tube exams NEUTRAL CREVICE CHEMISTRY AT TIME OF TUBE PULL e Consistent with 3/93 refueling shutdown chemistry data i i l l
ij P i i V. C. SUMAER SG INVESTIGATION ! CHEMISTRY EVALUATION i 1 i l , 1 i t J. L. Barkich T. A. Pitterle ! Westinghouse Nuclear Services Division J
i
. Causative Mechanism Summary l 1
1
- Synergistic relationship between three causative :
factors led to rapid tube degradation within ; crevice deposits. Dissolved copper species . Localized caustic environment Partially packed hot leg FDB location '
- Copper presence in tube OD oxide film and crack face oxide film indicates its direct role as a '
corrosion accelerant. l i e Localized caustic environment has been l associated with Alloy 600 SCC. l l Steam generator blowdown sodium and chloride ; concentration data indicate a likely alkaline ! operating chemistry environment in the steam i generators during most of Cycle 7, leading to- i potentially strong caustic environments in regions of high superheat. i e Partially packed hot leg FDB locations provide the crevice regions of highest superheat in the steam generators.
1 Crevice Geometry , e As indicated in earlier presentations, the following FDB crevice geometry observations would lead to greater propensity for corrosive attack: . Tube to FDB contact during plant operations and shutdown (Iow probability event). Deposit formation at tube / FDB contact region. i Provides matrix for soluble contaminant i accumulation. ! Partially packed FDB crevices have larger ; available surface area for contaminant ; absorption. Longer time to develop than at tube support plate. i High hot leg FDB available superheat Leads to greater concentration of soluble impurities.
- More alkaline than tube support plates due j to chloride volatility. l
Copper influence on Corrosion
- Reducible copper compounds can act as oxidants which may accelerate SCC in an alkaline environment.
- The presence of copper in the oxide film on the crack face is indicative of dissolved copper transport through the crack network. This copper transport ,
would elevate local potential on the crack face and serve as an oxidant.
- Postulated copper transport paths:
Copper ions accumulate in deposit and directly enter crack network. Copper precipitates in deposit and is subsequently ; oxidized, dissolved, and transported into crack network.
- The major source of secondary system copper has been removed (MSR tube bundle). Reduced feedwater copper concentrations can reduce the tubing corrosion potential.
i i l
~
i Chemistry Influence on Corrosion .; i
- Highly soluble sodium and chloride are primary j contaminants influencing the chemistry l environment in concentrated steam generator ,
crevice solutions at power operating conditions. l t
- While steam generator blowdown sodium to i chloride molar ratios of I would appear to ;
4 indicate neutrality, concentrated steam generator ! crevice solutions typically experience sodium to chloride molar ratios 2 to 2.5 times higher than in { steam generator blowdown due to chloride volatility.
- While low absolute concentrations have typically I been reported for these impurities, evaluation of V. C. Summer steam generator blowdown j chemistry. indicates periods of high sodium to i y chloride molar ratios during the first twelve
- months of Cycle 7. Thus, a caustic environment - '
was highly likely for steam generator crevice ! regions. 3
. 1
- Increases in sodium to chloride molar ratio appear to have occurred at approximately the same time as ammonia-breakthrough of steam generator blowdmyn demineralizers.
s i I i Molar Ratio Co - w w A u o q a e y 6 i i i i i i e e S/D 11-22 y- - =:-
~~
l 00 ! h 5 M==us S/D 12-28 I 1
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v 2 i - l y- -
% Train B Dennin I/S 2 3 ,
3 i ~
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> D> I, C
& l' Demia NH. Break 4-6 $a i 1 --
c. i 6 ~
& Domin O/S 4-20 C N
P % S/D 5-12 4 m
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& Domia I/S 6-1
(-% 2* Demia NH. Break 6-10 y I[ W ]
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; -4= mum 2' Domin NH. Break 9-8 llC % V
- h "
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- d. & S/D l-13
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}' 1 Molar Ratio
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4 M===ms S/D 11-22 g l 9' L 1 h 4===== S/D 12-28 i ? M- 1-- l $ M === Train B Demia I/S 2-3 e i _ N
~_
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> , % l' Domin NH. Break 44 g.
c. j > 3e '
-4=== Domin O/S 4-20 o
% S/D 5-12 f I l
l Q ( =
-4==ms Domin I/S 6-1
% % 2' Domin NH. Break 6-10
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4~
~
- l- $
o i @m w G - W - % 1* Domin NH. Break 8-15 N M
~
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& 5' Y 1 I p -
= n ea v v
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. .,O Train B Demin 1/S 12 5 W 5 -
"+== S/D l-13 I
% e l ~
h - i e r i h S/D 34 l Y
! K 4 .
i
Chemistry Influence on Corrosion
- Cycle 6 chemistry data indicate a potentially more benign corrosion environment than Cycle 7 data.
Periods of slightly elevated sodium to chloride molar ratio were followed by shutdown and contaminant removal or periods of low sodium to chloride molar ratio.
- Steam generator blowdown demineralizers, a primary source of sodium during Cycle 7, were not in service during Cycle 6.
- Since blowdmyn was not recycled during Cycle 6, the presence of organic chloride from makeup water resulted in lower sodium to chloride ratios.
- Cycle 8 steam generator blowdown sodium to l chloride molar ratios have recently been l maintained around 0.4_ i l
1
- Feedwater copper concentrations have decreased from an average of 0.06 ppb in Cycle 7 to around i 0.02 ppb in Cycle 8.
V. C. Summer - SG B Blowdown Na / Cl Molar Ratio - Cycle 6 9 g ._ 7 - l 6 - 1 e y5 - 4 - n j 2 3 - i , s g a 2 - l i i 1 \ j
~
f-
~
l (+ 1
! h-[
s
' '~^ ' ' 1 0 l 07-May-90 15-Aug-90 23-Nov-90 03-Mar-91 11-Jun-91 19-Sep-91 28-Dec-91
~
1990 / 1991 l l
Comparison of V. C. Summer Cycle 7 and Plant C Cycle 6 Causative Circumstances: V. C. Summer Plant C , Reinitiated blowdown demineralizer Initiated full flow, deep bed operation condensate polishing Sodium leakage from demineralizers Elevated sodium concentrations 12 ppb to 20 - 30 ppb spikes l Alkaline blowdown sodium to chloride ratios indicate a caustic Hideout return indicates free caustic SG crevice environment environment in SG crevices Copper present in tube OD oxide Sludge on FDB due to chemical film and on crack face - requires cleaning and spalling off tubes - transport of dissolved copper to residual oxidized copper due to tube surface incomplete dissolution during chemical cleaning.
~
Unfavorable, low probability crevice geometry influenced copper and caustic accumulation n _ _ _ ___ _ ____________.___________-___________-______.._____________._.______.-__----------------_--_-----u
Comparison of V. C. Summer Cycle 7 and Plant C Cycle 6 Remedial Actions: V. C. Summer Plant C Optimize blowdown demineralizer to limit Suspended condensate polisher sodium leakage operation Maintain boric acid treatment Crevice flushing with boric acid, low temperature boric acid soak, on line boric acid treatment Removed largest secondary system copper Plug all degradation indications source , Administrative leak limit of 350 gpd I Blowdown sodium to chloride molar ratio l control Continuous air ejector monitoring and .{ projections following 25 gpd change Plug all degradation indications Administrative leak limit of 150 gpd Continuous air ejector monitoring and trending following significant increase in radiation monitor cpm Maintain at least 1 PORV operable Reduction in allowable coolant activity by factor of 2
il Comparison of V. C. Summer Cycle 7 and Plant C Cycle 6 Results: V. C. Summer Plant C Copper transport reduced by No reported large indications and i factor of three and still declining number of indication 3 reduced Sodium to chloride molar ratio l currently maintained at around 0.4. 1 1 Continued safe operation through to end of Cycle 8 l l
-____--__-_-_____-______________________________________________________________-a
Remedial Actions to Address Chemistry and Copper Transport
- Reduced copper transport to steam generators !
MSR tube bundles replaced in RF-7. Initial reduction of copper concentration in feedwater by a factor of 3 and still declining. e Reduced potential for caustic crevice environment Steam generator blowdown demineralizer operation modified to reduce potential for sodium emry to steam generators. Monitoring Na / Cl molar ratio on a daily basis. Current ratio in the range of 0.4.
- Goal to control molar ratio to s;0.7 in bulk water to reduce crevice caustic corrosion potential Existing procedures (response to out of spec chemistry) to be modified for ratio control Method (chemical addition or alternatives) to control molar ratio to be implemented
Conclusions
- Elevated corrosion levels were likely to have been caused by copper transport and an alkaline environment - corrosive conditions are enhanced at partially packed FDB crevices. i
- Copper transport has been decreased by MSR tube bundle replacement.
- Actions to control alkalinity accumulation are being implemented.
- Partially packed FDB crevices are more susceptible to unusual growth over short periods of chemistry imbalance due to the greater surface area of the crevice deposits and higher superheat which facilitate accumulation of impurities.
- With remedial actions undertaken, the risk of additional elevated corrosion reoccurring is substantially reduced.
~
V.C. SUMMER NRC MEETING JULY 27,1993 e TUBE INTEGRITY ASSESSMENT l T.A. PITTERLE WESTINGHOUSE NUCLEAR SERVICES DIVISION
r BURST PRESSURE GUIDELINES R. G. 1.121 NORMAL OPERATING CONDITIONS . e R. G. 1.121: Section 3.a.2
- The margin between maximum internal pressure to be contained by the tubes during normal olant conditions and the pressure that ,
would be required to burst the tubes should remain consistent l with the margin incorporated in the design rules of Section 111 of the ASME code". F e Generally interpreted as 3 APuo = 3996 psi e Applicable at normal operation which includes FDB and TSP constraint ACCIDENT CONDITIONS e R.G. 1.121 : Section 3.a.3
- " Loadings associated with a LOCA or SLB.... accommodated ,
with margins determined by the stress limits specified in NB-3225 of Section ill of the ASME code...." e Generally interpreted as ~ 1.4 APste o Applicable with degree of tubesheet constraint effective under SLB conditions
q ~ BURST PRESSURE GUIDELINES i ACCIDENT CONDITIONS i l
-j PORV AVAILABILITY !
e Single active PORV adequate to limit pressure increase in accident conditions e Containment isolation in accident conditions also cuts off air ; supply to PORVs j
- Must be manually restored to PORVs in accident sequence j e SLB Simulator runs made to assess operator actions l
- SLB: 4 crews with and without a rupture j
- All events terminated with PORV air restored to limit AP to' 2335 l psi e FLB Simulator Runs (4 crews) l
- 3 crews limited AP to 2335 psi 1
- 1 crew restored air to limit AP to 2420 psi !
e Stuck Open Safety Valve (3 crews) !
- Terminated with a AP range of 1600 to 2300 psi 1
FOR CONSERVATISM, APsts = 2560 psi (CONSERVATIVE SAFETY - 1 VALVE OPERATION) USED FOR TUBE BURST ANALYSES AND i 2335/2560 psi (1.4 APsta = 3269/3584 psi) USED FOR LEAKAGE. I ANALYSES l
-l
d FDB DISPLACEMENT IN SLB EVENT SLB TSP DISPLACEMENT ANALYSIS e Catawba-1 analysis (WCAP-13494, Rev.1) applicable to V.C. Summer FDB e Common loads and FDB support structure FDB WELL SUPPORTED TO MINIMlZE DISPLACEMENT e 4 stayrods in hot leg I
- 4 segments welded to partition plate
- 8 vertical bar supports at plate edges FDB DISPLACEMENT e Displacemen: < 0.1" for > 95% of tube locations
- All S/G B locations with FDB indications
- Displacements up to 0.26" for a few locations l
l
. a l
l l l Overall Finite Element Model Geometry 3-28 l
1 e h WWE.1 Y n 8 i [ g L 2 M E,4 T-i i b N e
}i FDB CONSTRAINT FACTOR EFFECTS ON BURST CAPABlUTY PRESENCE OF FDB WITH 0.072" DIAMETRAL CLEARANCE PROVIDES CONSTRAINT INCREASING BURST PRESSURE o Burst tests performed on EDM slots to define constraint factor on burst capability as function of TW crack extension outside FDB e Significant affect although less than smaller clearance TSP holes FDB CONSTRAINT FACTOR
- Ratio of Burst Pressures with FDB constraint to free span burst e TW crack lengths of 0.70" and 0.75" tested e Burst capability approximately independent of crack extension outside FDB up to ~0.2"
- Crack within FDB (FLB Application)
- Mean constraint factor = 1.72
- Crack extending 0.1" outside FDB (SLB Applications)
- Constraint factor at lower 95% confidence = 1.54 I
i
Burst Pressure of Tubes with Degradation within the FDB 0.750" x 0.043" Alloy 600 SG Tubes, Prototypic Tests
-., a, b, e-N eum a lCGE_BHST XLW)8urat & Butse (Meybel Page1 nrx if24/93, a 22 Pu
___, _ , . ____--c- tv+--* * * #"'"* _ _
l l Pulled Tube Burst Capability for Indications Within TSP
- . a , b, e-1
+
1 I
I i i i BURST CAPABILITY ASSESSMENT EOC 7 BURST CAPABILITY l 1
- Assessed for pulled tube results adjusted for FDB constraint and j temperature i
- All results exceed R.G.1.121 guideline for 3 AP uo i
i CONSERVATIVE ASSESSMENT FOR CYCLE 8 ; i
- Conservative Assumptions j
- Same crack growth as Cycle 7 l l
- Same indications occur on tubes with lower tolerance limit (LTL) l properties
, - APsts = 2560 psi rather than expected 2335 psi i i e All expected burst capabilities exceed R.G.1.121 guidelines !
- includes a postulated 0.75" TW crack !
i
- Even assuming the invalid measured burst capability for R28C41 '
leads to a 1.54 factor on SLB burst at 2560 psid (with LTL properties, a 1.34 factor is maintained) l I i
.l l
Summary of V.C. Summer Burst Capability Burst Pressure - R.T. Adjustment Factors Burst Pressure Ratio Tube Basis Value FDB Temp. Mat. Burst 3aP=3996 SLB = 2560 Constraint Prop. Capability 3 APuoand FLB at EOC 7: Crack Within FDB R33C20 Actual 5082 1.72 0.94 1.0 8216 2.06 3.21 R42C43 Actual 3618 1.72 0.94 1.0 5850 1,46 2.29 R28C41 Expected 3249 1.72 0.94 1.0 5253 1.31 2.05 SLB at EOC 7: Crack O.1" Outside FDB R33C20 Actual 5082 1.54' O.94 1.0 7357 (1.84) *
- 2.87 R42C43 Actual 3618 1.54 0.94 1.0 5237 (1.31) 2.05 R28C41 Expected 3249 1.54 0.94 1.0 4703 (1.18) 1.84 SLB at EOC 8 With LTL Properties Assuming Cycle 7 Indications
~
R33C20 Expected 4739 1.54' O.94 0.92 6311 (1.58) 2.47 Actual 5082 1.54 0.94 0.92 6768 (1.69) 2.64 , R42C43 Expected 3906 1.54 0.94 0.80 4523 (1.13) 1.77 Actual 3618 1.54 0.94 0.80 4190 (1.05) 1.64 R28C41 Expected 3249 1.54 0.94 0.87 4092 (1.02) 1.60 , 0.75"TW Expected 3024 1.54 0.94 0.91 3984 (1.00) 1.56 R.G.1.121 Guideline 1.0 1.4 Constraint factor at lower 95% confidence
- Values in parentheses at accident conditios are not R.G.1.121 guideline
...,-c..- ...,,4- ,-c,,w-.--, . . . - ---_--<iv~,-----=ir--m--- ~- ,,--e-----y. -----,----,i,, .--.-m,-4. - . . - - . - - - .m-< ~ ,_ -,-g %---- r- ,---+m-+..~. w., , , - _ .-<- e_.- - w _ . __ - _ - _ _ _ __- _ - _ _
TUBE LEAKAGE GUIDELINES ALLOWABLE SLB LEAKAGE PER STANDARD REVIEW PLAN
- Factor of 10 margin against 10CFR100 limits e Conservative iodine spike models of the SRP e Allowable SLB leakage of 8.6 gpm for primary coolant activity of 1.0 pCi/gm DE l-131 ALLOWABLE SLB LEAKAGE LIMIT OF 17.2 gpm FOR CYCLE 8 e Coolant activity limited to 0.50 pCi/gm e Cycle 7 activity <0.05 pCi/gm !
t l l l 1
SLB TUBE LEAKAGE ASSESSMENT ANALYSIS METHOD
- Log leak vs log (volts) linear fit to V.C. Summer R33C20 and R42C43 data points
- Consistent with upper range of APC database ,
- Linear fit well within 90% prediction interval of APC database
- Apply linear fit to 5 largest EOC 7 indications
- Assumes 7.6 and 5.9 volt indications have probability of leakage = 1.0
- Next largest indication is only 3.1 volts (POL ~ 0.3)
SLB LEAK RATE RESULTS AT EOC 7
- Leak rate of 3.9 gpm at expected AP = 2335 psi is less than allowable limit of 8.6 gpm at Technical Specification coolant activity level
- Leak rate of 11.8 gpm at conservative AP = 2560 psi is less than allowable limit of > > 17.2 gpm for actual Cycle 7 coolant activity l
2560 psi SLB Leak Rate vs. Bobbin Amplitude
'3/4" Tubes, Model Boiler & Field Data
_., 6, e. - N i 34Lb560 SPS
- . . . . ~ _ . _ . . . . . . - . . . . . . _ . . . . . . . . , _ . _ _ . . . , . _ . . . . . . . . _ . _ . . . . . . . _ . . _ _ . . . . _ . . . . . . . . - . - . . . _ . . _ . . . . . . _ _ . . . _ . . - . . . . . . . . ~ . . . . . . . _ _ . . _ _ . . _ . , _ _ . . _ . . _ , _ _ . . _ . . . _ _ . . . _ , . . _ _ _ . . . . _ . . . _ _ _ _ , _ .
I i l POTENTIAL SLB LEAKAGE FOR FIVE LARGEST INDICATIONS AT EOC 7 2335 psid 2560 psid Volts 1/hr gpm 1/hr gpm ' 5.9 11 0.05 26 0.11 7.6 24 0.11 60 0.26 ' i 9.8 52 0.23 137 0.60 l 11.8 92 0.41 252 ,1.11 22.9 707 3.11 2210 9.73 i Total 886 3.91 2685 11.8 w 1 4 b J
SLB TUBE LEAKAGE ASSESSMENT J CONSERVATIVE ASSUMPTIONS FOR CYCLE 8
- Same crack growth and 5 large indications as Cycle 7 e APst, = 2560 psi SLB LEAK RATE CONCLUSION FOR CYCLE 8
- Conservative SLB leak rate estimates for Cycle 8, like Cycle 7, result in less than allowable limits
I l OVERALL TUBEINTEGRITY CONCLUSIONS i EVEN THOUGH LARGE CRACK GROWTHS OCCURRED ON A FEW TUBES IN CYCLE 7, TUBE INTEGRITY WAS MAINTAINED AT EOC 7
- Burst capability satisfying R.G.1.121 guidelines e Potential SLB leakage less than that resulting in a small fraction of '
10CFR100 dose limits using SRP methodology REMEDIAL ACTIONS IMPLEMENTED AT V.C. SUMMER FOR CYCLE 8 CAN BE EXPECTED TO RESULT IN: e Reduced crack growth comparable to that from prior operation and typical domestic plant experience ;
- Enhanced margins against radiological limits even if it is postulated that a SLB event occurs subsequent to reoccurrences of large growth rates FULL CYCLE 8 OPERATION TO SCHEDULED REFUELING IS ACCEPTABLE i
i b
i l l 1 REMEDIAL ACTIONS PLANT CONTROLS TO ENHANCE SAFETY MARGIN OPERATING LEAK LIMIT REDUCED TO 150 GPD '
- Increase leak before break capability ENHANCED LEAKAGE MONITORING GUIDELINES TO INCLUD5 TRENDING ANALYSIS FOLLOWING LEAKAGE ALARM
- Increased detection of potential rapidly propagating crack with reduced time to plant shutdown.
INCREASED OPERATOR TRAINING / SENSITIVITY TO S/G TRANSIENTS
- Further enhance likelihood of appropriate / timely response to S/G .
transients such as secondary pipe break events. REDUCTION IN COOLANT ACTIVITY LIMIT BY FACTOR OF TWO e Reduce potential radiological consequences of a secondary pipe break. MAINTAINING 1 PORV AVAILABLE (UNBLOCKED) AT ALL TIMES , DURING CYCLE 8 e increase confidence that S/G Tube AP in a transient will be 2350 psid with associated reduction in radiological consequences i ALL FLAW INDICATIONS FOUND IN 1993 WERE REPAIRED e Reduced likelihood of significant indications left in service subject to potential for large growth than following 1991 inspection
V C Summer Nuclear Station l Requirements Compliance , O Technical Specifications RF-7 Inspection & Repair Demonstrate S/G are Operable O R.G.1.121 Burst Capability Demonstrated Pulled Tube Test Results Together with FDB Restraint Satisfied 3 OP/1.4 SLB R.G.1.121 Burst Guidelines. Remedial Actions Further Increase Margins Over Cycle 7 9 Operating Leakage Limits Set Operating Limit at 150 gpd Enhance Leak Before Break 9 10 CFR 100 Dose Limits Reduce Allowable RCS for Accidents Activity. Demonstrated Pulled Tube Results are Within Limits. Remedial Actions Further Increase Margin over Cycle 7. n.......n.,
V. C. SUMMER NUCLEAR STATION
.. s ..
l Action Result Status OPERATING LEAK LIMIT REDUCED TO 150 GPD INCREASED LEAK BEFORE COMPLETE STP-114.002 BREAK CAPABILITY REDUCED RCS ACTIVITY LIMIT REDUCES POTENTIAL RAD. PROCS. TO BE REVISED BY 08/15 CONSEQUENCES OF A -(STP-604.001) SECONDARY PIPE BREAK (CP-602) & (CP-614) MAINTAIN 1 PRESSURIZER PORV AVAILABLE
~
INCREASE CONFIDENCE COMPLETE STATION - THAT S/G TUBE APIN A ORDER (50-93-12) TRANSIENT WILL BE 2335 PSID WITH ASSOCIATED REDUCTION IN RAD. CONSEQUENCES IMPROVED OPERATOR TRAINING / SENSITIVITY ENCHANCE LIKLlHOOD OF LICENSED OPERATOR RE-QUAL. TO APPROPRIATE TIMELY TRAINING TO BE COMPLETE IN S/G TRANSIENTS RESPONSE TO S/G NEXT CYCLE - 8/9 TO 9/11 TRANSIENTS ENHANCE PLANT RESPONSE CRITERIA BASED ON EARLIER PLANT RESPONSE PLANT PROCEDURES TO BE RAD. MONITORS SENSING S/G TUBE LEAKAGE TO RAPIDLY DEGRADING REVISED BY 08/31 S/G TUBING
V. C. SUMMER NUCLEAR STATION Action Result Status ENHANCE NDE CAPABILITY & REPEATABILITY - STD. NDE METHODOLGY COMPLETE - RF7 WCAP 13522 TRACABLE TO INDUSTRY DATA BASE REMOVE MSR TUBE BUNDLES SIGNIFICANTLY REDUCED COMPLETE - RF7 COPPER TRANSPORT CHEMISTRY ACTIONS REDUCES POTENTIAL FOR CAUSTIC ENVIRONMENT A) 5/G BLOWDOWN /DEMINERALIZER OPERATION A) COMPLETE MODIFIED B) MONITOR Na/Cl MOLAR RATIO DAILY B) MONITORINGIN PROGRESS PROC. CP-613 TO BE REVISED BY 08/15 C) MODIFY EXISTING PROCEDURES FOR CONTROL C) PROC.CP-613 REVISED OF MOLAR RATIO BY 08/15 D) IMPLEMENT METHODS TO CONTROL MOLAR D) METHODS TO BE RATIO DETERMINED & IMPLEMENTED BY 08/31
i! I f 't V C Summer Nuclear Station l fj . i e Plant was Safe in Cycle 7 - l L e Plant will be Operated Safely in Cycle 8 l t, Effective Remedial Measures to Prevent Reoccurrence of Atypical Cracks Additional Plant Controls Enhance ; Safety Margins e No Midcycle Outage is Required ! l A}}