ML021330264
| ML021330264 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 04/30/2002 |
| From: | Harden P Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| N-566-1 | |
| Download: ML021330264 (49) | |
Text
Committed to Nuclear Excellence Palisades Nuclear Plant Operated by Nuclear Management Company, LLC April 30, 2002 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 DOCKET 50-255 - LICENSE DPR PALISADES NUCLEAR PLANT PRIMARY COOLANT PUMP CASING STUD EVALUATION - CODE CASE N-566-1 By letter dated August 26, 1999, the Nuclear Regulatory Commission (NRC) approved for Consumers Energy Company the use of American Society of Mechanical Engineers Boiler and Pressure Vessel Section Xl Code Case N-566-1, "Corrective Action for Leakage Identified at Bolted Connections,Section XI, Division 1," as an alternative to the requirements of paragraph IWA-5250. This code case allows inspection and evaluation as an alternative to casing stud removal when determining suitability for continued operation with degraded bolted connections. Nuclear Management Company, LLC (NMC) is notifying the NRC that NMC is using the provisions of code case N-566-1 to support continued operation of primary coolant pump P-50C for the Palisades Plant until the next refueling outage. The prescribed evaluation was performed in accordance with subparagraph IWB-3142.4 and is submitted in accordance with subparagraph IWB-3144(b). Based on the enclosed evaluation, use of the code case for this application is acceptable. contains the overall evaluation of the degraded bolted connections and contains the NMC engineering analysis that calculates the degradation of the bolted connections and determines acceptability for continued operation.
27780 Blue Star Memorial Highway
- Covert, MI 49043 Telephone: 616.764.2000 P\\,O
SUMMARY
OF COMMITMENTS This letter contains one new commitment and no revisions to existing commitments.
The new commitment is:
The primary coolant pump P-50C casing joint will be visually inspected and assessed at each plant shutdown to below Mode 4 conditions, until repairs can be completed in the next refueling outage.
Paul A. Harden Director, Engineering CC Regional Administrator, USNRC, Region III Project Manager, USNRC, NRR NRC Resident Inspector - Palisades Enclosures
ENCLOSURE1 NUCLEAR MANAGEMENT COMPANY PALISADES NUCLEAR PLANT DOCKET 50-255 April 30, 2002 Primary Coolant Pump Casing Stud Evaluation 7 pages follow
Primary Coolant System COMPONENT:
Primary Coolant Pump P-50C CLASS:
ASME Class 1 INTRODUCTION:
By letter dated August 26, 1999, the Nuclear Regulatory Commission (NRC) approved for Consumers Energy Company the use of American Society Mechanical Engineers (ASME) Boiler and Pressure Vessel Section Xl Code Case N-566-1, "Corrective Action for Leakage Identified at Bolted Connections Section Xl, Division 1," as an alternative to the requirements of paragraph IWA-5250. This code case allows inspection and evaluation as an alternative to stud removal when determining suitability for continued operation with degraded bolted connections. Leakage in primary coolant pump (PCP) P-50B was the subject for application of the requirements of code case N-566-1 as approved on August 26, 1999, by the NRC. Nuclear Management Company, LLC (NMC) is using the provisions of code case N-566-1 to support continued operation of P-50C for the Palisades Plant until the next refueling outage. The prescribed evaluation is being performed in accordance with subparagraph IWB-3142.4 and submitted in accordance with subparagraph IWB-3144(b).
FUNCTION:
There are four PCPs (designated P-50A, B, C and D) in the primary coolant system (PCS). They are Byron-Jackson Company designed vertical, single suction, centrifugal pumps. The casing joint design and service conditions are the same for each pump. During normal operation, the four pumps circulate water through the reactor vessel that serves as both coolant and moderator for the core.
NUMBER, SERVICE AGE AND MATERIAL OF BOLTING:
Each pump is part of the PCS pressure boundary. The casing material is ASTM A 351, Grade CF8 low alloy steel. The 16 pump studs (numbered 1 through 16 sequentially around the pump casing) securing the upper casing to the lower casing are ASTM A 193, Grade B7 low alloy steel. The threaded portions of the studs are chrome plated. The stud material is considered susceptible to boric acid wastage.
CORROSIVENESS OF LEAKING MEDIUM AND ENVIRONMENTAL CONDITIONS:
The leaking medium is primary coolant. Service conditions inside the pressure boundary are 2060 psia and approximately 537 OF. PCS Boron concentration SYSTEM:
varies from approximately 1600 ppm at beginning of cycle operating conditions to potentially 0 ppm at end of cycle operating conditions. Leakage is characterized as a fine steam and water mist when the PCS is at operating conditions. The leak rate is not directly measurable. The leaking medium is characterized as moderately corrosive under these conditions.
LEAKAGE LOCATION:
Evidence of minor leakage was identified from boric acid accumulations in the vicinity of the pump P-50C casing joint prior to and during the 2001 refueling outage. Evidence of leakage was confirmed during surveillance pressure testing at the conclusion of the 2001 refueling outage (May 2001). The leakage was reconfirmed during forced outage (FO) 01-5001 in December 2001 (FO 01-5001 began in June 2001 and concluded in January 2002). The leakage at the joint was identified as the source of boric acid build-up on the PCP casing studs adjacent to the component cooling water (CCW) piping at the rear of the pump.
There is no equipment other than the PCP studs and CCW piping in the immediate vicinity that is affected by boric acid accumulations.
LEAK RATE:
The total PCS unidentified leak rate determined on January 29, 2002 at 0320 hours0.0037 days <br />0.0889 hours <br />5.291005e-4 weeks <br />1.2176e-4 months <br /> was approximately 0.01 gpm. The leak rate at the pump P-50C casing flange contributing to this total cannot be determined. Therefore, a conservative treatment of the leakage would attribute the entire leakage from the casing flange area.
LEAKAGE HISTORY AT SIMILAR CONNECTIONS:
Plant Experience A comparison of the FO 01-5001 pump P-50C leakage with historical data related to similar leakage in pump P-50A (Palisades condition report CPAL9801939) is provided in the following table and text.
Parameter P-50C (FO-015001)
P-50A (CPAL9801939)
Leak Location Between studs No.16 Between studs No.16 &
& No.1 No.1 Stud diameter 4.575 - 4.590 inches 4.575 - 4.590 inches Stud material A-193, Gr. B7 A-193, Gr. B7 Stud No.16 minimum 4.238 inches 3.720 inches measured diameter Stud No.1 minimum 4.482 inches 3.700 inches measured diameter PCS leak rate 0.01 gpm Approx. 0.02 gpm
A review of the table above indicates the occurrence of leakage for pump P-50C is similar to that of pump P-50A.
Industry Experience Industry experience indicates the pump P-50C leakage will not lead to catastrophic failure. Electric Power Research Institute (EPRI) Nuclear Maintenance Applications Center (NMAC) publication TR-1 02748, "Boric Acid Corrosion Guidebook," April 1995, describes operating experience associated with similarly designed PCPs at the Fort Calhoun Plant. Fort Calhoun observed significant wastage on three adjacent studs on two pumps due to a leak in the same area (under a CCW pipe). The wastage on some of the studs at Fort Calhoun was such that only 20% of the stud was left remaining. This wastage was identified during a plant shutdown. The wastage on the Fort Calhoun studs was much more severe and affected more studs than is the case at Palisades.
This information was also contained in NRC Information Notice, IN-80-27, "Degradation of Reactor Coolant Pump Studs."
VISUAL EVIDENCE OF CORROSION AT THE ASSEMBLED CONNECTION:
Current Physical Condition ASME Section Xl, 1989 Edition, Paragraph IWB-3517 requires stud removal when the level of wastage exceeds 5% of the cross-sectional area. The actual level of wastage is conservatively estimated to be 14% of the cross-sectional area for stud No.16 and 5% of cross-sectional area for stud No.1. These values are conservative because the minimum measured diameter was applied to the entire stud circumference. Field measurements indicate the degraded area of stud No.16 extends its entire circumference to varying degrees. The minimum measured diameter was 4.238 inches, while other areas of this stud show less wastage. The degraded area of stud No.1 measures only approximately three inches along the circumference, and a minimum measured diameter of 4.482 inches. Therefore, the actual reduction of cross-sectional area for stud No. 16 is less than 14% and for stud No.1 is much less than 5%. Stud dimensions are provided in the following table:
Pump P-50C Measured Nominal Measured Minimum Stud Number Diameter (inches)
Diameter (inches)
No.1 4.584 4.482 No.16 4.579 4.238
Results of Visual InsDections Pump P-50C studs were inspected during FO 01-5001. After cleaning, all studs other than No.s 15, 16, 1 and 2, were inspected for wastage and found to be in good condition. These four studs were VT-1 visually examined and measured.
Only studs No.16 and No.1 were found degraded in the area just above the pump casing. Wastage did not continue into the stud hole or up toward the upper casing flange.
Additional Inspections Additional inspection of pumps P-50A, P-50B and P-50D casing flanges and studs was performed. Stud diameter measurement and visual inspection was performed. Evidence of minor leakage was discovered at the pump P-50B casing flange in the same area of interest. However, inspection, and evaluation in accordance with IWB-3142.4 for joint integrity, indicated all pump P-50B studs were clean and no wastage was occurring.
Results of Other Test/Inspection Methods An ultrasonic test has not been performed on the degraded studs. This method is not effective in detecting stud wastage of the magnitude affecting pump P-50C. Visual inspection and examination indicates no stud breakage in the accessible areas. Inaccessible areas are not affected by wastage due to lack of oxygen and fluid flow as documented in EPRI TR-102748, Section 3.2.
Threaded areas are chrome plated, further reducing susceptibility to wastage.
TIME COMPONENTS HAVE BEEN DEGRADING:
Wastage is conservatively assumed to have begun after start-up from the cycle 14 refueling outage in December 1999, which is the last time pump P-50C is known to have been free from significant boron accumulation. Calculations are based on the 18 months from the end of the cycle 14 refueling outage to the beginning of FO 01-5001 in June of 2001. From these 18 months, three months are subtracted for forced and refueling outage time with the PCS below 21 0°F.
Industry experience has shown that wastage rates are most severe when metal temperatures are near 212'F. A review of plant operating history records for this period of time indicates the PCS was below 21 0°F for approximately 59 days. An additional 30 days is subtracted to add a level of conservatism.
CHARACTERISTIC ADVERSELY AFFECTED:
Leakage is adversely affecting the stud cross-sectional area at the pump P-50C casing flange. The studs identified as No.16 and No.1 are adversely affected.
The maximum measured wastage is determined to be 0.341 inches. Using this wastage and the 15 months of wastage time, NMC engineering analysis EA CPAL0104122-01 calculates an approximate wastage rate of 0.023 inches per month (0.272 in/yr). This value is similar to the wastage rates documented for pump P-50A. This wastage rate is reasonably consistent with those found in EPRI TR-1 02748, Test Reference F (page 4-23) and Test Reference K (beginning on page 4-34). A comparison of all available wastage rates associated with pumps P-50A and P-50C and available rates from the EPRI guidebook are contained in the following table:
NOTE: All wastage rates are converted to units of inches per year (in/yr) to be consistent with EPRI NMAC data.
Reference Date Stud No.1 Stud No.16 Data Source Document Wastage Wastage EA-05/22/98 Not estimated Not estimated Initial EA for CPAL981067-01 P-50A.
EA-11/19/98 0.216 in/yr 0.384 in/yr Based on actual CPAL981939-01 measurements at two dates for P-50A.
CPAL990588 05/12/99 0.120 in/yr 0.264 in/yr Based on actual measurements when studs were retired from P-50A.
EPRI Test N/A Minimum Maximum Immersion at Ref. F 0.108 in/yr 0.124 in/yr 212°F & 4000 ppm boron.
EPRI Test N/A Minimum Maximum Immersion at Ref. F 0.042 in/yr 0.050 in/yr 352°F & 4000 ppm boron.
EPRI Test N/A Minimum Maximum Directed steam Ref. K 0.639 in/yr 0.833 in/yr spray; stud temp. < 1750F.
Unknown boron.
EPRI Test N/A Minimum Maximum Directed steam Ref. K 0.042 in/yr 0.050 in/yr spray; stud temp. < 3500F.
Unknown boron.
EA-12/29/01 0.474 in/yr 0.273 in/yr Assumed for CPAL01 04122-P-50C. Higher 01 rate assumed for lesser wasted stud for conservatism.
COMPONENT REPAIR SCHEDULE:
Component repair is presently scheduled for the next refueling outage.
RISK OF FAILURE:
The primary structural concern to be addressed is degradation of the pump joint integrity due to casing stud wastage resulting from the contact of boric acid on carbon steel. Stud wastage could result in increased leak rates that may exceed the limits of Technical Specification (TS) Limiting Condition of Operation 3.4.13, "PCS Operational Leakage," or radiological effluent releases greater than limits specified in TS 5.5.4, "Radiological Effluent Controls Program." NMC performed an engineering analysis to assess these studs and the integrity of the pump P-50C casing joint using field measurements. By comparing operational data with that associated with the previous pump P-50A leakage, and by reviewing inspection records, a wastage rate for pump P-50C was determined. Using this wastage rate, a linear extrapolation was made to the next refueling outage. The wastage rate is reasonably consistent with those found in EPRI TR-1 02748 and NMC's experience with pump P-50A. The calculation determined that joint preload was maintained at a level that ensures operational requirements are met and structural integrity is maintained.
Probability of Casing Joint Failure The casing leak rate is relatively small and stable and may slowly degrade, which would be detected by PCS leak rate monitoring. Industry and NMC's experience indicates the casing joint will not fail catastrophically. Based on information provided in EPRI TR-102748, Section 8.0, and due to the location of the degraded studs (adjacent to each other) the joint will exhibit leakage prior to a catastrophic failure. Therefore, the probability of failure is very low.
Consequence of Degraded Stud Failure NMC has determined from expected wastage rates that joint integrity will be maintained. The results of the finite element analysis described in EPRI Publication NP-5769, "Degradation or Failure of Bolting in Nuclear Power Plants," of similarly designed pumps constructed of similar materials and with a 16 stud casing flange similar to the Palisades' pumps, indicates that if the 2 studs in question were to fail, the adjacent (and all other) studs would remain intact. This publication also displays through the graph on page 8-31, an expected leak rate of approximately 10 gpm following failure of two adjacent studs. This leak rate is significantly less than Palisades' charging system capacity for PCS makeup (33 to 133 gpm).
Core Damage Frequency (CDF) or Large Early Release Frequency (LERF)
The risk due to the postulated failure of two adjacent studs is that of a required, controlled plant shutdown, since the expected leak rate following the failure is significantly less than Palisades' charging system capacity for PCS makeup. A conservative number of controlled plant shutdowns are incorporated in the NMC Probabilistic Safety Assessment for Palisades. The increase in CDF or LERF from operating in this condition, due to the possibility of a controlled plant shutdown, is negligible.
IMMEDIATE ACTIONS RECOMMENDATION:
Inspection Schedule The Pump P-50C casing joint will be visually inspected and assessed at each plant shutdown to below Mode 4 conditions, until repairs can be completed in the next refueling outage.
Protective Measure to Prevent Further Wastage No additional protective measures are recommended for pump P-50C casing studs. Protective measures, including zinc coating and shielding, applied to pump P-50A were demonstrated to be ineffective in surviving the operating environment and protecting the studs from wastage.
LONG-TERM ACTIONS:
Component repair is presently scheduled for the next refueling outage; therefore no additional actions are necessary.
CONCLUSION:
NMC concludes that the Primary Coolant Pump P-50C casing joint meets operational requirements based on the current corroded condition and using a reasonable rate of degradation.
Pump P-50C is considered operable based on visual inspection, engineering analysis of the degraded studs, industry experience with leakage at similar pump flanges, and the approved use of code case N-566-1. The analysis supporting this operability recommendation indicates pump P-50C will remain operable at least until the next refueling outage.
ENCLOSURE2 NUCLEAR MANAGEMENT COMPANY PALISADES NUCLEAR PLANT DOCKET 50-255 April 30, 2002 Engineering Analysis EA-CPAL0104122-01 Evaluation of Wastage on Studs Between Casing and Cover of Pump P-50C 37 pages follow
consmmI Power PMNA MESS PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS COVER SHEET EA-CPAL0104122-01 Title Evaluation of wastage on studs between casing and-cover of Pump P-50C 16-+ A41k e INITIATION AND REVIEW Calculation Status Preliminary Pending Final Superseded 0
0 6
0 Initiated Int Review Method Technically Rev Rev Appd Reviewed Appd CPCo Description By Detail Qual By Appd By Date Alt Review Test By Date Cal c 0Ori Issue 012 V
B 6/0
/0 1 DSRIAT 01 o
Attachment 1-Stud Measurements, NDE Suplementary sketch etc.
Attachment 2-CPAL 0104122 Attachment 3-Stud Design Configuration, Byron Jackson dwg.
Attachment 4-Record of Telecon with Flow Serve (Byron Jackson)
Attachment 5-Fax from flow Serve dated 12/18/98 Attachment 6-ASME IIl, 1965 Table N-422 Attachment 7-Palisades FSAR Table 5.2-3 (Sheet 2 of 22)
Attachment 8-Combustion Engineering - Engineering Specification #70P-005 for Primary Coolant Pumps Attachment 9-BW/IP International Inc. to Consumers Energy, STUD ANALYSIS,12/20/98 0- Primary coolant Pump 50C, Casing Leak Timeline to Forced Outage 01-5001 1-EPRI, NMAC, Boric Acid Corrosion Guidelines Report TR-102748, 4/95 2-ASME 1995 Section III Appendices, FIG. 1-9.4, Design Fatigue Curves 3-Primary Coolant Pump dwg.
t/#.*
consum s
Power NHMf PALISADES NUCLEAR PLANT EA-CPAL0104122-01 ANALYSIS CONTINUATION SHEET Sheet 9
pev # n Reference/Comment 2
1.0 OBJECTIVE 3
4 The objective of this EA is to evaluate the acceptability of 5
the Primary Coolant Pump P-50C joint between the casing and 6
the cover with local degradation of two of the 16 studs.
Each 7
stud is 4.58" diameter.
CPAL0104122 documented the condition 8
of these studs during Forced Outage 2001 and that condition 9
report will determine operability.
10 11 12 2.0 APPLICABILITY 13 14 This EA is applicable to studs for any of the four (4) 15 Primary Coolant Pumps at Palisades, provided the specific 16 degradation and the degradation rate is enveloped by the 17 conditions evaluated in this EA.
18 19
3.0 REFERENCES
21 22 2.1 ASME B&PV Code Section XI, 1986 Edition.
23
2.2 Drawings
M1-EA-5, M1-EA-5001, M1-EA-2006 24 2.3 Byron Jackson Tech manual, Vendor file.
25 2.4 ASME B&PV Code Section Ill, 1965 Edition.
26 2.5 Palisades FSAR, Rev. 23, Table 5.2-3, Sheet 2 of 22 27 2.6 EPRI Report (EPRI TR-104213s), Bolted Joint 28 Maintenance and Application Guide. December 1995 29 2.7 Telecon Record with Flow Serve dated 12/18/98 30 2.8 Fax dated 12/18/98 from Flow Serve.
31 2.9 ASME B&PV Code Section IIl, 1998 Edition 32 2.10 AISC Steel-Construction Manual, 8th Edition 33 2.11 Marks' Handbook, page 5-30, Ninth ed.
34 2.12 EA-C-PAL-98-1939-01 Rev 0 35 2.13 Structural Engineering Handbook by Gaylord Jr., 1968 36 edition, pp 6-52 thru 6-54 37 2.14 NRC letter to Mr.
NL Haskell,"Evaluation of Inservice 38 Inspection Program Releif Request No.
39 MA4420)", dated January 28,1999 40 43 44 45
Cans.uml Power FOMMU PALISADES NUCLEAR PLANT EA-CPAL0104122-01 ANALYSIS CONTINUATION SHEET Szhan 11 Paw J n 1
4.0 DESIGN CRITERIA 2
Reference/Comment 3
4.1 References 2.4 and 2.5 provide the design criteria.
4 5
4.2 Original detailed design calcalations for the pumps 6
are not available.
7 8
9 10 11 5.0 DESIGN INPUT 12 13 5.1 During the forced outage in year 2001, the P-50C pump 14 casing to cover joint was inspected to assess the 15 condition of the studs and document the condition of 16 local degradation of the two studs.
Measurements 17 show that the stud in position # 16 is 0.341 in 18 thinner on the diameter and stud in position # 1 is 19 thinner by 0.102".
The corrosion is presumably from boric acid leaking from a degraded casing to cover gasket in the area. This adverse condition is 22 documented by CPAL0104122. See Attach 2.
23 24 25 5.2 The Primary Cooling Pump P-50C is one of the four 26 pumps that circulate water through the reactor.
The 27 joint between the casing and cover of the pump is 28 characterized by 16, 4.58" diameter studs which 29 reflect a measure of redundancy in bolt reactions. The 30 studs are made of A-193, Grade B7 material with flash 31 chrome plate and phosphate coating on the tap end.
32 The nut material is A-194 Class 2H.
The casing and 33 driver mount/cover material is A-351 Grade CF8M and A 34 216 Gr WCB respectively. See Attach. 5 and Ref 2.2.
35 36 5.3 The original stud dimensions are given in Reference 37 2.2.
Per Reference 2.3, the applied preload stress on 38 the stud is 25,000 psi. This stress level is about 1/3 39 of the stud's yield strength of 75 Ksi per Reference 40 2.4, and stud's deformation during preload is well Al within the elastic range.
43 The stud has upset threaded ends, 4.73" in diameter 44 per Attach 3.
45
Consumhi Power POMEs.
Mr A PXMGMMF.S PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-CPAL0104122-01 Qhnt 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 16 17 18 19 0
22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 A 1 43 44 A
Dx/ -4 n
-m a
Reference/Comment 5.4 5.5 5.6 5.7 5.8 The degraded stud configurations are depicted in of this EA.
Only two studs have visible degradation.
There is no visual evidence that the casing, cover or the chrome plated threads are degraded. VT-1 was performed to determine the corroded dimensions.
In the current corroded condition, Stud # 16 measures 4.238 inches and # 1 measures 4.482 inches in minimum diameter.
Time line regarding casing leak for Pump 50C is shown on Attach 10. Wastage is assumed to begin after start-up from REFOUT 14 in 12/1999, which is the last time pump 50C is known to have been free from significant boron accumulation. Calculations are based on the 18 months from the end of the REFOUT 14 to the beginning of force outage 01-5001 in June 2001.
From these 18 months, 3 months were subtracted for forced and refueling outage time with the PCS below 210 degrees F. A review of Reactor Engineering records for this period of time indicates the PCS was below 210 degrees F for approximately 59 days. An additional 30 days is subtracted to add a level of conservatism.
Design Code for the pump is ASME 1965 ed. Use of later editions of the ASME Codes is acceptable because:
A. Design philosphy has not changed.
B. The allowables are consistent.
C. Additional information has been added to enhance/clarify the code.
In the corroded condition, conservatively,the studs will experience less than 10 operational (heatup/cooldown) load cycles.
§ onismM Power PMEu MW PFRMUSW PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-CPAL0104122-01
.R Paw A n
-U ____
5.9 6.0 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 16 17 18 19 Reference/Comment a
This evaluation is not proposed as a design analysis.
It represents an assessment of operating data and physical component characteristics combined with design code methodology to determine component adequacy for a finite period of time i.e. one fuel cycle.
ASSUMPTIONS Major:
None Minor:
The total wastage took place in 15 months as stipulated in para 5.6. The corrosion from now until Refout 2003 is considered to be linear with time.
Because the next refueling outage is 15 months away, the corroded diameter is expected to be (4.238-0.341) 3.897 in. This assumption is similar to the the degradation rate assumed for Pump P-50A in 1998. The analysis and the "RELEIF REQUEST" was submitted to NRC.
(See Ref. 2.14)
The height of the corrosion for stud # 16 is taken to be uniform over a 1.5 inch length as shown on Attach
- 1. Stud # 1 corroded condition is considered minor compared to # 16. Therefore, the analysis for stud #
16 envelopes stud # 1.
The other studs adjacent to the studs being evaluated have no visual signs of degradation as evidenced by the data in Attachments 1.
/I 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 A I 43 44 45
consummrs Power FRMM PALISADES NUCLEAR PLANT EA-CPAL0104122-01 ANALYSIS CONTINUATION SHEET qhppt
- r.
Pow i n 7.0 ANALYSIS Reference/Comment 2
3 The ASME Codes stipulate that the bolted joint be designed for 4
effects due to preload, operating pressure and any 5
differential thermal expansion stresses.
The code does not 6
specify how these stresses are to be combined.
7 8
During discussions with Byron Jackson (Flow Serve, see REF 9
2.12) Palisades was informed that if a Finite Element Analysis 10 model is performed for the pump casing, stud and cover to 11 include the effects of preload and system pressure loads, the 12 pressure loads will not add significantly to the calculated 13 bolt (stud) stresses. Generally, the high strength bolts may 14 be tightened to 70% of the specified tensile strength (see Ref 15 2.13).
16 17 The studs are over 4.5 inches in diameter, which makes them 18 considerably stiff members of the joint. The pump is connected 19 to massive piping and supported with very stiff structural "I
members/components.
1 22 23 24 THERMAL EXPANSION 25 26 Because, all of the components are at or very close to the 27 same temperature ( same fluid), there is very little relative 28 thermal expansion. Additionally, the thermal expansion loads 29 are of secondary nature and self releiving. Therefore, these 30 loads are neglected. (Ref. 2.9 para NB-3213.13 (1)b).
31 32 33 34 35 SEISMIC LOADS 36 37 Seismic inertia loads are relatively small because the massive 38 structure will experience a very low ground acceleration for 39 the horizontal direction and even less for the vertical 40 direction.
.2 43 44 45
consumem Power PRO69M PALISADES NUCLEAR PLANT EA-CPALO014122-01 ANALYSIS CONTINUATION SHEET Sýhppt 7
Paw J nl 1
DEAD WEIGHT A WReference/Comment 2
3 Dead weight of the motor affects the loads on the studs and 4
will be considered in the analysis. The motor dead load 5
affects the pre load of the studs.
6 7
8 9
10 CRITICAL DESIGN FEATURES 11 12 13 The critical design consideration is the preload for 14 connections using bolts or studs.
The applied preload is 15 intended to be greater than the operating load on the studs.
16 This ensures adequate compression of the joint and no leakage.
17 As the bolt or stud material is removed by corrosion, the 18 strain energy and, thus the preload force is reduced.
19 However, the reduction can be accepted if sufficient preload zi force is maintained.
Preload is a displacement-dependent 22 load.
With a smaller stud cross section, the stiffness of the 23 stud is reduced and the stress in the stud increases.
24 25 Because, the operating loads on the bolts do not always behave 26 as theoretically designed, EPRI Report for "Bolted Joint 27 Maint. And applications Guide" ( Ref. 2.6) provides guidance 28 to calculate a "f" factor which depends on the relative 29 stiffness of the fastener and the joint members.
This "f 30 factor is applied to the calculated pressure stress and the 31 resultant factored stress is combined with the preload stress 32 in the bolt, to compare with an allowable bolt stress.
EPRI 33 developed these rules to integrate the practical aspects with 34 the theoretical design.
35 36 37 FUTURE WASTAGE 38 39 As stipulated in para 5.6, the wastage started in December 40 1999 and was noticed in June 2001, a duration of 18 Al months.There was approximately three (3) months downtime 2
during that period. So, it took 15 months of operation to 43 corrode stud # 16 by (4.579"-4.238") or 0.341".
Assuming the 44 plant will run for another 15 months until Refout 2003, the 45 stud may corrode further.
Power M9 PALISADES NUCLEAR PLANT EA-CPALO014122-O1 ANALYSIS CONTINUATION SHEET h
_ Pa g n 1
Anticipated corrosion = 0.341"*15mo./15mo. = 0.341" Reference/Comment 2
3 Anticipated corroded stud diameter is 4.238"-0.341" = 3.897" 4
5 This wastage rate is higher than 0.13 in/year as recommended 6
in EPRI Report TR-102748 ( Attach. 11 ). The higher wastage 7
rate would cause the stud to be thinner and produce higher 8
stud stresses. The degradation rate for this EA is similar to 9
past rate for Pump P-50A. See NRC letter dated January 28, 10 1999. (Ref 2.14) 11 12 Although the second stud (# 1) has corroded less than the 13 first, it will also be conservatively assumed to corrode to 14 3.897" between now and Refout 2003.
15 16 The measurement of the stud diameters taken for CPAL0104122 17 does not show any degradation on the other studs adjacent to 18 these two studs.
Therefore, the joint will be evaluated 19 considering only two studs being degraded.
Z1 Figure 1 on page 9 is a conservatively postulated cross 22 section of the degraded stud used in this EA to analyze the 23 impact of corrosion of the two studs on the pump joint.
24 25 26 27 28 PRELOAD STUD STRESS 29 30 View AA on page 9 shows the stud configuration used to 31 calculate the effective stiffness of the corroded stud with 32 respect to the non-corroded stud.
33 34 Preload Force, F = K x 35 36
- Where, A = stud elongation in inches 37 K = Stiffness = EA/L 38 E = modulus of elasticity 39 L = length 40 A = Cross-section Area 2
The corroded stud cross section is considered as shown on page 43
- 9. Because the wastage is uniform all around the stud, there 44 are now two cross sections that need to be considered.
- Also, 45 there is a 0.5" diameter hole in the stud. (See Attach 3)
Scommers Power PE-Mr MWrrl APuWS PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-CPAL0104122-:i qhoat
-I F/:/ 4 ak6 I
g4eo- #
Cf~
/D IL(
14 1 i 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 16 17 18 19
?30 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 AI
.341 CT 1.0 )
-4 A-A
,z i1 A55i A-5C654AT*'/C 1r Y q/
flý Reference/Comment
- 4 rrr---
Power PMEAMM Caffiutn PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET 1
2 3
4
-5 6
7 8
9 10 11 12 13 14 15 16 17 18 19 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 41
.2 43 44 45 (See L1 =
Attach 3 and Ref 2.2) 1.5L/17 and L2 = 15.5L/17 K1 = EA1/(1.5L/17)
K2 = EA2/(15.5L/17) 1/Ke = 1/K1 + 1/K2
= (L/17E)*( 1-.5/A1+ 15.5/A2) 1/Ke = (L/17EA)*[1.5/0.72 + 15.5/1.0]
1/Ke = (L/17EA)*17.5833 Ke = (17/17.5833)* (EA/L) = 0.9668 (EA/L)
Therefore, the effective stiffness of the 96.68 percent of the original stud.
corroded stud is Overall stiffness of the joint = (14 X 1 + 0.9668 X 2)/16
=0.996 EA-CPAL0104122-O1 1n n D,,,,
Reference/Comment For a stepped member, the stiffness is calculated as two series connected springs.
The effective stiffness Ke is given by 1/Ke =1/K1 + 1/K2 KI = EA1/L1 and K2 = EA2/L2 Al = Area of reduced cross section minus Area of 0.5" dia hole
= (1/4* 3.8972)
(n/4* 0.52)
= 11.928 - 0.2
= 11.728 sq in Cross sectional area of the original stud A2 = A n/4 (4.582 -0.52) = 16.28 sq.in, A1/A2 = 11.728/16.28 = 0.72 The actual thickness of the bolted joint is 18".
The 17" dimension is the length of the stud between the upset ends.
f)
-a Id -
66r.16 W
consumme Power MO PALISADES NUCLEAR PLANT EA-CPALO014122-OI ANALYSIS CONTINUATION SHEET Sheet I1 PON/ A N 1
This shows that the joint effeciency with the local 2
degradation on two studs is 99.6 percent of what it would be Reference/Comment 3
if there were no degradation.
This shows a very insignificant 4
decrease in the overall effectiveness of the joint and is, 5
therefore, acceptable.
6 7
The preload stress in the corroded stud 8
9
= 25 X 0.9668/0.72 = 33.57 kSI 10 11 where 0.72 is the area ratio of the locally corroded stud to 12 the non-corroded stud and 25 KSI is the initial preload.
13 14 15 Differential Thermal Expansion Stress 16 17 Most of the stud length is in the cover material and they both 18 expand at the same rate. The pump casing material is 19 different from the stud material. But, because only a small portion of the stud is inside the casing, thermal stresses, if 21 any, are very small due to minor differential thermal 22 expansion. Stresses due to this expansion are secondary in 23 nature, generally self relieving and very low.
Therefore, 24 these stresses are deemed to be acceptable without further 25 evaluation. See Ref 2.9 para NB-3213.13 (1)b.
26 27 28 29 30 31 Seismic/ Dead Weight Design Considerations 32 33 Even though, there will be some horizontal and vertical loads 34 due to the earthquake, the loads will be very small because 35 this pump is attached to massive and very stiff piping and 36 supports and the combined response of the system and 37 components reflects rigid body motion. For this reason these 38 loads can be ignored. The external dead weight loads have been 39 considered in the pressure stress calculations. All pertinent 40 data is part of the Palisades FSAR Section 5.7.5.1. Following S1is a brief history of the NSSS seismic design:
43 The initial seismic design of Palisades NSSS components was 44 conducted in a very simplistic, static manner.
In the 1986 45 time frame, Palisades elected to use the ASME Section III, I
Power MW PMEUS MMnORm PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-CPALO014122-01 1 0 D-,
4 0'
1 2
3 4
5 6
7 Reference/Comment The implementation of these curves required the determination of seismic anchor movements for piping attached to the primary coolant system.
At about the same time, it was necessary for Palisades to determine time history input to the reactor vessel in order to conduct nonlinear seismic analysis on new fuel bundles.
Each of these analysis demands required that the primary coolant system vessels and piping be decoupled from the stick model for the development of an interaction model.
The vessels themselves (Reactor Vessel, Steam Generator, Pressurizer and Primary Coolant Pumps) were characterized by stick model members.
This enabled the analyst to calculate seismic loads directly at critical cross sections.
This experience demonstrated that the interaction of the primary system components with the concrete internal structure was such that the combined system/structure (original lumped mass) model was adequate for overall structural response and seismic loading as previously calculated by the static ZPA methods was acceptable. The seismic loading was very modest and noncontrolling with respect to the other load cases and the seismic anchor movements of the primary coolant components could be ignored. This is the basis for the conclusion that seismic loadings cases and combinations are not limiting or significant in the analysis of primary system components.
Pressure Stress Operating pressure = 2060 psi Operating pressure-force = (n/4)*482*2060 = 3,727.7 Kips where the effective pump diameter for pressure is 48 inches.
(See Attach 4)
Load of equipment on the joint = 122.3 Kips Net pressure force = (3727.7-122.3) equals 3605 Kips Force per stud = 3605/16 = 225.3 Kips Tension stress in the corroded stud = 225.3/11.728 = 19.21 Ksi where 11.728 is the cross sectional area of the corroded stud in square inches (see Preload stud stress calculation).
10 11 12 13 14 15 16 17 18 19 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 43 44 45
Sconsume Power S M.,
A MKWNS 5IM PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-CPALO014122-01 Ckxd 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 16 17 18 19 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 43 44 45 Le = grip length + 1/2 engaged thread thickness.
length + 1/2 bolt head Le = 18 + 10/2 + 4.58/2
= 25.29" 4.58" is the assumed thickness of the nut.
Le/D = 25.29/4.58 g 5.52 From Fig 6-17, page 6-23 of Reference 5.52, the ratio Kj/Kb is 3.
2.6, for Le/D equal to From page 6-21, equation 6-4 of Reference 2.6, S= Kb/(Kj + Kb) 14* = Kj/Kb +1
= Kj/Kb +1
=3 +1 i
Reference/Comment Allowable stress intensity, Sm at 600 deg F = 19.8 Ksi (Attach 6)
Pressure stress is below the ASME 1965 Code Sm.
Stress Combinations Operating Pressure + Preload For design pressure or operating pressure conditions, the pressure stresses and differential thermal stresses are not directly additive to the preload stress. to this EA shows the original design details of the stud.
Reference 2.6 provides guidance to calculate the factor by which the pressure stresses should be modified.
Refer to page 6-21 of Reference 2.6
(ýk I
D,', -
4
consumeu Power MER M PMNSP PALISADES NUCLEAR PLANT EA-CPAL0104122-01 ANALYSIS CONTINUATION SHEET qheaf
-A Paw # n_
14 Reference/Comment 2
3
- =0.25 4
5 6
The preload stress plus pressure stress in the locally 7
corroded stud = 33.57 + 0.25 X 19.21 = 38.37 Ksi 8
9 10 Reference 2.4 specifies the allowable stress at the operating 11 temperature equal to 39.6 Ksi which is two times Sm at 600 12 degrees operating temperature.
13 14 15 The calculated stress in the corroded stud using the operating 16 pressure and preload is 38.37 Ksi which is less than the 17 allowable stress intensity.
This is acceptable and, 18 therefore, the corroded stud can take the stress imposed on it 19 by the preload and operating pressure.
21 22 Bolt Eccentricity 23 24 For stud # 16, there is no eccentricity because the wastage is 25 360 degrees around the stud in a uniform fashion.The fact that 26 the wastage is only on one side of the stud #1 may cause an 27 eccentric load path through the corroded length of the stud.
28 But the wasted area for that stud is much smaller and the 29 height and the width of the wastage is also less than one-half 30 of stud # 16. So, the insignificant eccentricity may cause 31 small local bending in the stud, but the resulting moments 32 will be minimal and do not impact the load carrying capacity 33 of the studs as shown below.
The two adjacent degraded studs 34 being less stiff than the other 14 may cause a slight joint 35 asymmetry. But this is too minute to be a concern.
36 37 38 Assume that there is a 0.25 in eccentricity from the center 39 for the bolt load path.
This will provide additional design 40 conservatism for stud # 16.
Ii
.2 NOTE: Similar evaluation was performed for Pump P-50A in 1998 43 where there was eccentricity. P-50A analysis was submitted to 44 NRC in 12/98. See Ref. 2.14 45 i
Pwer rMDUSm PALISADES NUCLEAR PLANT EA-CPAL0104122-O1 ANALYSIS CONTINUATION SHEET qhppt r;
pP,, j n 1
NOTE: Similar evaluation was performed for Pump P-50A in 1998 Reference/Comment 2
where there was eccentricity. P-50A analysis was submitted to 3
NRC in 12/98. See Ref. 2.14 4
5 Max. Bending moment M due to eccentric path 6
= 38.37* 11.728* 0.25 7
= 112.50 in.kip 8
9 Moment of inertia, I = n/64(3.8974- 0.504) = 11.3181 in4 10 11 Section Modulus, S = n*(3.897 4 0.504) / (32*3.897) 12
= 5.8086 in3 13 14 Bending stress due to moment = (112.50) /( 5.8086) KSI 15
= 19.368 Ksi 16 17 18 Stud Cyclic (FATIGUE) Evaluation 19 1
NB 3222.4 (e) of Reference 2.9 states that a strength 21 reduction factor for fatigue need not be greater than 5. So by 22 using a factor of 5, 23 Peak Bending Stress = 19.368* 5 = 96.84 ksi 24 25 26 Total Peak Stress = 38.37 + 96.84 = 135.21 KSI 27 28 Peak Alternating Stress = 135.21/2 = 67.60 KSI 29 30 31 FIG. 1-9.4 of Reference 2.9 shows two design fatigue curves 32 for high strength-steel bolting for temperatures not exceeding 33 700 degrees F. One curve relates alternating stress values 34 and number of cycles for maximum nominal stress less than or 35 equal to 2.7 Sm. The other curve relates alternating stress 36 values and number of cycles for maximum nominal stress equal 37 to 3.0 Sm. Using the conservative curve the number of allowed 38 cycles is more than 1000 cycles. See Attachment 12.
39 40 Number of cycles, experienced by the studs in the corroded 1
condition is less than 10. Therefore, usage factor = 10/1000 =
,2 0.01 which is extremely small as compared to 0.1 allowed in 43 the industry practice.
44
Consumms Power NWIMAI"P PALISADES NUCLEAR PLANT EA-CPALO014122-OI ANALYSIS CONTINUATION SHEET qheet
-A Pow i n 1O SReference/Comment 2
8.0 CONCLUSION
3 4
Based on the current corroded condition and using a linear 5
rate of degradation until REFOUT 2003, it is concluded that 6
the Primary Coolant Pump P-50C joint meets the ASME Section 7
II1 allowables and, therefore, complies with the FSAR 8
requirements.
9 10 This EA has made some conservative assumptions fom the 11 analytical standpoint such as the number of cycles for the 12 studs and assumed stud eccentricity. Based on this analysis, 13 it is concluded that sufficient preload still exists (and will 14 exist until such time when the studs are replaced i.e. 3/2003) 15 to prevent the joint from separating under operating 16 conditions.
17 18 19
FA-CPALI o1o4.iz2-oj Lakntyw~wy NDE a
EA-CPALO14-1z z-c I EXAMINATION OF PCP CASING FLANGE BOLTS CPAL0104122 (These instructions are provided for guidance. Field personnel may add information as conditions may reauire?*
For P-50C, perform Visual Examination VT-1 and record thinnest dimensions for bolts 15, 16, 1 and 2 as shown in Permanent Maintenance Procedure PCS-M-47, Page 71.
EM VT-1 Complete Micrometer Readi g Indicate by a/
(inches) 1,4, A,.
P-5OC 15 16-16-LV 1-1-
/i.s 7 2-2-
'.3'-q.(579 Visually scan the remainder of the casing flange bolting for boric acid accumulations. IE boric acid is present, AMD the location is accessible, clean, perform Visual Examination VT-1 and obtain micrometer readings per the following table. Otherwise indicate Not.
Required.
EUM VISUL GAQ CLEANED MUCRQMETIE iAPRESENT?
Yes or L
READiNG (Bolt #
Yes, No or Not Required or Yes or and inches)
Not Accessible Not Accessible Not Required Not Accessible P-50C 3-3-
3 4-4-
4 5-5-
5 6-6-
6 7-7-
7 8-8-
8 9-9-
9 10-10-10 11-11-11 12-12-12 13-13-13 14-14-14- -_
C:\\WNDOWS\\DESKTO*NMEWBRJI~ !LANGEIN.WPD 1,
consumer--
A A -CA
'I "'
Z 3 -4o 5 Laboratory Se NDE -Visual Examination Report Examiner: BrianLenlus Level: II Receipt Date: 8.2901 FExam Date: 12.2401 lReport Issue Date:
ISheet No.: BJL-01 Examiner: re/a Level: n/a NDE Company: Conuvan.n EWy Total Hours Worked: Wea Project No.: 0100510 Customer Location/Address: Pa*sad*s NucOer P*
atExamination Location: 0 Laboratory CM Customer Facility 127780 Blue Star Mem. Hwy. Covert, Mi. 49043 NDE Procedure: NDT-VT-01 Rev: 14 Method:
Direct Q Remote _
Technique: N VT-1 3 VT-3 [3 Other:
Surface Condition: as found Evaluation Requirements Light Meter: 006931 lCai. Due: 6.26.02 Intensity at Surt 200ft.
Code:
Year:
Seti on:
Piart I
IProcedure Adequacy Demonstrated:
ASME 1989 Xi 1141"517 Observation Distance.: 12,to 24" jObserlvaton Anglwe: 4y N
Other
Reference:
re Visual Aldo Eamination Checidlat F i Scatcheu/Cuts/Gouges
[
Connection Integrity
[I Weld Profile Flashlight Mirror I
0 Lack of Fusion 0 Weld Reinforcement C1 Wear Material Type: CS Joint Design: Wa Ruler Pit Gauge 0
C Unear Indication
[
Physical Damage
[3 Undercut Fillet Gauge 0
Contour Gauge 0
[
Physical Displacement El Clearance Verification [
Cracks Nominal Diameter: 4.75w Nominal Thickness: 3w Binoculars 0
B Loose/Missing Parts
[
Erosion
[
Corrosion Item Type: Pump Stud El Debris Other Vlsual: Ofher00w Abnonnlle Item ID System Indication Lo.
Indication
. Weld
.E0valuaio L
- 1.
WOI1O No. n I
Number Name Line No.
No.
Location 1
Leg Longthe Acceptab.e Rue Tye $1 No.
Remork 1
PCP P5OC 1
o" erosion 3"
n/a n/a No 4
EVeNe of/ eakeg was noted a&nwW statfs 101 and I 2
PCP P60C n/a n/a n/a n/a wn/
n/a Yes.
4 15 PCP PSOC wa n/a n/a n/w n/a n/w Yes 4
16 PCP PSOC ION" erosion 16" n/a n/a No 4
See aftchment BJL.01 for sketch.
,2LA Accreditation Certificate Number 1097.03 for ANSI B31.1, AWS Di.1, AWS DI.3, md ASME SecUon V. Aricle 9.
Report shall not be reproduced4coog in full. woku the w
wrMi aiprows! of Consumnru Enwgy.
Examiner:e 1 of
..Reviewed BYQ j 1
1
-ovel:
]Date:
Revision 09/0 1 41ý,X
Stud #16 taken on December 22, 2001
-4
Stud #1 taken on December 22, 2001 vt-I2
--. ;S r-4 0*
http:. pipwn05 sý.base. reports cr rpt htm 'cVJi-=
P \\1_
Condition Report EA.C
-CPAL 0/,4 CPAL0104122
Title:
BORIC ACID ACCUMULATION NEAR PRIMARY COOLANT PUMP P-50C COVER CCW FLANGE Discovery Date and Time: 2001-12-21 08:00 Condition Discovered By: HEALTH PHYSIC3 System(s) Affected:
CCS PCS Component(s) Affected:
P-50C Description of Occurrence or Condition:
On December 21 "A" Shift, Health Physics personnel observed an accumulation of material below the Primary Coolant Pump P-50C cover Component Cooling Water (CCW) connection flange.
This is located at approximately 618' elevation, behind the Cýun mechanical seal, outside and below the motor driver mount.
Until very recently, the area was only accessible by traversing slippery pump insulation covers, or by passing directly by the pump shaft when the pump was tagged out (now an expanded metal grating allows safer access).
The only known potential sources of such an accumulation in that area are boric acid from the PCS or nitrites from CCW.
There were also whitish coatings on smaller piping above the accumulated buildup, a piece of rag, and possible wire debris.
Immediate Action Taken:
The observer took digital photographs of the area and reported observations after exiting.
The System Engineer was notified upon arrival for day shift.
A planner with primary coolant pump engineering and maintenance experience inspected the area later that morning.
The System Engineer, planner, ISI lead, Health Physics, and a System Engineering section head met that afternoon to plan cleanup, fastener inspection, and leak inspection.
The team, which had extensive experience with primary coolant pumps, agreed that further inspection was necessary for fastener evaluation and possible leakage, but the accumulation probably was old boric acid that had built up from many years of pump seal leaks and drips from seal instrument lines.
A System Engineer jump was planned for "A" Shift December 22 after PCS pressurization to 250 psia.
System Engineer Walkdown Report:
Buildup appeared to be old boric acid.
Deconners had cleaned the pump flange so stud-cover interfaces were visible in the vicinity of the CCW line.
There was moisture around one stud, but it appeared to result from cleaning, not a leak.
No flow was noticed.
A deconner stated afterward that considerable liquid had been used to try to clean off the area.
One stud had what appeared to be relatively shallow surface corrosion, and there was some pump cover wastage around that stud hole.
The lower two of four CCW studs were still partly buried in the boric acid "rock" and could not be more closely examined.
There was no evidence of CCW leakage; the pressurized pipe and gasket area were dry.
The P-50D CCW flange area was also examined since it is the only pump that has not been opened since initial criticality.
There was no boric acid buildup right at the CCW flange, but there was considerable old boric acid buildup a few inches away beneath insulation covers.
Loosened boric acid had fallen to the 607' area below the pump.
P-50D studs did not appear degraded.
There was some maintenance debris in the area.
Health Physics was notified.
Recommendations (Operability and corrective Action):
History:
P-50C internals were replaced in 1985.
Since then, the pump cover and CCW flange have not been disassembled.
Some dry buildup was visible from the 625' floor near P-50C at the beginning of the present outage, but it did not appear to be any different from what has been seen before.
A close examination of the flange area was not made at that time for safety reasons.
12/26/2001 10:14 AM I of 2
A *{,.
4
.5t4 Z
http: plpwnt05 s-vbase reports crTthtm'pa!C?
,L 6A -CPAL o 412Z-01 Ar n-pumps nave tended tt azC=U' mu e
Cr::
- z tne Pas-.
A c:ose examination of all primary coolant pump :over ln:es
=n not show any evidence of cover leakage except on 2-50A :ref. C;AL}
1333; Recommendations:
- 1. (Decon Crew)
Remove more boric acid to fully expose the two lower CCW studs an:
fasteners for examination.
- 2.
(Engineering Programs)
Evaluate CCW fasteners, rusted pump cover stud, and stud hole for wastage & acceptability.
- 3.
(Operations)
If Engineering determines fasteners to be acceptable, consider P-50C OPERABLE.
- 4.
(Engineering)
After final cleanup, take more pictures.
Compare to 2003 REFOUT pictures to determine if boric acid buildup is continuing.
References:
CPAL980000963, CPAL9801067, CPAL9801080, CPAL9801939 Problem Resolved Yes 17 No Initiator:
BEMIS DA Origination Date:
2001-12-22 02:50 Does the condition involve an equipment or programmatic issue related to the ability of an SSC to perform its safety or safety support function?
Yes - Complete Admin 4.13 Attachment 1 fl No Immediately Reportable?
C Yes - Complete Admin 3.03 Attachment 5 C No Safety Assessment per Maintenance Rule Policy Required? fl Yes 17 No Reportable:
Cl Yes El No 10CFR Part#
PRC: U Yes 11 No Licensing
/
Maintenance Rule Applicable?
[] Yes fl No Significance Leveler (Circle one) 1 2
3 4
Industry Experience? 03 Yes 13 No Does past operability need assessment? 0 Yes 0 No If yes, CRTL will ensure completion Comments:
N/A CRG Chair:
Date:
Assigned to:
Due Date:
Evaluated by:
Date:
Approved by:
Date:
CARB Chair Approval:
Date:
Closeout by:
Date:
12/26/2001 10:14 AM 2 of 2
17, 9 wso~ uomw N.Aow.".m
~
,!~"'"
- 9L
$love II 0 V,4
-'-CAO"4
-2~
.~Jo PINc "6Ut-Cft
'~~~~.a q
i~q
'-4 6
~
.4CA A-
, N0 G
tf;
- P 0
a -
Telephone Conversation with FkyesrV,, December 181 On December 18, 1996, Palsade conducted a telecon with ongineering pernnel frm Flowerve Coporation to discuss the draft review of EAC-PAL-Si-93i-01, "Ev,,lution of Corrosion on Studs between Cn and Cover of Pump P-50, Rmerence C-PAL-1939."
Flowserve participants Included Frank Coetanzo, Managor of Enginaeing, Pump Division; and Gerard Lenzen, Senior Project Engineer, Pump Division. Among the Palisades participants were Don Riot, N Lyon, Raj Gupta, and Don Bemis. Other people came and went during the conversation.
Floweerve comments on the draft calculation were discussed. In general, the commenft were similar to comments generated Internally at Palisades.
Flowserve assumes the hydraulic lifting are to extend to the outer diameter of the outer casing gasket, which is 48 Inches. This is based upon the conservative assumption that the inner gasket is nonfunctional, allowing full pressurization up to the outer gasket. They believed that a slightiy smaller effective diameter would be justifiable In accordance with standard gasket pnnacples.
The code maxdmum yield steb allowable for the stud at the assumed temperature (800 F.) is 61 ksi.
Flow*erve engineers said that we should not add bolt preoed stress to acua pressure stes.
Finite element analysis shows only a marginal increase In bolt st-ess due to pressure. The larger of the two stressas (pressure or preload) should govern design.
Section 2B of the original code stress report dealt with bolting design. Unfortunately, Flowserve cannot physically locate that section.
Based upon Flowuerve's eperience with pump maintenance, proed relaxation is very hard to quantify. Their suggested value is 15%. This is based upon tensioning the stud In order to determine the lift-off load.
Stud temperature would be close to pump operating temperature inside the flange. In the exposed area, it may be a hundred degrees cooler.Most plants Insulate tho studs and flanges, but those at Palisades are uninsulated. Due to the nature of the lceaized boiling. suiface temperature of our stud is cooler in the exposed area.
Flowserve also provided a fax detailing pump and stud materials, as well as gas dimnensWo.
4 I~iUMIRMIS LTitllL*
'am pWU
,,U-,,s;*.;,-*.W, U
IN
FAX TRANSMOTAL TO 211-M2M9 S*jm pallodu RCIP po ggow&g is som johmm" requemod an do PdbWa RCP.
Ddw m&jw I - A-216 Go& WCS cowmeedd
- A-351 Omb CNM cmumudd
- A-351 Omb CFIM ftdmsimdd
-A-l"GmftB7 Nu9mMWW
- A-194 Chm 2H otft
_ goo
-4CIDmd4g-VrOD 0 u w !p do!
A&MCIDW46-VICOD tirlDod4rOD
.43-9ACWwd44-1S/lIrOD Mwd*iý offt aisO.175*.
so 12129 12-1" 3: 29 MA 3m
- 4. 1 1.
JUINOf PWW tMW Cm~
we is I
ý
ý"
W -,!
P-M low I
P!IP ARTIC:LE 4 DESICN
.6
'0'l4
~TABLLE N.422 "Z IK r
DESIGN STRESS INTENSITY VALUES, S. FOR STEEL BOLTING MATEmA.S*
For motel tompaerluto net euceeding *F 200 300 400 500 600 700 l00 900 1000 33.000 31.400 30,300 28,00
- 27.700 26.300 24.400
- ,,-u 29.800 28,400 27,400 26.200 2S,100 23,800 22,100 23,500 22.400 21.600 20,700 19,800 18.800 17,400
.,f.4m) 34.000 33,200 32.500 31.800 30.900 29,400 27,700 30.700 301,00I0 29,400 28,'90 27,900 26,600 25.100
- .,.:1111 27,500 260900 26.300 25.:00 25.000 23.800 22,400
-..t'I3 33.000 31.900 30,600 29,500 28.100 26.400 24.200 WI TABLE H142 (in coure of preparation)
- 1 ii
'I
.4 LI
~
C I
t 4122--oi I Table t-422 SECTIONi111 NUCLEAR VSSEis -lt ot TABLE *4#.ý 09SM69 STREZSS NIMEJSITYi VALU tSi.
N OR STOLSLLI9'.AWLS' MIMIA T.mu.rttg:
YME ow
Spas N om in al D i m ter T
p..
Af Number Gr*.,
Comp** iltion In.
PSI'-s Low Alley Stees.
SA-193 7
1 Ct-0.2Mo 2
omdundter 12,006id0
.12S.00
,0 Or1/2r 2% Lo 4 Inc.
I 100 11S000.
9500w..
Ovt'r 4 to 7 lIc.
1100 I00,000 7.,OfR.
2% and under 1'200 125.000 105.000 SA-193 814 1 Cr-0.3Mo-V Over 24 to 4 Ic.
1200 110.O00 95,000.,
SA-193 B16 1 Cr-YMo-V Over 4 to 7 inc.
1200 100,000 MO.0OD SA-3VO L43 2NI-O.8Ct-4Mb 4 and tinder 12SO0 105,0M, aj'be allowable "suess valves lot boeling materials gives IN this table do not exceed the lesser of osthlkird of the specified emii~musm e
yield strenth ow sue-ild of the yield a..t at temperature, with credit ganted for the eahbacemegt of prupeitiesproducod by hast yestmeni. Tey ar inende** o o h. Nwin the design formulas of Appeodi-Io. For allowable velts of actual Pr*ied sad serviet st Oes.
see M-414.
4.4 A
'32 YIW..
- iy.
-
sat23 f~~Ive H B
'It'
e
- iii
! ! ! I E~ro MII H5 11,1 E
U U
P.I 0
a I
. TV'
-"I.
I,
( 0
-I "I
l's TI
oo C
I.
L I
I I!
I i 8I Fill
- q!v
r I I!
.it F
y I
mm.
p "i
- 0.
w w
u in
- 0.
0 t "
'L
[
mm I IK S
I w
.1 F
- 5. '1 I I I F
[ I I
till 11 I
I,
.1 p5!
K I'"
I..
A II
oow_
/0412 0
4.0 4.1 noe pump "aD be a welled cmntritho iyps with bottom sued" and horaconld disehuwi. The driving motor dial be of d conventional bmuto type.,
u duwon hll be made for-die collction and removal of -daft nal leakam' Sulb&~ mIclfoniitaInd lubrication -system fur pump mad molur mnust be Iuaekdsd with doe pump sociably. The pump and Its pcrormuice The duln Ulfa of the pump duall be forty yam. The predicted Ufe of does positions of the pump mmwnbly with Ico. dian aforty yam life sin be Holed In the tochinicad manual.
1im proswn containmenit desigm of die pump assembly didl conform to Setidon.LII of the ASME Boiler and Pmmsur. Veral Cods.
Vime pump esuing, pressure housing and auctien elbow shall be coodedered as i Clam *A" vame per paragraph 14-131 of Section 11t.
nom pump did be designd to require maintenance at a midmlmm intend of emj (1) yewr. It dd be a dutlp objeedve to require mdlommocme at dwse yew Provisions didl be made to change auk without drulinhg the pump ea**ng The pressure Inside the casing wil be betweem atuieqhahert andl 20p4&
noe pump amsembly didll be Apdupe-d for continuous operation at mny point an fth
-crcedteumv ifro 80,0 to I1.0 M
M hen in dug op..ll.. at uqwdf gravity 1.0 as qasfedWin 4.1.7.
.h i
"A3 heursi under dmoecodtin diall he within ledo "11ipuIdiu h
- 41A8 Fmp asommly dwd be despiged for ceunin dokwin rotadma OWNe ViWd
~from abom.
eM. 2 Fp 5., 81 47 0
4.1.2 4.1.3 Ir 4.1.4
- 4.1.6 I
K"tl 1%
O EFipmolpn RItenhruifclnc5
II "Ii" 9 IiiI U
iI I-l
PRIMARY COOLANT PUMP 50C CASING LEAK TIMELINE TO FORCED OUTAGE 01-5001 LA -CPA L 2.04
-22 Refuelina Outatge #13: June 1998 Refueling outage pressure test RT-71A did not record any indication of leakage or evidence of leakage in the area of interest for pump 50C.
Primary Walk-Down, December 26, 1998 System Engineering Mode 3 walk-down of P-50C indicated "No sign of leakage or boric acid buildup in the flange area corresponding to the P 50A leak.
Primary Walk-Down, May 8, 1999 System Engineering Mode 3 walk-down records indicate P-50A and P 50B were closely examined in the area near the CCW pipe. P-50D flange was not examined for safety & ALARA reasons. P-50C was examined from a distance (note: from 625' floor near pump). No evidence of leakage was recorded.
Refueling Outage #14: December 1999 Refueling outage pressure test RT-71A did not record any indication of leakage or evidence of leakage in the area of interest for pump 50C.
September 6. 2000 The System Engineer initiated work order 24013898 to clean boric acid from the CCW pipe flange in the area of interest for pump 50C. Activities were completed April 8, 2001. The CCW flange was determined to be in good condition. No cleaning or inspection was performed on the pump casing.
Refuelina Outaae #15: May 2001 Refueling outage pressure test RT-71A recorded an indication of evidence of leakage in the area of interest for pump 50C. However, no active steam plume or water dripping was seen.
Forced Outaae 01-5001, December 21, 2001 During Health Physics activities in the area of P-50C a boron accumulation was discovered at Component Cooling Water (CCW) flange located in the cooling water piping to P-50C (see CPAL0104122). After cleaning and inspecting, stud wastage was recorded (WO 24114404). No active leakage was seen.
EPRI Licensed Material Boric Acid Corrosion Guidebook Test Ref.:
Test Type:
Org/Date:
Reference:
"r"
- I (
EACAcPAL 01041ZZ-ol Immersion - Aerated Boric Acid Brookhaven National Laboratory (1982)
"Boric Acid Corrosion of Ferritic Reactor Components."
Brookhaven National Laboratory, July 1982, NUREG/CR-2827 (D).
Test Configuration, Procedure, and Results The testing performed by BNL was not described in any of the references. However, it can be inferred that A-193 Grade B7, AISI 4130 and 4135 low alloy steel specimens were immersed in an aerated environment containing borated water at various temperatures.
The average corrosion rates for these materials were then determined for test durations of 350-1300 hours.
Tests were conducted in solutions with 4,000 ppm boron and 4,000 ppm boron plus UOH to bring the solution to pH 7.3 which is close to typical PWR operating conditions.
The test cases and key test results are as follows:
Average Corrosion Rate (in/yr)
Temperature (°F) 4000 ppm boron 4000 ppm boron
+ LiOH to 7.3 pH 212 0.108 - 0.124 0.112 - 0.130 352 0.042 - 0.050 0.046 - 0.054 600 0.022 - 0.028 no data Key observations from these tests are:
"* The corrosion rate decreases for temperatures above 212°F.
"* Lithiated primary water has essentially the same high temperature corrosion rate as non lithiated water. This result is not consistent with testing conducted by others at lower temperatures. BNL hypothesizes that the increased ionic species present at higher temperatures increase reaction kinetics over that at lower temperatures.
Conclusions The main conclusion from these tests is that corrosion rates in aerated water near typical operating boric acid concentrations can be as high as 0.13 in/yr. Work by others has shown that this corrosion rate can increase significantly at higher concentrations.
5Pr
-jl[
iC
4 IJ 4-23 7--
o2-
-74 S
/,1/*%.'.P k9 5$
AL Poiec FIG. 1-9.4 DESIGN FATIGUE CURVES FOR HIGH STRENGTH STEEL BOLTING FOR TEMPERATURES NOT EXCEEDING 700"F Table 1-9.1 Contains Tabulated Values and a Formula for Accurate Interpolation of These Curves a'
1o2 S
-I S
iU 1.0 I
U' I z
I M
NOTE:
E -30 X o0 1W 7...
I!.
- I.
I.*;
-CpAl 0 o l6 4-10. -0 (
)
J
A F-
-., 4 ý ic 2 o(o-4ý22-o I 13 CA - CP/A Z-Sedlon A -A Drtw Couplim SpKw -000 C4WIrq Watw I n1d Somi CoWing HW Ex qe Connedlon Face SW lmpWlsr--..
CoWing wew Oudd Sucomoolector POWER COMPANY PALISADES PLANr
'FSAR UPDATE lmý pRINAAV COOLANT CIRCUUTINS PU14P F I OUK NO 44 1
51, IT, Area where lPImo casing zas*ketEs ar'e located.