ML18064A860
| ML18064A860 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 08/11/1995 |
| From: | Gamberoni M NRC (Affiliation Not Assigned) |
| To: | NRC (Affiliation Not Assigned) |
| References | |
| NUDOCS 9508180061 | |
| Download: ML18064A860 (53) | |
Text
I UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555--0001 LICENSEE:
Consumers Power Company FACILITY:
Palisades Nuclear Plant August 11, 1995
SUBJECT:
PALISADES REACTOR VESSEL ANNEALING MEETING
SUMMARY
A meeting was held at NRC Headquarters on July 24, 1995, between Consumers Power Company and the NRC to discuss the Palisades reactor vessel annealing.
This was the second working level meeting between the staff and the Palisades Annealing Team.
Consumers came into NRC headquarters on June 6, 1995, for the initial meeting on this issue.
The June 6 meeting consisted of the licensee presenting a broad overview of their plan and schedule for the annealing of the Palisades reactor vessel. A meeting summary was published on June 18 and is available in the Public Document Room.
I The July 24 meeting consisted of individual sessions to discuss a number of technical issues in more detail. These discussions included (1) thermal and stress analysis, (2) bioshield, (3) environmental qualification of equipment and instrumentation, and (4) fire protection. These technical areas are summarized below.
A list of attendees is provided as Enclosure 1. contains the slides presented at the meeting.
PRESENTATIONS/DISCUSSIONS Thermal and Stress Analysis The licensee and Westinghouse, the primary contractor for the thermal/stress analysis, described the results of a preliminary assessment of the feasibility of performing an in situ thermal annealing treatment of the vessel for a Westinghouse 4-loop plant.
The analysis is documented in EPRI report TR-104934 (March 1995) along with an analysis for the vessel in a 3-loop plant.
The analysis considers heating of the entire beltline region of the vessel (150 inches) to 850°F for 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> with an assumed linear decrease in temperature to 70°F at the vessel flange and lower head, respectively.
The vessel base plate material considered was A533B with 309 stainless steel cladding.
Three-inch thick Transco-type vessel insulation was assumed to completely surround the reactor vessel. The air gap between the vessel Afin/IJ\\J{)f insulation and the bio-shield wall was 3 inches.
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The preliminary analysis showed that the highest stresses generated were predicted to be at the nozzle safe ends.
A large differential bending moment.
is expected in the safe ends, but this can be controlled by limiting the thermal gradients and is predicted to be within the ASME Code allowable stress limits.
Code Cases N-47 and N-499 are the appropriate ASME Code references.
The elastic-pl~stic analyses showed that the strains (through-wall, local and surface) were predicted to be within the Code Case N-47 allowables.
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thermal analysis showed that the reactor pressure vessel (RPV) supports and primary system piping were predicted to be at lower temperatures than those experienced during normal operation.
The inside surface of the bio-shield wall was predicted to be below 150°F.
Thermal barriers in the nozzle were not simulated in the analyses but will be used at Palisades. These barriers will further reduce temperatures in the piping by limiting radiation and convection down the loops.
The staff discussed the possibility of finite element model mesh refinement potentially affecting the predicted stresses in the RPV wall.
Four elements were used across the vessel wall in the model that was presented and a comparison with an eight element mesh showed maximum differences at the outside diameter of the vessel.
The licensee took an action to further investigate the effects of mesh refinement on the stress distribution.
Consideration of the potential adverse effects of creep and fatigue were discussed.
The licensee considers creep effects to be negligible at the temperatures of the anneal.
However, creep will be addressed in the 20 stress analyses to be performed for Palisades.
Creep effects will not be considered in the 30 analyses unless they are judged to be significant.
Fatigue for the annealing process was considered a negligible effect since it is expected that the stresses in the annealing cycle will be bounded by the normal heatup/cooldown cycle. The licensee will provide confirmation in their submittal.
ISI during the annealing outage will include 100% of the reactor vessel beltline welds and base metal.
In addition, highly stressed areas, such as the RPV nozzl~ safe ends will also receive 100% inspection.
The licensee asked the staff if consideration of the effects of the annealing cycle on under-clad cracking needed to be considered i~ detail.
The staff response was that this should not be a major issue as the clad vessel (with piping not attached) had previously experienced higher temperatures (greater than 1100°F) during ~tress relieving and the predicted interfacial stresses were low.
The licensee expects to complete a preliminary thermal/stress analysis specific to the Palisades plant by the end of September 1995.
Environmental Qualification/Instrumentation Since the licensee has not yet finalized the routing of the ductwork, this
.. session of the meeting primarily consisted of the staff stating potential concerns that should be addressed in the licensee's submittal.
The concerns included (1) actions that will be taken to ensure that safety-related equ;itpment is not affected by the heat, (2) actions taken to ensure safety related equ~pment required to be operable during the annealing will remain
- operable, and (3) actions taken to ensure instrumentation (e.g. hot leg temperature sensors and neutron detectors) are not effected by the annealing.
- Concerns relative to heavy loads and placement of the burners on the sperit
.fuel pool building roof were also discussed.
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' *.! Structural Issues As discussed above, the licensee expects the bioshield temperatures during annealing to be less than the bioshield temperatures during normal operation.
This expectation is based on the plant not operating resulting in no gamma radiation and a reduction in thermal radiation from surrounding equipment.
The primary source of heat to the bioshield during the annealing will consist of thermal radiation and convection across the cavity between the reactor vessel insulation and the bioshield wall, approximately 32 inches.
The licensee installed thermocouples during the 1995 outage to monitor the temperatures during normal operation.
The results of these measurements and the results of the demonstration project to verify lower temperatures during annnealing will be key information in addressing the annealing's effect on the design basis of the bioshield.
Fire Protection The representatives from Cooperheat described the safety features associated with the gas train system such as the ultra violet eye to monitor the burner flame, the electrical interlocks to prevent flammable gases from entering the ductwork, the relief valves to prevent overpressurization, and the blocking solenoids and shutoff valves to terminate gas flow in the event of an abnormal condition.
All safety features are in accordance with the standards issued by the Compressed Gas Association and all safety components are listed for the intended use by Underwriters Laboratories. According to the Cooperheat representatives, the only gases that will be present in the ductwork and heat exchanger will be air, carbon monoxide at approximately.01% (lower flammable limit 12.5%), and nitrous oxide.
The staff advised the licensee's representatives during the meeting that the following areas are of interest and will be addressed during the NRC review:
(1) siting and quantity of flammable gas/liquid storage containers and appliances and exposure hazard to safety-related structures, (2) emergency shutdown systems/safety shutoff devices and venting provided for storage and appliances, (3) flammable gas/liquid piping (type and routing), (4) fire brigade training and resources to respond to a flammable gas leak or fire, (5) control of ignition sources in
~lant areas where flammable gas is stored, used or routed, (6) electrical installations associated with the gas train system, and (7) the monitoring provided for detecting leaks of flammable gas into safety-related structures.
SUMMARY
Following discussions on the technical issues, there was a brief discussion regarding the licensee's schedule for the thermal annealing report.
Palisades plans-to**submit the report in segments and expects the first*segment to be submitted this fall, with the final report to be submitted sometime next summer.
The staff requested that Palisades provide a submittal schedule for the individual segments within the next few weeks.
In preparation for the review of the thermal annealing report, future meetings will include a trip to Cooperheat to view a model of the heating assembly currently scheduled for August 21 and a trip to the Marble Hill site, one of the annealing demonstration project sites, currently scheduled for August 30 and 31.
'* If there are any questions regarding this meeting summary, please contact me at (301} 415-3024.
Enclosures:
As stated (2}
cc w/encls:
See next page DISTRIBUTION w/encls. (2):
Docket File (50-255}
PUBLIC PD3-l R/F WRussell/FMiraglia RZimmerman JRoe EAdensam (E-mail}
AThadani GHolahan BS heron ACRS (4}
OPA OGC EJordan BMcCabe KWichman EHackett BElliot JFair Llois CHinson JTatum EConnell JMa I Ahmed MMayfield GMillman ATaboada MVassil aros MZobler WAx.e l son, RI I I WKropp, Riii MHo l mberg, RII I ORIGINAL SIGNED BY Marsha Gamberoni, Project Manager Project Directorate 111-1 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation DOCUMENT NAME:
G:\\WPDOCS\\PALISADE\\ANNEAL\\PAL0724.SUM To receive a copy of thla document, Indicate In the l;>ox: "C" = Copy without attachment/enclosure "N" = No copy OFFICE PM:PD31 BC:EMCB NAME JStrosni DATE
Consumers Power Company cc:
Mr. Thomas J. Palmisano Plant General Manager Palisades Plant 27780 Blue Star Memorial Highway Covert, Michigan 49043 Mr. Robert A. Fenech Vice President, Nuclear Operations Palisades Plant 27780 Blue Star Memorial Highway Covert, Michigan 49043 M. I. Miller, Esquire Sidley & Austin 54th Floor One First National Plaza Chicago, Illinois 60603 Mr. Thomas A. McNish Vice President & Secretary Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Judd L~ Bacon, Esquire Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Regional Administrator, Region III U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, Illinois 60532-4351 Jerry Sarno Township Supervisor Covert Township 36197 M-140 Highway Covert, Michigan
- 49043 Office of the Governor Room I - Capitol Building -
Lansing, Michigan 48913 U.S. Nuclear Regulatory Commission Resident Inspector's Office Palisades Plant 27782 Blue Star Memorial Highway Covert, Michigan 49043 Palisades Plant Nuclear Facilities and Environmental Monitoring Section Office Division of Radiological Health Department of Public Health 3423 N. Logan Street P. 0. Box 30195 Lansing, Michigan 48909 Gerald Charnoff, Esquire Shaw, Pittman, Potts and Trowbridge 2300 N Street, N. W.
Washington DC 20037 Michigan Department of Attorney General Special Litigation Division 630 Law Building P.O. Box 30212 Lansing, Michigan 48909 Mr. Dennis Harrison U.S. Department of Energy NE 451 Washington, DC 20585 Mr. Kurt M. Haas Plant Safety and Licensing Director 27780 Blue Star Memorial Highway Covert, Michigan 49043 August 1995
MEETING ATTENDEES JULY 24, 1995 Marsha Gamberoni Cynthia Carpenter Barry Elliot Marian Zobler Dean Houston Al Taboada Mel Holmberg Tae Kim Ed Hackett Jack Strosnider Keith Wichman Mike Vassilaros John Fair John Ma James Tatum Edward Connell John Hannon Dick Smedley Jack Hanson R. W. Philips, Jr.
George Goralski Bill Beckius James Wong R. B. Jenkins David Howe 11 J. Bowen Rick Rishel Lee Tunon-Sanjur John Warren John Houstrup Joseph Burger Robert Nugent Glenn Campbell John Badrock Bill Server Eric Blocher Lloyd Zerr AFFILIATION NRC NRC NRC NRC NRC NRC NRC NRC NRC NRC NRC NRC NRC NRC NRC NRC NRC CPCo CPCo CPCo CPCo CPCo CPCo CPCo Westinghouse Westinghouse Westinghouse Westinghouse DOE ABB-CE ABB-CE Cooperheat Cooperheat Cooperheat ATI Consulting Gilbert/Commonwealth STS/EPRI ENCLOSURE I
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Palisades Thermal Annealing Topics for 7/31/95 Meeting with NRC
- 1.
Previous Annealing Studies
- 2.
Palisades Reactor Vessel
- 3.
Palisades RV Surrounding Structures
- 4.
Thermal/Stress Finite Element Model(s)
Mesh Size Cladding Thermal Boundary Conditions Insulation Thermal Loading Plasticity/Creep Residual Deformation/Stresses Vessel Supports Accounting for reactor coolant piping and equipment
- 5.
Upper Bound Analysis to establish limiting HU and CD rates
- 6.
Stress Criteria
- 7.
TAR Contents
PALISADES REACTOR VESSEL ANNEALING THERMAL/STRESS ANALYSIS PLAN July 24, 1995 Prepared by:
Lee Tunon-Sanjur, Senior Engineer Structural Mechanics Technology e
Westinghouse Electric Corporation Nuclear Technology Division
PALISADES REACTOR VESSEL ANNEAL Previous Annealing Studies Palisades Reactor Vessel Palisades RV Surrounding Structures Thermal/Stress Finite Element Model(s)
Upper Bound Analysis to establish limiting HU and CD rates Stress Criteria TAR Contents
PREVIOUS ANNEALING STUDIES e NUREG 20 studies Westinghouse KAPL Efforts EPRI Generic RV Studies Recent Efforts
RECENT EFFORTS
- 20 parametric studies
- 30 elastoplastic analysis
- Three-Loop plant with circumferential weld
- Four-Loop plant with longitudinal weld
2D MODEL RESULTS
- Heat zone can extend only slightly above beltline region.
- Without forced cooling:
- Vessel close to ambient after 20 days.
- Some heat will remain in the system after 30 days.
- Creep and residual stresses are small.
- Dimensional stability maintained.
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Temperature Profiles at Selected Pipe Locations
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RESIDUAL DEFORMATIONS DISPLACEMENT TIME*720
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~SSEL SUPPORT (6) REQUIRED Layout of Three-Loop Reactor Vessel
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Layout of Four-Loop Reactor Vessel
Four-Loop Thermal Stress Analysis Finite Element Model, outside view
Blowup showing how the hiqh cladding stresses are lor.ril in nature.
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I 3D MODEL RESULTS
- Temperature differences across nozzle:
- 270 °F for circumferential weld case
- 450 °F for longitudinal weld case
- Temperatures at the nozzle safe ends, piping, and internal supports are less than 600 °F.
- The elastoplastic analysis shows that the plastic deformation of the cladding is very localized and does not affect the base metal.
- The concrete temperature at the inner wall and vessel supports is well below 150 °F.
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WHAT IS AN ADEQUATE STRESS CRITERIA?
- ASME Section 111 design requirements met for all service conditions ocurring during anneal.
- Elevated temperature operation considered using ASME Section Ill Code Case N-47.
- Creep considered using ASME Section Ill Code Case N-47.
- Determine if a fatigue evaluation is necessary.
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30 MODEL RESULTS Table 3-7 Inelastic Strains and Allowables Item
£plastic E*llow*ble Strains averaged through the 0.03°/o 1°/o thickness Strains at the surface 0.08°/o 2°/o Local strains at any point
.34°/o 5°/o (in lhe clad)
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PALISADES REACTOR VESSEL Material Characterization:
Base Metal is SA 302 B Cladding :is 304 Stainless Steel Flow Skirt is lnconel SB 168 Insulation outside of the vessel is TRANSCO insulation All Physical and Material Properties are temperature dependent.
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30" ID Inlet Nozzle CROM Nozzle Instrumentation
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PALISADES SURROUNDING STRUCTURES
- Primary Loop Piping
- Reactor Vessel Supports
- SG/RCP Supports
- Biological Shield Wall
- Flange Head Clearance
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THERMAL STRESS ANALYSIS Mesh Size Studies 20 Radiation Model Thermal Loading Overall 30 Model Loading/Boundary Conditions Adiabatic Boundaries
- Thermal Loading
- Insulation
- - Nozzle Barriers Conduction Heat Losses Local Models
- Inlet Nozzle
- Outlet Nozzle
- Flow Skirt
- Reactor Vessel Supports Accounting for Reactor Coolant Piping and Equipment Plasticity/Creep Issues Residual Deformations/Stresses
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THERMAL BOUNDARY CONDITIONS 4 ELEMENTS THROUGH THE THICKNESS MODEL 120 NODE ISOPARAMETRIC SOLID ELEMENTS)
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HEAT UP THERMAL TRANSIENT 2
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TIME (HOURS) 5 6
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4 ELEMENTS THROUGH THE THICKNESS MODEL (20 NODE ISOPARAMETRIC SOL !O 8.EMENTSI 43.94 In. RADIUS 9 NODES THROUGH THICKNESS
8 ELEMENTS THROUGH THE THICKNESS MODEL 120 NOOE ISOPARAMETRIC SOLID ELEMENTS)
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43.94 in. RADIUS 17 NODES THROUGH THICKNESS
MESH CONVERGENCY CHECK USING 4 ELEMENT MODEL AND 8 ELEMENT MODEL AT END OF HEATUP(THERM. STRESS) 40,000
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8 THROUGH WALL DISTANCE (INCHES)
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MESH CONVERGENCY CHECK USING 4 ELEMENT MODEL AND 8 ELEMENT MODEL AT END OF HEATUP(THERM. STRESS) 20,000 en 10,000 a.. -
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Elastic stress clad(1 elem) 0 elastic-plastic clad(~ elem) 150 Distance through wall (mm) 200
PALISADES INLET NOZZLE LOCAL MODEL
1, PALISADES INLET NOZZLE LOCAL MODEL
PALISADES OUTLET NOZZLE LOCAL MODEL
THERMAUSTRESS ANALYSIS 30 Heat Transfer The heat-up rate shall be 25 °F/hr.
A hold period of 850 °F +50 °F/-O °F for 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />, and a cooldown temperature transient shall be specified.
The boundary conditions for the outside of the Palisades vessel shall be accounted for using effective film coefficients.
If the measured heat loss through the insulation is high/non-uniform, the thermal model would use different h values for each insulation panel.
Thermal analyses which will be performed include the following:
Normal Operation -
Using the expected thermal gradients established by the ANSYS radiation model and the characteristics of the heater.
Abnormal Operation -
The extreme thermal gradients shall be established based on postulated heater malfunctions.
THERMAUSTRESS ANALYSIS Stress Analysis A computer program simulating the relationship between these computed stresses, strains, and deflections and the field measured axial and circumferential thermal gradients at the inside of the reactor vessel wall shall be created.
The stress analysis of the vessel shall first be performed using the overall 30 model.
The 30 stress analysis shall use the same finite element models as the thermal analysis. The vessel cladding elements will be included in the stress analysis.
The overall model will include elastoplastic and creep effects if judged to be significant to assure that the Palisades reactor vessel internals fit-up required clearances will still exist after annealing.
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COMPARISON OF INSTRUMENTATION LOCATIONS - PALISADES AND MARBLE lllLL (PRELil\\IlNARY)
Zone Description Surface Instrument Palisades Marble Hill Bouom of Reactor Vessel OD of RV Thermocouple I
I Displacement I
I Strain Gauge 0
I Coincident with Core Suppon OD of RV Thermocouple 3
7 Lugs Vessel Beltline. Mid-Region OD of RV Thermocouple 3
7 Displacement 3
4 Vessel Top of Beltline into OD of RV Thermocouple 3
7 Transition RV Nozzle. Bouom Side OD of RV Thermocouple 0
8 Strain Gauge -
0 8
RV Nozzle. Upper Side OD of RV Thermocouple 0
7 RV Between Nozzle and Flange OD of RV Thermocouple 0
7 RV Flange OD of RV Thermocouple 0
7 Nozzle to Pipe Weld. Lower OD of RV Thermocouple 0
8 Side Strain Gauge -
0 8
Nozzle to Pipe Weld. Upper OD of RV Thermocouple 0
8 Side Strain Gauge-0 8
Nozzle Support On support Thermocouple 2
4 Reactor Cavity Liner. Beltline On liner Thermocouple 2
4 Region Each Loop Located Midway On piping Displacement 6
8 Between S/G or Pump & Vessel
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Outside of Bioshield Wall Nozzle to Pipe Weld. Bottom ID of RV Thermocouple 6
8 Side Nozzle to Pipe Weld. Upper ID of RV Thermocouple 6
8 Side Vessel Beltline. Heating Zone ID of RV Thermocouple 42 42 RV Shell Above RV NozzJes ID of RV Thermocouple 12 12 TOT AL LOCATIONS OD of RV Thermocouple 14 75 Displacement 10 13 Strain Gauge 0
25 TOT AL LOCATIONS ID of RV Thermocouple 66 70
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CTI N 127.08 G095JUL20 DUCT WORK PLUS HEAT EXCHANGER COOPERHEAT
~ Hot air only inside the ductwork, no combustible gases 4t Hot air is under pressure 2-3 p.s.i.g.
e Ductwork is insulated externally to a safe temperature e System will have tests, prior to installing into the reactor I
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er 1-u N121.09 G095JUL20 COOPER HEAT PROPANE PLUS WASTE GASES I) The propane supplied *from a road tanker via a docking station, not stationary tanks
- The tanker 50' plus from the containment
- Electrically heated units vaporize the gas. not di re ct fired
~ Gas pressure reduced to 15-20 p.s. Lg. after the vaporizers and 3-5 p.s.i.g. at the control train e Exhaust gases from the heat exchanger will be
- vented above the spent fuel building roof
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