I\ • :... consumers Power l'OWERIN& llllCHl&AN"S l'RO&RESS Palisades Nuclear Plant: 27780 Blue Star Memorial Highway, Covert, Ml 49043 March 27, 1996 US Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR-20 - PALISADES PLANT PRELIMINARY THERMAL ANNEALING REPORT, THERMAL ANNEALING . OPERATING PLAN, SECTION 1.7, THERMAL AND STRESS ANALYSES, AND SECTION 1.10, SUMMARY OF THE THERMAL ANNEALING OPERATING PLAN , At a meeting on June 6, 1995, we discussed with the staff our plan to anneal the Palisades reactor vessel (RV) during the refueling outage currently scheduled for the middle of 1998. In support of this effort, we presently plan to submit the final Thermal Annealing Report (TAR) in November of 1996 after the results of the Marble Hill reactor vessel annealing demonstration have been evaluated. The TAR . will include the information recommended in Regulatory Guide 1 . 162, Format and Content of Report For Thermal Annealing of Reactor Pressure To permit NRC .review of the TAR to begin before the Marble Hill results are known, we have made a series of submittals of preliminary TAB sections as they have been developed. This letter provides the eighth of those submittals. Attachment 1 to this letter contains the Thermal Annealing Operating Plan Section 1. 7, Thermal and Stress Analyses. Attachment 2 to this letter contains the Thermal Annealing Operating Plan Section 1.10, Summary of the Thermal Annealing Operating Plan. These sections are presented in the format recommended by Section C.1 of the Draft Regulatory Guide DG-1027. 9604020373 960327 . PDR ADOCK 05000255 \ p PDR I. OZ(Jo41 1 A CMS' ENER5Y COMPANY 1\()\) \ I \
Transcript
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consumers Power l'OWERIN&
llllCHl&AN"S l'RO&RESS Palisades Nuclear Plant: 27780 Blue Star Memorial Highway, Covert, Ml 49043
March 27, 1996
US Nuclear Regulatory Commission Document Control Desk Washington, DC 20555
. OPERATING PLAN, SECTION 1.7, THERMAL AND STRESS ANALYSES, AND SECTION 1.10, SUMMARY OF THE THERMAL ANNEALING OPERATING PLAN ,
At a meeting on June 6, 1995, we discussed with the staff our plan to anneal the Palisades reactor vessel (RV) during the refueling outage currently scheduled for the middle of 1998. In support of this effort, we presently plan to submit the final Thermal Annealing Report (TAR) in November of 1996 after the results of the Marble Hill reactor vessel annealing demonstration have been evaluated. The TAR
. will include the information recommended in Regulatory Guide 1 . 162, Format and Content of Report For Thermal Annealing of Reactor Pressure Vessels~ To permit NRC .review of the TAR to begin before the Marble Hill results are known, we have made a series of submittals of preliminary TAB sections as they have been developed. This letter provides the eighth of those submittals.
Attachment 1 to this letter contains the Thermal Annealing Operating Plan Section 1. 7, Thermal and Stress Analyses. Attachment 2 to this letter contains the Thermal Annealing Operating Plan Section 1.10, Summary of the Thermal Annealing Operating Plan. These sections are presented in the format recommended by Section C.1 of the Draft Regulatory Guide DG-1027.
9604020373 960327 . PDR ADOCK 05000255 \ p PDR I.
OZ(Jo41 1
A CMS' ENER5Y COMPANY 1\()\) \ I \
... ' . ·•
During the technical review of Section 1 . 7, several other sections were affected and have been revised. Attachment 3 provides a description of those changes. Attachment 4 contains a listing of the revised pages and provides the revised pages. Attachment 5 provides a Table of Contents of the entire Thermal Annealing Report to assist in your reviews.
The remaining two items of the Thermal Annealing Report; the Appendices to Section 1 . 7 containing the two dimensional and three dimensional thermal and stress analyses, are expected to be submitted by April 30, 1996.
SUMMARY OF COMMITMENTS
This letter contains ilo new commitments and no revisions to existing commitments.
This section summarizes the thermal and stress analyses performed to demonstrate that the operability of the Palisades reactor will not be detrimentally affected by the annealing operation. The annealing parameters described in Section 1.4 are considered in these analyses. In turn the results of these analyses are used to establish administrative .limits on key parameters (such as heatup and cooldown rates, and axial, azimuthal, and through-wall temperature gradients) also described in Section 1 .4. Additionally the results confirm the conclusions established in Sections 1.2 and 1.3 on the affect,of the annealing operation on other equipment, components, and structures.
The thermal and stress analyses summarized in this section and documented in Appendices 1. 7 .A and 1. 7.B for selected cases build upon an~lyses previously performed (Mager and Rishel, 1982; Vroom et al., 1984; Schwall et al., 1984). In each of these analyses -it was concluded that an appropriately applied annealing operation does not pose a reactor vessel integrity problem for post-annealing reactor operation.
1. 7 .B Analysis Approach
The detailed Palisades-specific thermal and structural analyses performed to demonstrate the acceptability of an in-situ thermal anneal were conducted using a combination of 2-D and 3-D inelastic finite element analyses as well a~ other calculation methodologies, where applicable.
The 2-D analyses were used to identify a heating distribution that met the annealing temperature requirements defined in Section 1.4, while still -insuring the structural integrity bf the reactor vessel and nearby equipment during the annealing operation. This heating distribution was used in the 3-D analyses. The 2-D analyses were also used to perform parametric studies of heat exchanger and RV emissivity, heat exchanger design configurations, RV wall thickness, heatup rates, material properties of the cladding, and equipment boundary conditions. Additionally it was used to determine the maximum residual stresses, and the maximum residual deformations; and to determine the effects of creep. This
_allowed b_etter_ utilization of the mor~ complex 3-D model_. _
The 3-D model analyses provide the best quantitative picture of the thermal loadings, th.ermal stress gradients, and displacements particularly_ in the areas at
- and-above the-RV nozzles. The 3-D-analyses were used to establish a boundingthermal and stress analysis, and to evaluate the effects of changes in the steam
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generator and vessel support boundary conditions, of hotter annealing and vessel temperatures, of changes in the heat exchanger temperature distribution, and of the steam generator support friction force during cooldown. Administrative limits on temperature gradients were also established with the 3-D analyses.
Other calculation methodologies were used to evaluate the annealing effects on simple geometry regions and localized areas as well as determine the effects of convection in the region above the heat exchanger top guard zone.
1. 7 .B.1 Two-Dimensional (2-D) Model
The 2-D finite element thermal model shown in Figure 1. 7 .B-1 was used to calculate temperature distributions. This model is based on the dimensions of the Palisades reactor vessel provided in Section 1.2 and includes the RV bottom head, the flow skirt and core support lugs, the RV shell plates, and the RV flange. This model also factors in the heat exchanger, and the reactor vessel top cover (RVTC) design described in Section 1.6 and the RV nozzle support structure and PCS piping. The model is axisymmetric and is supported at the RV nozzles in such a way as to allow radial expansion and contraction of the reactor vessel. In addition, the effects of the six RV nozzles, PCS piping, and three RV supports are simulated to give the same effect for the axisymmetric model as for the entire reactor vessel. The effect of the stud hole·s in the RV flange region have been approximated. The surveillance holder assemblies and core stabilizing lugs will only produce localized effects that would not affect the overall behavior of the reactor vessel, thus they are not included in the model. They are evaluated in the separate calculations described in Section 1. 7.B.3. The RV insulation, PCS piping insulation, RVTC insulation, RV flange seal ledge, and RV support structure are accounted for using effective film coefficients on the outside surfaces. The vessel cladding (approximately 0.25-inch thick) is assumed to be all stainless steel ignoring the relatively small areas of the Palisades vessel clad with Ni-Cr-Fe material. The justification for this assumption is the closeness in the coefficient of thermal expansion and the relatively thin layer of material with respect to the base metal.
This thermal model consists of a mesh of rectangular elements with six elements through the thickness of the RV wall and approximately eighty elements in height. This mesh configuration was selected to efficiently provide acceptable engineering accuracy without being excessively large. In order to model the thermal profiles through the thickness, the six rows of elements are distributed such that the inner elements are smaller than the outer rows. This distribution was chosen because the highest stresses and the greatest changes in stress occur at the inner surface, where the temperature loadings-are applied and where a finer. mesh will .provide a. more accurate calculation. The thermal model consists of about 1440 nodes and
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560 elements. The RV base metal and cladding are represented using high order 8-noded isoparametric elements.
For this model, heat is dissipated away from the reactor vessel to the heat sinks (steam generators, primary coolant pumps, biological shield wall, and ambient air) by all modes of heat transfer; conduction, convection, and radiation. Heat travels from the heat exchanger to the reactor vessel primarily by radiation. Figure 1. 7 .B-2 shows the radiating surfaces of the thermal model. Convection within the reactor vessel particularly in the- air space above the heat exchanger was evaluated in a separate calculation as having a negligible effect on the temperature profile in the RV upper shell and the RV flange region in comparison with the other modes of heat transfer, even considering an imperfect convection barrier seal at the heat exchanger top guard zone. Thus convection within the RV interior was not included in the 2-D- thermal model. This convection analysis is described in Sections 1. 7 .B.3 and 1. 7.D.3. Such convection effects would make the RV thermal gradients in the upper part of the reactor vessel smaller which would reduce the stresses. The temperature solutions were obtained with the ANSYS Version 5.1 finite element computer program on a workstation environment.
The structural model used the same basic axisymmetric model as shown in Figure 1. 7.B-1 however, with fewer nodes and elements since some of the elements were deactivated to allow for the structural analysis. This model had the RV support on the PCS piping centerline which will result in larger downward displacements of the RV bottom head and smaller upward displacements of the RV flange during the annealing operation than if the RV support was placed below the RV nozzle as it actually is. This was an inelastic analysis using the ANSYS Version 5.1 finite element computer program on a workstation environment.
More details on the 2-D model, particularly the assumptions, material properties, boundary conditions; and film coefficients, are provided in Appendix 1. 7 .A.
1. 7 .B.2 Three-Dimensional (3-D) Model
The 3-D finite element model shown in Figures 1. 7.B-3 and 1. 7.B-4 was used to calculate temperature and stress distributions. This quarter model is based on the dimensions of the Palisades reactor vessel provided in Section 1.2 and includes the RV bottom head, the flow skirt and core support lugs, the-RV shell plates, the RV flange, a RV inlet nozzle without a support pad, one:half of a RV outlet nozzle with a support pad, the vessel support structure, and the PCS piping. This model also factors in the heat exchanger and the RVTC design described in Section 1.6. This quarter section is considered to give more conservative results than the other section-because the PCS piping from the RV inlet nozzle to the primary coolant pump has the most severe angle and only the RV outlet nozzle is supported. The
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use of this quarter model of the reactor vessel was justified by developing a second-quarter--model with the inlet piping having. a less severe bend and comparing the resultant stress intensities at a given point in time at the end of heatup. Additionally the inlet piping was modeled in each of the four ·quadrants and displacements at the end of heatup were applied. Both studies determined that the loading is uniform.
The effect of the stud holes in the RV flange region have been accounted for in this model. The surveillance holder assemblies and core stabilizing lugs will only produce localized effects that would not affect the overall behavior of the reactor vessel, thus they are not included i~ the model. They are evaluated in the separate calculations described in Section 1. 7 .B.3. The RV insulation, PCS piping insulation, RVTC, and RV flange seal ledge are accounted for using effective film coefficients on the outside surfaces. The vessel cladding (approximately 0.25-inch thick) is assumed to be all stainless steel ignoring the relatively small areas of- the Palisades vessel clad with Ni-Cr-Fe material. The justification for this assumption is the closeness in the coefficient of thermal expansion of the Ni-Cr-Fe to the base material and the relatively thin layer of cladding with respect to the base metal.
This 3-D model consists of a mesh of high order 20-noded isoparametric brick elements with four elements through the thickness of the RV wall and approximately thirty elements in height. This mesh configuration was selected to efficiently provide acceptable engineering accuracy without being excessively large. In order to model the thermal and stress profiles through the thickness, the four rows of elements are distributed such that the inner el~ments are. smaller than the outer rows. As with the 2-D model, this distribution was chosen because the
·highest stresses and the greatest changes in stress generally occur at the inner surface, where the temperature loadings are applied and where a finer mesh will, provide a more accurate calculation. The thermal model consists of about 15,874 nodes and 3,068 elements.
For this model, heat is dissipated away from the reactor vessel to the heat sinks (steam generators, primary coolant pumps, biological shield wall, and ambient air) by all modes of heat transfer; conduction, convection, and radiation. Heat travels from the heat exchanger to the reactor vessel primarily by radiation . .Figure 1. 7 .B-2 shows the radiating surfaces of the thermal model. Differences between the 2-D and 3-D models are more pronounced in the RV nozzle and RV flange areas because the heat losses are more representative in the 3-D model. C_onvection within the reactor vessel particularly in the air space above the heat exchanger was evaluated in a separate calculation as having a negligible effect on the temperature profile in the upper shell and RV flange in comparison with the other modes of heat transfer, even considering an imperfect convec~ion barrier seal at the heat exchanger top guard zone. Thus convection within the RV interior was
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not included in the model. This convection analysis is described in Section 1. 7 .8.3 and 1. 7 .D.3. Such -convection effects would make the RV thermal gradients in the upper part of the reactor vessel smaller which would reduce the stresses. The temperature solutions were obtained with the ANSYS Version 5.1 finite element computer program on a CRAY C90 supercomputer environment.
The structural model used the same basic model as shown in Figures 1. 7 .8-3 c;ind 1. 7 .8-4 however, with fewer nodes and elements since some the elements were not needed for the structural analysis. Since the RV nozzle supports were applied at a location more representative of the Palisades plant configuration the vertical displacements determined by this analysis will be more realistic, however, a rise in the RV support structure due to thermal expansion needs to be accounted for in the 3-D results. This was an inelastic analysis performed with the ANSYS Version 5.1 finite element computer program on a CRAY C90 supercomputer environment.
More details on the 3-D model, particularly the assumptions, material properties, boundary conditions, and film coefficients, are provided in Appendix 1. 7 .B.
1. 7 .B.3 Additional Calculations
Additional calculations were performed either with small subsets of the models described above, simple computer mod.els using the ANSYS Version 5.1 computer program, or using simple hand-calculation analysis methods which factor in the results of the larger models to determine the effects of the annealing operation at simple geometry regions and localized areas. Such areas include the surveillance holder assemblies, the core stapilizing lugs, the flow skirt and the core support lugs, the RV seal ledge, and the lateral supports of the primary coolant pumps. Convection effects in· the upper vessel interior were also analyzed.
Surveillance Capsule Holder Assemblies
Calculations were performed on the surveillance capsule holder· assemblies (Figure 1.2.D-1, pages 8 and 9 of 14) to evaluate their structural integrity since they were excluded from the model, as discussed in Section 1.7.B.1. The evaluation included the transient thermal stresses in the full penetration weld between the bracket and the RV wall, the fillet welds joining the bracket to the surveillance capsule holder box beam, and the surveillance capsule holder box beam itself. The ANSYS model used for the evaluation included the tube box and nine pairs of brackets attached to the RV wall.
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Core Stabilizing Lugs
The core stabilizing lugs were evaluated to determine their structural integrity at the full penetration weld to the RV wall and at the core stabilizing lug shim bolts. They were also analyzed to determine whether they would come in contact with the heat exchanger during the annealing operation as a result of thermal expansion.
Azimuthal Temperature Variation on the RV Flow Skirt
A subset of the 3-D model (from the lower portion of the RV beltline downward) was used to quantify the effects of an azimuthal temperature variation around the RV flow skirt area and its impact on the structural integrity of the flow ·skirt. This analysis was accomplished by forcing a 50°F azimuthal temperature difference on the RV inner diameter in the model.
RV Seal Ledge
A small 2-D ANSYS model was developed of the RV flange and seal ledge to analyze whether the imposed stresses would be acceptable. This ledge was not included in the 3-D model since it was not considered to have a large effect on the overall model. A temperature of 900°F was applied ·below the RV flange (a worst case temperature) and a temperature of 100°F, representin_g ambient air, was applied to the _outer uninsulated surfaces of the seal ledge as well as the outer insulated surfaces of the RV flange. Different film coefficients were applied to the outer uninsulated surfaces of the seal ledge, the upper insulated surfaces of the RV flarige, and p'art of the lower surface of .the seal ledge adjacent to the flange. The inside RV flange surface was assumed to be perfectly insulated. A thermal analysis was perfo~med to establish the temperature distribution for use in a stress evaluation.
Lateral Supports of the Primary Coolant Pumps
The 2-D and 3-D model analyses assumed that the primary coolant pumps would not come in contact with their lateral supports. Thus an ANSYS 3-D model of the PCS inlet piping. was developed which assumed a horizontal constraint at an angle of 33 degrees from the pump centerline through the reactor vessel centerline to
. simulate a .contact with the pump support. Additionally a hard bound~HY condition was applied at the pump location by constraining the piping at the pump interface in all rotations and in the vertical direction. This model intended to evaluate the structural impact of this condition .
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Convection Effects in the Upper RV Interior
Calculations were performed to evaluate the effects of convection on the temperature profile in the RV upper shell and the RV flange regions above the heat exchanger top guard zone. A 2-D axisymmetric model of the region between the heat exchanger top guard zone and the RVTC was developed. The analysis considered the effects of convection with no heat leakage from the heat exchanger top guard zone convection barrier, a nominal leakage based on an imperfect seal, and a leakage ten times the nominal.
Biological Shield Wall Concrete and RV Insulation
The analysis and effect of the temperature distribution on the biological shield wall concrete as well as the RV insulation temperatures are addressed in Section 1.3.C.
1. 7 .C Analyses
The 2-D model was e_xercised to evaluate the effects of nine postulated annealing conditions, including nine stress analysis cases which used seven unique thermal input cases. These cases are described in Table 1. 7.C-1. The intent of these analyses was to evaluate-how variations in the heat exchanger and RV emissivity, heat exchanger design configuration, material properties of the cladding, ·heatup rates, RV wall thickness, equipment boundary conditions, and heating_ distributions will affect the annealing process, and to establish a "bounding" thermal/stress case ("design case") for use in the 3-D model that is reasonably consistent with the current heat exchanger design configuration. The analysis cases cover a du.ration from 0 to 360 annealing operation hours from initial heatup, through the soak or hold period, and to cooldown to less than 200°F: Inelastic strain due to creep was also evaluated using the 2-D model. More detail on the design Case T1/S1 is provided in Appendix 1.7.A.
The full 3-D model was exercised to evaluate the effects of nine postulated annealing conditions, including eight stress analysis cases whiph used six unique thermal input cases. These cases are described in Table 1. 7 .C-2. The design Case T1 of the 2-D analysis as shown in Table 1.7.C-1 was used as a basis for these 3-D analyses with the exception that the 3-D analyses do not include creep. Creep, as discussed in Section 1. 7 .D.1, was shown to be small in the 2-D model analyses. -This is consistent with past analyses (Mager.and Rishel, 1982; Vroom et al., 1984). The intent of the 3-D analyses was to establish a bounding thermal and stress analysis, and to evaluate the effects of changes in the steam generator and vessel support boundary conditions, of hotter annealing and vessel temperatures, of changes in the heat exchanger temperature distribution, and of the steam generator support friction force during cooldown. Administrative limits
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on temperature gradients were also established with the 3-D analyses .. These . administrative -limits were derived. from thermal distributions. beyond what .is required or even possible to achieve for the annealing operation. The thermal analysis cases cover a duration from 0 to 360 annealing operation hours from initial heatup, through the soak or hold period, and to cooldown to less than 200°F. The stress analysis cases were done for times of 30 hours, 36 hours, and 204 hours. These points were chosen because they coincide with the maximum RV strains, the maximum nozzle stresses, and the maximum temperatures and deformations, respectively, as determined in the 2-D analyses. More detail on the Cases T1 . 1,. Thermal and T1 . 1 F, Stress are provided in Appendix 1 . 7. B.
For all 2-D and 3-D analysis cases, except where otherwise noted in Tables 1. 7 ."c-1 and 1. 7 .C-2, the heating parameters in the annealing zone given in Table 1. 7 .C-3 were applied. The average heatup rate up to 850°F is 25.8°F/hr followed by a slowdown in the heat up rate to the bounding annealing temperature of 900°F. After a hold period of 168 liours the annealing zone is cooled down using an average temperature decrease on the heat exchanger surfaces of 100°F/hr.
1. 7 .D Analysis Results
The evaluations of Sections 1.2 and 1.3 to determine the effects .of the annealing operation on internal and external attachments and on other equipment, components, and structures have referenced Section 1 . 7 for the quantification of temperatures, stresses, and displacements. Below is a list of these required values and the reference to the appropriate table or section in which this information is summarized and discussed.
Loadings at the nozzle extensions (Table 1. 7 .D-7) Thermal stresses in the core stabilizing lugs (Table 1.7.D-7, Section 1.7.D.3) Thermal stresses in the ·core support lugs (Table 1. 7 .D-7) Thermal stresses in the RV flo_w skirt (Table 1. 7 .D-7, Section 1. 7 .D.3) Thermal stresses in the surveillance capsule holders (Table 1. 7 .D-7, Section 1.7.D.3) PCS piping temperature (Table 1. 7 .D-6) Temperature at the steam generators (Table 1. 7 .D-6) Steam generator nozzle loads (Table 1. 7 .D-7) RV support pad temperature (Table 1. 7.D-6) RV .support temperature (Table 1. 7 .D-6) RV bottom head displacement (Table 1. 7.D-9) RV flange displacements, radial ·and vertical and impact to the 0-ring leakoff piping (Table 1. 7.D-9) Thermal profile down the PCS piping (Table 1. 7 .D-6) RV nozzle extension temperatures (Table 1. 7 .D-6)
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Radial displacement at steam generators (Table 1. 7.D-9) Radial displacement at-primary coolant pumps (Table 1. 7 .D-9) Differential thermal loading between the core stabilizing lugs and shim bolts (Table 1.7.D-7 and Section 1.7.D,.3)
1. 7 .D.1 2-D Analysis Results
Detailed time temperature history, temperature contour, time history radial and vertical displacement, stress contour, and strain plots are provided for Case T1/S1 in Appendix 1. 7 .A. A summary of the maximum temperatures (rounded within 5 degrees) at key node locations is provided in Table 1. 7 .D-,1 for all of the temperature cases. The radial and vertical displacements, peak stress intensity, peak strain, and the peak residual deformation for eight stress cases are provided in Table 1. 7 .D-2. Case 52 was not conducted because the thermal distribution (Case T2) was similar to Case T1 and thus would yield similar results to Case 51.
The effect of creep is shown in Table 1. 7 .D-3. The guidelines of ASME Code Case N-47-32 are provided for comparison sake only, as this Code Case applies to the more limiting condition of operational transients. The Palisades anneal will be governed by Code Case N-557 which limits stress and time at temperature to control creep .
Results of the 2-D analysis are useful as a qualitative comparison of effects among the various sensitivity cases and to demonstrate that Cc;ise T1/S1 as shown in Table 1.7 .C-1, is bounding for the 3-D analysis. They are also used to
. quantitatively establish the effect of creep, residual deformations, and residual stress at the completion of annealing.
The Case T1/S1 is considered bounding because the annealing temperatures (900°F) used are at the high end of the target for the actual annealing operation, and the ambient air temperatures (100°F) are lower than expected. This combination will. result in larger axial and through-wall thermal, gradients than will occur during the annealing and lead to conservative RV stress results. The acceptability of the stress is demonstrated by the resultant peak stress intensity for the Case S 1 anywhere in the model as shown in Table 1. 7 .D-2. This stress intensity is located in the RV cladding below the RV nozzle at 3Q hours. Lesser stress intensities are in the base metal.
Also in Case T1 it was noted that the maximum temperature differential between the reactor vessel and the flow skirt occurred during heatup and is 103°F. This differential is less than 60°F during the hold or soak period. These temperature differentials are large enough that the flow skirt is included in the 3-D analysis model in order to quantify such stresses .
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The temperature profile of the RV beltline (for Case T1) during heatup~ in the hold period, and during coeldown is shown in Figure 1. 7.D-1. This figure shows-the temperatures at the bottom (8), middle (M), and top (T) of the Palisades RV beltline both on the inner diameter (I) and on the outer diameter (o). The heatup rate was consistent with the heating parameters of Table 1. 7.C-3. The inside surface temperature in the RV beltline region was held very closely at 900°F particularly at the end of heatup. This is because the 2-D analysis allowed multiple iterations of the heat exchanger surface temperatures until a consistency could be achieved. This is an analytical equivalence to the variable control that an annealing equipment technician has. The analyses maintained a constant heat input throughout the full 168 hour hold period. This accounts for the temperature variations in the annealing zone during the hold period.
The equivalent cooldown rate of the RV beltline using the 100°F/hr reduction in the heat exchanger surface temperatures is less than 25°F/hr as shown in Figure
. · 1. 7 .D-1. At 500°F, the cool down of the RV beltline occurs at even a slower rate until it reaches about 1.25°F. This is because the forced cooling of the heat exchanger, as described in Section 1.5.D.1, loses its efficiency at lower temperatures. The through-wall temperature difference in the RV beltline region during the annealing operation is shown in Figure 1. 7.D-2 for Case T1. The" largest through-wall temperature gradient is approximately 100°F at the end of
· heatup. During the hold period it dr<;>ps to less than 50°F, and it reaches approximately 70°F near the beginning of cooldown.
Figure 1. 7. D-3, from Case T1/S1, shows the "coke bottle" effect as seen in previous studies. This is caused by .the axial thermal gradient across the RV nozzles. This gradient results in variable radial thermal displacements along the length of the RV cylindrical wall. The variable axial deformation results in the bending of the RV cylindrical shell. This bending causes higher stresses at the upper shell to intermediate sh~ll transition and to a lesser degree at the lower shell to bottom head transition. The most significant effect of this RV rotation is the production of a bending of the PCS piping particularly at the RV nozzle extension and at the RV nozzle reinforcement area transition. The "coke bottle" effect shown in Figure 1. 7 .D-3 is less pronounced than that observed in previous studies primarily due to the.wider heating zone (175% of the RV beltline length as shown in Figure 1. 7 .B-2). This is compared to 120% of the RV beltline length in previous studies (Server, W.L., 1985). However the bending is still presenL
There are very little differences between the Case T1/S1 resul_ts and the variations prescribed in the other thermal and stress cases described in Table 1 . 7. C-1 . The RV stresses due to the annealing operation will be predominantly dictated by the temperature distribution and not by variations in the equipment' emissivities (Cases T2/S2 and T5/S5), RV material thickness (Case T4/S4), steam generator boundary
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conditions (Case T1 /S6), and the reference temperature (Case T1 /S7). This conclusion is dearly shown ·by comparing the Case T3/S3 results with the Case T1 /S1 results. As expected, without insulation on the top guard zone cover, the RV nozzle and RV flange regions become hotter, as shown in Table 1.7.D-1. A corresponding reduction in the thermal gradient across the RV nozzle, however results in a significant reduction in stress, as shown in Table 1 . 7. D-2.
A faster heatup rate, as assumed in Case T6/S9, will also result in hotter temperatures outside the annealing zone since the heat exchanger must put in . more heat in a shorter period of time. Like Case S3, this results in a peak stress intensity less than the design case (Case S 1) due to the reduction in the thermal gradient across the RV nozzle.
Hotter temperatures throughout the reactor vessel as seen in Case T7/S 10 result in greater radial and vertical displacements than Case T1/S1 due to greater thermal expansion. However the stresses, strains, and displacements are similar.
The maximum r~sidual deformation for all the analyzed cases is 0.008 inch (radial) and this occurs··limmediately below the- RV nozzle area. A residual surface flatness of 0.0003 inch is calculated for the 0-ring surface.
Table 1. 7 .D-3 provides the results of an inelastic analysis performed to evaluate the RV strains. This was performed since the annealing is conducted at elevated temperatures above the creep threshold temperature of 700°F~ The largest total strains occur at the end of the heatup period and are small. The maximum strains occur just below the RV nozzles. The largest strains due to creep are also small. Thus the strains on the re_actor vessel res!Jlting from the annealing are negligible consistent with the basis for ASME Code Case N-557.
Since the residl:lal deformations and strains, including the effetts of creep, are insignificant, the residual deformations and creep need not be ·evaluated in the 3-D analysis. Since creep produces stress relaxation, the stresses obtained in the 3-D analysis will be higher than during the actual annealing operation and thus conservative. However. the effects of creep will impact the residual deformations thus the 2-D analysis will be· used to establish these values.
The residual stresses were calculated to be up to 16.1 ksi and are mainly due to bending of the- RV wall. It occurs on the inner diameter surface near. the upper shell to intermediate shell transition area .
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1. 7 .0.2 3-0 Analysis Results
Detailed time temperature history, temperature contour, time history radial and vertical displacement, stress contour, and strain plots for Thermal Case T1 .1 and Stress Case T1 .1 F are provided in Appendix 1. 7 .B. Relevant plots from the 3-D analyses are included as Figures 1. 7.D-4 through 1. 7.D-6 in this section. A summary of the maximum temperatures at key locations is provided in Table 1. 7.D-4 for three of the temperature cases. The radial and vertical displacements and maximum stress intensities at selected locations for severi stress cases are· provided in Table 1.7.D-5. Tables 1.7.D-6 through 1.7.D-9 provide a summary of the maximum temperatures, stress intensities, thermal gradients, and relevant displacements at key locations obtained mostly from the 3-D analyses. These latter tables include the highest values obtained in the analyses for conservatism.
Thermal Case T.1.1 of the 3-D analysis used the same characteristics as the Case T1 of the 2-D analysis including the target heating parameters· of Table 1. 7 .C-3. For this case the RV flange gets about 200°F hotter than in the 2-D case at the RV flange to shell interface at the end of the hold period (e.g. 750°F versus approximately 550°F). The RV nozzle extension reaches a temperature of about 540°F at the end of the hold period which is about 250°F hotter than ihe 2-D case. The RV nozzles and PCS piping heat losses are more realistic in the 3-D analysis than the 2-D thus these higher temperatures are expected. The Case T1 .1 temperatures at key locations are provided in Table 1. 7 .D-4. Temperature contour
· plots are shown for these 2-D and 3-D cases in Appendices 1. 7 .A and 1. 7 .B.
Stress Cases T1 .1, T1 .1 P, and T1 .1 F use the Thermal Case T1 ;1 temperature distributions. The differences in- these cases are the manner in which the nozzle support and the steam generator are constrained. Case T1 .1 and T1 .1 P results provide a comparison on the effects of allowing the RV nozzle support to rotate. By design, the nozzle support rotates on a convex shoe. Case T1 .1 P includes this feature. Table 1. 7. D-5 indicates that the changes in stress intensities for these cases in the RV nozzles are small (0.1 to 1.1 ksi). The stress intensity of 58.5 ksi occurring at a very localized area at the edge of the RV nozzle support pad in Case T1 .1 is eliminated by allowing the nozzle support to rotate (Case T1 .1 P) and thus the emphasis on stress intensities is in the RV outlet nozzle transition; Cases T1 . 1 P and T1 . 1 F provide a comparison on the effects of changing the steam generator nozzle from being able to rotate to being fixed, respectively. The magnitude of change is approximately 1.1 to 4.5 ksi in the RV nozzles with the Case T1 .1 F stress intensities being greater (Table 1. 7. D-5). The largest stress intensity is that in the RV outlet nozzle transition of 46.8 ksi. In none of these cases is there plasticity in the base metal.
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Table 1.7 .D-5 also indicates for these cases that the maximum stress intensities in the RV nozzles occur essentially at the end of the hold period (204 hours). This is inconsistent with the 2-D analysis results which show that the maximum stresses in the RV nozzles occur at the beginning of the hold period (36 hours). The primary cause of these nozzle stresses at the later time, however, is riot due to large axial thermal gradients, as it is at 36 hours, but rather due to a conservatively assumed thermal expansion of the reactor vessel supports, a feature not included in the 2-D analysis. The reactor vessel support structure rises approximately 0.23 - 0.25 inch. The effect of this thermal expansion has been calculated to be on the order of 20 ksi without having a stress intensification factor due to the geometrical discontinuity between the pipe and the nozzle reinforced area (nozzle transition area). Without this conservatism on the thermal expansion, the maximum nozzle stresses revert back to the beginning of the hold period consistent with the 2-D analysis.
The largest radial deformation in the reactor vessel in Cases T1 .1, T1 .1 P, and T1 .1 F occurs in the RV beltline area and is similar to that of Case T1/S1 of the 2-D analysis ( 0.69 inch as shown in Table 1. 7 .D-5 versus 0.64 inch as shown in Table 1. 7.D-2, respectively). Figure 1. 7 .D-4 provides a picture of the. displaced vessel at t = 204 hours. Whereas the "coke bottle" effect is observable it is again less pronounced due to the Wider heating zone which extends into the RV nozzle region. The largest vertical displacement is the downward movement of the RV bottom head of 2.17 inches at the end of the hold period (t = 204 hours). This value conservatively includes the 0.25 inch rise in the RV support structure. The RV flange moves radially 0.53 inch and vertically 0.92 inch also at the end of the hold period. The steam generator end and the primary coolant pump end move 1.01 inch and 0. 72 inch, respectively, also at the end of the hold period.
The Case T1 .2 was designed to show that even at a higher annealing zone temperature of approximately 950°F (this temperature is even beyond the ASME Code Case N-557 guidance of 940°F) there remained a margin in terms of structural integrity. Larger vertical displacements (t = 204 hours) were observed in the RV bottom head (2.23 inches) and in the RV flange (0.95 inch) compared to the Cases T1 .1, T1 .1 P, and T1 .1 F. These values conservatively include the 0.25 inch rise in the RV support structure. A small increase of the RV flange radial displacement (0.55 inch) and the RV beltline displacement (0. 70 inch) were also observed. Radial movement in the steam generator end and in the primary coolant pump end was observed to be~greater (0.94 inch versus 0.91 inch, and 0.69 inch versus 0.64 inch, respectively) than the similarly constrained Case T1 .1 F . These differences in displacements are due to the higher temperatures throughout the reactor vessel caused by the higher annealing temperature (Table 1. 7 :D-4). The maximum stress intensity was observed at a time of 36 hours and it reached 58.8 ksi. Again it occurred at the edge of the RV support block as a result of the nozzle
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support being not allowed to rotate. The stress intensity in the RV outlet nozzle transition is 47.8 ksi.
In the Case TC1 an imposed 50°F azimuthal temperature difference was put on the heat exchanger surface at the lower portion of the RV beltline during heatup because it represents the time of maximum circumferential variation. Such a difference only produced a maximum azimuthal temperature difference of 25°F to 35°F in the annealing zone as shown in the temperature contour plot at t = 40 hours of Figure 1. 7 .D-5. This indicates that the reactor vessel tends to act as a thermal averaging mass and thus can average out local hot spots ·that might occur in the heat exchanger. The reason for the lower temperatures above the annealing zone compared to Thermal Case T1. 1 is that the net thermal energy imposed on the reactor vessel is less.
Case T2 was an intended hotter case, in comparison with .Case T1 . 1 to establish larger bounding temperatures. In this case a higher ambient air temperature consistent with that used in the concrete temperature evaluation in Section 1.3, 33% better insulation properties, and no internal convection in the PCS piping were assumed. This resulted in higher temperatures in the RV flange ·(760°F at the vessel shell interface to 637°F at the RV flange top compared to 749°F and 623°F, respectively), and in the RV nozzle support pad (563°F compared to 548°F) for Case T1 . 1 . A temperature gradient across the RV wall in the beltline region of 20°F was observed (t = 204 hours). In comparison with Cases T1 .1, T1 .2 and TC 1, Case T2 resulted in only larger radial movements of the steam generator end (1.12 inches) and the primary coolant pump end (0.81 inch). No plasticity is occurring in the RV base metal.
Case T3 was designed to evaluate the effect of a forced axial thermal gradient across the ends of the annealing zone (top and bottom). An axial temperature difference of 400°F was imposed across the flow skirt and a 100°F axial temperature difference was imposed in the RV nozzle region. This resulted in a maximum stress intensity in the RV flow skirt area of 50.4 ksi and a stress intensity of 50.3 ksi in the RV outlet nozzle transition. The maximum.total strain, including plastic strain, in the flow skirt is small at 1.91 %. No plasticity is occurring in the RV base metal.
Case T4 imposed an extreme artificial temperature loading across the RV nozzle as shown in Figure 1. 7. D-6. This distribution was intended to establish an administrative limit for an axial temperature gradient across the RV nozzles. Large axial thermal gradients lead to bending of the PCS loop piping potentil;llly large enough to cause the RV nozzle ends to form a plastic hinge. The largest stresses due to a temperature gradient will happen at the end of heatup when the RV beltline region is reaching its prescribed temperature of approximately 900°F but
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the flange is still considerably cooler. With a temperature difference of 470°F, which is extl"emely-high, the maximum stress-intensity was determined to be 65.4 ksi. Again this is a very localized stress at the end of the RV support pad caused by the RV nozzle not being able to rotate. Elsewhere, the largest stress intensity in the RV nozzle occurs at the RV outlet nozzle transition (49.9 ksi). At this point localized plasticity occurs. The total strain is about 0.0026 in/in.
Case T6 was performed to determine that if the reactor vessel would ovalize during the anneal due to the steam generator support friction loads it would be elastic. This effect was considered in the 2-D analysis (Case S8) but the 2-D analysis could not account for the stresses caused by this ovalization due to nonaxisymmetric friction loads. Two cases were evaluated, one using the cross-over leg stiffness and the other with no cross-over leg stiffness. With a cross-over leg stiffness, the inlet piping was restricted from displacing horizontally and ovalization of the reactor vessel was not observed. It was determined in this case that the maximum radial displacement of the RV flange could be 0.079 inch and that the maximum stress intensity would be 4. 7 ksi. With no cross-over leg stiffness the
. reactor vessel does ovalize hO\~ever the resulting displacements and stresses are smaller (0.027 inch and 3.4 ksi, respectively). The actual c;:ase will probably· be in between these cases.
1. 7 .D.3 · Additional Calculation Results
Surveillance Capsule Holder Assemblies
Maximum temperature differences of 275°F to 312°F between the surveillance - capsule holder and the RV wall occur very early in the heating process (approximately 5 hours). The surveillance capsule holder bracket weld to the RV wall has a calculated stress intensity of 40.4 ksi using an elastic analysis. This stress intensi~y value will not be reached since th~ analysis of the box beam and the support brackets determined that the axial load would exceed the yield stress _ of the box beam. Consequently this box beam yielding will reduce the loads and stresses being .developed. An evaluation of the fillet welds that join the bracket and the box beam indicates that the stresses are acceptable. These results indicate that the surveillance capsule holder assembly will rernain intact during the annealing operation. The box beam, however, is expected to exceed the elastic limits and may be affected by creep which could· result in permanent deformation.
Core Stabilizing Lugs
The six core stabilizing lugs are directly opposite the lower guard zone of the heat exchanger and are closer to the heat exchanger than the RV wall. Thus the maximum temperature gradients occur early in the heatup when the temperature
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difference between the heat exchanger and the RV wall is at its maximum value. The average axial temperature of the lugs may vary from 106°F to 176°F higher than the RV wall while the temperature of the RV wall is varying from 200°F to 800°F, respectively. This temperature distribution causes an axial thermal gradient across the lugs. Since one end of the core stabilizing lug is unconstrained, this loading will not cause large stresses on the reactor vessel. When the reactor vessel inside diameter surface reaches 900°F, the end of the core stabilizing lug away from the reactor vessel may reach 1 025°F.
The critical items on the core stabilizing lugs are the core stabilizing shim bolts. The excess margin in these bolts can accommodate a differential temperature between the bolts and the lug of 84°F. The thermal expansion coefficients and thermal conductivity coefficients of the lugs, shims, and bolts are well matched. In addition the bolts are oriented circumferentially, i.e. in parallel to the radiant heat exchanger. Thermal gradients between the bolts and the lugs are anticipated to be between 10 to 20 degrees. Thus the bolts are not adversely affected by this annealing process.
It was also shown that the core stabilizing lug will have a small inward displacement toward the heat exchanger of 0.02 inch which, despite the expected growth in the heat exchanger, will not cause an impact between the end of the lug and the heat exchanger outer surface during the annealing operation.
Azimuthal Temperature Variation in the RV Flow Skirt Area
The forced azimuthal temperature variation of 50°F imposed in the RV flow skirt region-, including the flow skirt, resulted in a maximum stress intensity of 22.9 ksi. This stress intensity is located in the RV base metal but would bound that seen in the RV flow skirt. The maximum principal total strain, including the flow skirt, is 0.153% and it is elastic.
RV Seal Ledge
The temperature distribution in the RV seal ledge is axial varying approximately 130°F across the length of the seal ledge. This temperature distribution will not cause any significant bending. A maximum radial displacement of 0.61 inch was calculated. The maximum stress intensity calculated is 43.4 ksi and is located on the tip of the seal ledge. The stress intensity at the RV flange to seal ledge interface is approximately 24.5 ksi.
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Lateral Support of the Primary Coolant Pumps
If the primary coolant pumps were to contact one of their lateral supports during the anneal operation it would still be acceptable in terms of structural integrity of the PCS piping. The maximum stress calculated occurred at the nozzle to pipe boundary condition in the piping where the displacements were applied and is very localized. This maximum stress intensity value is 51.6 ksi and is artificial in that it occurs at the structural model boundary. Stresses away from the imposed boundary are less than 38 ksi.
Convection Effects in the Upper Reactor Vessel Interior
Figure 1 . 7. D-7 shows the calculated temperature profile along the RV wall inside surface above the top guard zone of the heat exchanger as a result of radiation and convection in this region. It shows that the effects of convection are minimal even considering a conservative leakage of heat from the RV beltline region through the top heat ·exchanger convection barrier. For this conservative case there is a 10% reduction in the temperature difference of the RV inside surface from the top guard zone to the top of the flange, e.g. the flange gets hotter. For the case, given a nominal gap of 0.25 inch between the top heat exchanger convection barrier and the RV wall, the temperature difference is only 3.3% or 10°F.
1 . 7. D .4 Evaluation of Results
The primary criteria for evaluation of the analyses results is defined in the ASME Code Case N-557. The 3Sm stress intensity limit in the Code Case is predicated upon elastic analysis and is intended to limit the thermal strains, exclusive of creep strains, to approximately twice the yield strain as a reasonable limit (the creep strains are limited by the time at temperature limits), but these analyses use inelastic calculational methods. Thus, the comparison of the computed stresses to 3Sm in areas where plasticity occurred would be unconservative in limiting strains. However, the results from the finite element analyses have been examined to identify whether plasticity has occurred in the reactor vessel base metal, and its impact on the calculated stresses. Review of the plastic strains concluded that the vessel base metal did not experience plasticity for most of the cases shown in Table 1. 7.D-5 and thus the 3Sm criterion can 'be directly applied to the resultant stresses from the finite element analysis. The only case that experienced plasticity is the extreme axial temperature gradient across the RV nozzles (Case T4). The Sm values from the Code Case N-557 for use in evaluating stress intensities in SA-533 · Grade B, Class 1 plate, SA-302 Grade B (Ni-modified) plate an'd SA-508 Class 2 and Class 3 forging material at the annealing temperatures are provided in the table below. Sm values for the piping material, SA-516 Grade 70, experiencing
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lesser temperatures are found in the ASME Code Section Ill Appendices and are also provided in the table below.
The ASME Code Case N-557 on thermal annealing establishes a limit of 3Sm for primary plus secondary stress intensity in the reactor vessel excluding cladding. Table 1. 7.D-7 provides the resultant maximum stress intensities for key locations specific to the Palisades reactor vessel and. PCS piping and a comparison with the 3Sm limit, where appropriate. The comparison is made using the Sm value at 900°F of 24.0 ksi as defined in the above table for the RV materials, and the Sm value at 650°F of 18.4 ksi also defined in the above table for the piping material. This is conservative in that the required Sm value increases for lower temperatures. In addition to the stress intensities found in any of the cases, the values of the stress intensities for the Case T1 .1 F, Stress are provided in Table 1. 7.D-7 for comparison.
The maximum stress intensity, which is not required to be evaluated by the ASME Code Case N-557, found anywhere in the reactor vessel and in the analyses performed was 58.8 ksi or 82% of the allowable (Table 1.7.D-7). It should be noted that this value is a result of an analysis assuming no rotation at the RV outlet nozzle support, an unrealistic boundary condition. For the Case T1 .1 F, Stress, in which the nozzle support is allowed to rotate, the maximum stress intensity anywhere in the reactor vessel is 46.8 ksi or 65% of the allowable. This maximum value occurs at the RV outlet nozzle transition on the outer diameter surface. All other areas in the reactor vessel have lesser values. The maximum stress intensity in the PCS piping is at the RV outlet nozzle extension (actually in the A-508, Class 1 material adjacent to the pipe). This maximum value is 42.6 ksi or 77% of the allowable (Table 1. 7 .D-7). For the Case T1 .1 F, Stress the
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maximum stress intensity in the piping is also at the same point in the RV outlet nozzle extension . This maximum value is 29.5 ksi or 53% of the allowable (Table 1.7.D-7).
For Case T4 in which the extreme 470°F axial temperature gradient is imposed across the RV nozzles and in which the RV nozzle transition area experiences localized plastic deformation, the total strain is about 0.0026 in/in. The limit goal of Code Case N-55 7 is to keep the primary plus secondary strain to less than twice the yield strain.
From ASME Code Appendix I:
yield stress = 43.8 ksi at 600°F E = 26.4 x 103 ksi at 600°F
therefore:
yield strain
2 x (yield strain)
= yield stress/E = 43.8/26.4 x 103
= 0.00166 in/in = 0.00332 in/in)
•. Thus comparing the total strain of 0.0026 in/in to twice the yield strain of 0.00332 in/in results in an acceptable condition. This case, however, is considered the axial temperature gradient limit across the RV nozzles during
·. heatup. A more realistic axial temperature gradient is on the order of 280°F -300°F as observed in Case T1 .1.
The maximum radial displacement of the reactor vessel was calculated as 0. 70 inch in the RV beltline region (Table 1. 7.D-5) and it occurs outward from the RV centerline. This displacement can be accommodated without interference with other equipment. The maximum radial displacement of the steam generator was determined to be 1.12 inches (Table 1. 7 .D-5) outward from the reactor vessel. This displacement is less than the normal operating condition displacement of 1.125 inches. This displacement is well within the travel limits of the steam generators (Steam Generator A - 1.99 inches, Steam Generator B - 1.41 inches).
The maximum vertical displacement in the reactor vessel relative to the RV support was determined to be 2.23 inches using the 3-D analysis and it occurs at the bottom head in the downward direction (Table 1. 7.D-5). This value is approximately 1. 25 inches more than the thermal growth experienced during the normal plant heatup. The worst case effect is that the block insulation below the reactor vessel will be crushed and the stainless steel floor covering will be
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deformed up to 1.25 inches. This condition will not affect the operability of the floor or~the cooling coils in the floor. This effect has not been modeled in the 3-D analyses since it was assumed that sufficient clearances exist. The 2-D analysis result for this vertical displacement (Table 1. 7.D-2) is higher, however the placement of the RV support in the 2-D model was put on the PCS piping centerline, a half pipe diameter higher.
The RV flange was determined to have a maximum radial displacement of 0.55 inch and a maximum vertical displacement of 0.95 inch which are outward from the RV centerline and iri the upward direction, respectively (Table 1. 7 .D-5). This displacement is more than normal operating conditions and is conservative based on the assumed thermal growth of the reactor vessel support structure of 0.23 -0.25 inch. Its largest impact is on the 0-ring leakage monitor tubes off the RV flange. A structural analysis using a piping analysis program has beeri performed to compare the maximum range of stress intensity in the monitor tubes to ASME Code Section Ill allowables. The maximum range of stress int~nsity in the Alloy 600 monitor tubes resulting from the thermal growth of the reactor vessel occurs at the RV flange mating surface where the tubes are rigidly attached. This range of stress intensity (58.4 ksi) is within the Code allowable (69.9 ksi for material SB-166 at 800°F). For the attached stainless steel leakoff system piping the maximum stress occurs at the point where the piping penetrates the biological shield wall. This maximum stress (39.5 ksi) is within the Code allowable (46.3 ksi for A376, Type 316 material at 150°F). Thus the additional thermal loading on the RV monitor tubes and attached leakoff piping is acceptable and can be accommodated by the design configuration.
The peak residual deformations due to axial thermal gradients are less than 0.008 inch (radial). and occur immediately below the RV nozzle area. This value is less than the 0.460 inch hot vertical clearance between the core support barrel (CSB) and the RV flow skirt, and the 0.067 inch hot radial clearance between the CSB and the RV outlet nozzle. The relative surface flatness of 0.0003 inch of the 0-ring surface assures that the 0.005 inch flatness of the 0-ring seating surface requirement is maintained .. The hot tangential tolerances on the core stabilizing lugs are 0.017 inch on each side. A permanent tangential deformation will occur if the azimuthal temperature variations result in stresses exceeding the yield limit. This is unlikely since the heating method results in relatively uniform azimuthal heating of the reactor vessel. Since the axial thermal gr.adients are much greater than the azimuthal gradients, azimuthal r:esidual deformations will be less than 0.008 inch and thus within the tangential tolerances. The maximum radial displacement of the RV flange due to the steam generator support frictional loads is 0.079 inch, and the maximum stress intensity is 4. 7 ksi. This stress value is very low and indicates that the reactor vessel will not go into the plastic range due to the steam generator sliding base friction. Thus by jacking the steam generator
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at the end of cooldown the reactor vessel will return to its original shape. This assures-that the reactor-vessel head seating counterbore tolerance of 0.045 inch on radius and the CSB seating counterbore tolerance of 0.469 inch on radius can be maintained.
The residual stress was calculated to be less than 16.1 ksi and it is mainly due to bending of the wall. It occurs on the inner diameter surface near the upper shell to intermediate transition area.
The largest peak strains occur at the end of the heatup period and are small (Table 1. 7 .D-2). The maximum strains occur just below the RV nozzles. The largest residual strains due to creep are also small (Table 1. 7.D-3). Thus the strains on the reactor vessel resulting from the annealing are negligible consistent with the basis for ASME Code Case N-557.
The RV flow skirt was determined to exceed the elastic limits as a result of the ann~aling process. The bounding axial temperature gradient case (Case T3) resulted in a maximum localized total strain of less than 1.91 % which is small and therefore acceptable. The bounding azimuthal temperature gradient case resulted in a maximum localized total strain of less than 0.153% which was all elastic. Thus even if the temperature gradients were combined the localized total strain would be less. than 2.06%, which is small and also acceptablEl .
The surveillance capsule holder assembly box beams have been determined to undergo inelastic deformation. This deformation might make it difficult to put the surveillance capsules back in place after the anneal without some post-anneal modifications but does not threaten the integrity of the holders and their attachments. ·Alternately, modifications to the surveillance capsule holder assemblies prior to annealing such as reflective shielding, localized insulation, invessel annealing of the surveillance capsules, or a combination thereof will be evaluated to ensure compliance with the reembrittlement program described in Section 3.0.
Tables 1.7.D-6 through 1.7.D-9 provide a summary and discussion of.the maximum temperature, maximum stress intensity, maximum temperature gradient,
· and maximum displacement results at key locations integrated from all the analyses performed. Table 1. 7 .D-10 specifies the administrative limits determined
· · by the thermal and stress- analyses that need to.be factored into annealing procedures to ensure that the annealing operation remains well within the limits of Code Case N-557 .
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1. 7 .F Fatigue Analysis
This annealing operation is not intended to be a multiple cycle occurrence. In fact it is intended to be applied just once. Being a one cycle occurrence it will not impact the fatigue life of the reactor vessel. Thus, a fatigue analysis is not required. The ASME Code Case N-557 also does not require a fatigue analysis to be conducted.
1. 7 .G Conclusions
In these thermal and stress analyses specific. to the Palisades reactor vessel it was assumed that the worst case boundary conditions (worst insulation condition, lowest ambient air temperature, and conservative film coefficients) are. present to maximize the temperature differences and thus maximize the stresses. Higher
. annealing temperatures were imposed to show that the stresses are still acceptable. The boundary conditions were changed to induce higher temperatures throughout the reactor vessel and PCS piping. Severe temperature differences well beyond realistic conditions were artificially simulated in key locations such as the RV nozzles and the RV flow skirt area. Lastly, even the maximum stress intensity values were compared against the ASME Code Case N-557 allowable instead of the lower primary plus secondary stress intensities. In all such cases the temperatures, ·stresses, strains, displacements, and residual deformations were such to cause no harmful structural integrity effects to the reactor vessel and PCS piping.
1. 7 .H References
ASME Code Case N-557, "In-Place Dry Annealing of a Nuclear Reactor Vessel, Section XI, Division 1 ", Cases of the ASME Boiler and Pressure Vessel Code.
Mager, T. R., Rishel, R. D., 1982, "Development of a Generic Procedure for Thermal Annealing an Embrittled Reactor Vessel Using a Dry Annealing Method", NP-2493, Re~earch Project 1021-1, Electric Power Research lnstitut~,· Palo Alto, CA.
Vroom, D. W., Hofses A. S., Wong, J., Houstrop, J. P., 1984, "An Evaluation of the Consequences on an In-Place Anneal of a PWR Reactor Vessel", NUREG/CR
- 4212-:-ln-Place Thermal Annealing of Nuclear Reactor _Pressure Vessels (1985), EG&G Idaho, Inc., Idaho Falls, ID.
Server, W. L., 1985, NUREG/CR-4212 : In-Place Thermal Annealing of Nuclear Reactor Pressure Vessels, EG&G Idaho, Inc., Idaho Falls, ID .
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Schall, J. R., Cline, S. C., Wong, J., Houstrop, J. P., 1984, "Further Evaluation of the Consequences on an In-Place Anneal of a PWR Reactor Vessel", NUREG/CR-4212 : In-Place Thermal Annealing of Nuclear Reactor Pressure Vessels (1985), EG&G Idaho, Inc., Idaho Falls, ID.
ASME Code Case N-47-32, "Class 1 Components in Elevated Temperature Service", ASME Boiler and Pressure Vessel Code, Approved August 8~ 1994 .
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• Case Type Run
T1 Thermal
S1 Stress
T2 Thermal
Description/ Characteristics
The heat exchanger top and bottom is insulated. The emissivity of the heat exchanger and the reactor vessel is 0.55, and the RVTC insulation emissivity is 0.90. The average heatup rate of the RV 1.0. surface in the annealing zone is 25.8°F/hr up to 850°F. The RV beltline 1.0. surface is raised to 900°F and is held for 168 hours. Cooldown is controlled by a drop in the heat exchanger average surface temperature of 100°F/hr. The ambient air is assumed to be 100°F. The worst case thermal conductivity for the RV insulation as measured following the 1995 outage and adjusted for annealing conditions is used. The stress free reference temperature is 70°F.
The.rmal loading is that of T1. The steam generator end rotation is fixed but free radial movement is allowed. ASME material properties are used.
Same as Case T1 except that the RVTC insulation emissivity is 0.55.
Intent of Evaluation
This case uses the intended heat exchanger configuration and performance combined with thermal boundary conditions that lead to conservative thermal gradients, and thus conservative stresses compared to the intended anneal.
This case uses the intended heat exchanger configuration and performance combined with thermal boundary conditions that lead to conservative thermal gradients, and thus conservative stresses compared to the intended anneal.
This case evaluates the effect of a 40% decrease in the emissivity of the RVTC on the temperatures in the· upper part of the reactor vessel.
Tab~e 1.7.C-1 Description of Postulated Annealing Conditions Evaluated With 2-D Model for the Palisades Reactor Vessel Annealing (Page 1 of 5)
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• Case
S2
T3
S3
T4
S4
Table 1.7.C-1
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Type Run Description/ Intent of Evaluation Characteristics j
j 1
Stress Same as Case S 1 except that the thermal loading is This case was· not evaluated because the that of Case T2. temperature distributions caused by the
Case T1 and Case T2 were similar.
Thermal Same as Case T1 except that the heat exchanger This case evaluates the effect of an top is uninsulated. absence of the heat exchanger top
J' insulation on the temperatures in the upper part of the reactor vessel.
Stress Same as Case S1 except that the thermal loading is This case evaluates the effect of the RV that of Case T3. flange getting hotter than the Case T1.
Thermal Same as Case T1 except that the RV beltline, RV This case evaluates the effect of a wall '
nozzle and RV flange walls are assumed to be 0.3" thickness variation on the temperatures in thicker. the reactor vessel and attached equipment. .
' Stress Same as Case S 1 except that the thermal loading is This case evaluates the effect of added wall
that of Case T4. thickness of the RV beltline, RV nozzle, and RV flange on the stress distribution.
Description of Postulated Annealing Conditions Evaluated With 2-D Model for the Palisades · : Reactor Vessel Annealing (Page 2 of 5)
1.7-25
• Gase
T5
S5
S6
Table 1.7.C-1
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Type Run Descriptionl Intent of Evaluation Characteristics
Thermal Same as Case T1 except that the emissivity of This case evaluates the effect of a 20% the heat exchanger and reactor vessel is 0.45. decrease in the heat exchanger and RV
emissivity on the range of the heat exchanger design maximum heat flux capapility.
Stress Same as Case S 1 except that the thermal This case evaluates the effect of this therm~I loading is that of Case T5. loading on the stress distribution.
Stress Same as Case S 1 except that the steam This case evaluates the effect of a pinned generator end is free to rotate and free to move connection at the equipment end (steam radially. generator) on the moments at the RV nozzle to
pipe interface.
Description of Postulated Annealing Conditions Evaluated With 2-D Model for the Palisades Reactor Ve~sel Annealing (Page 3 of 5)
1.7-26
• Case Type Run Description/ Intent of Evaluation
Characteristics I
S7 Stress Same as Ca$e S1 except the stress-free reference This case evaluates a change in the temperature for the cladding and base metal is 400°F reference temperature on the displacement instead of 70°F and the yield stress of the cladding is 45 and the bending stresses as well as ksi. residual stresses and deformations of the
reactor vessel.
T6 Thermal Same as Case T1 except that the heatup rate has been This case evaluates the effect of a faster increased at the beginning of the heatup period to initial heatup rate on the temperatures in · approximately 40°F/hr however the average heatup rate the reactor vessel and piping. is still 25°F/hr up to 850°F.
S9 Stress Same as Case S1 except the thermal loading is that of This ·case evaluates the effect of a faster Case T6. initial heatup rate on the temperatures in
the reactor vessel and piping.
Tablle 1.7.C-1 · D.escription of Postulated Annealing Conditions Evaluated With 2-D Model for the Palisades. Reactor Vessel Annealing (Page 4 of 5)
TAR 3/25/96 1.7-27
• '
Case Type R~n Description/ Intent of Evaluation i.
17
S10
Tablie 1.7.C-1
TAR 3/25/96
Characteristics , .. I
Thermal Same as Case T1 except that a combination of effects has This case evaluates the effects of been added to make the RV temperatures hotter. These imposing characteristics intended to include: increasing the effectiveness of the RV insulation by make the temperatures hotter and 33%; increasing the ambient air temperatures consistent establishing the resultant temperatures with the forced cooling analysis described in Section 1.3 (i.e. throughout the reactor vessel and piping. 134°F at the seal ledge to 164°F at the bottom head region.); no heat loss due to convective currents inside the PCS piping; and, no RV support heat losses.
Stress Same as Case S 1 except the thermal loading is that of Case This case evaluates the effects of 17. imposing characteristics intended to
make the temperatures hotter and establishing the resultant stresses throughout the reactor vessel and piping.
Description of Postulated Annealing Conditions Evaluated With 2-D Model for the Palisades Reactor Vessel Annealing (Page 5 of 5)
1.7-28
• Case Type Run
T1 .1 Thermal
•• ·~
Description/ Characteristics
The heat exchanger has both top and bottom insulation. The emissivity of the heat exchanger and the reactor vessel is 0.55, and the reactor vessel top cover insulation emissivity is 0.90. The average heatup rate of the RV 1.0. surface in the annealing zone is 25.8°F/hr up to 850°F. The temperature of the RV beltline 1.0. surface is approximately 900°F and is held for 168 hours. Cooldown is controlled by a drop in the heat exchanger average surface temperature of 100°F/hr. The ambient air temperature is assumed to be 100°F. The worst case thermal conductivity for the RV insulation as measured following the 1995 outage and adjusted for annealing conditions is used. The stress free reference temperature is 70°F. These are similar to the 2-0 analysis, Case T1.
Intent of Evaluation
.·~ I I
This case uses the intended heat exchanger configuration and performance combined with thermal boundary conditions that lead to conservative thermal gradients, and thus conservative stresses compared to the intended anneal.
Table 1.7.C-2 Description of Postulated Annealing Conditions Evaluated With the .Full 3-D Model for the: . Palisades Reactor Vessel Annealing (Page 1 of 6)
TAR 3/25/96 1.7-29
• • ·,
' • I
Case Type Run Description/ Intent of Evaluation I Characteristics
T1.1 Stress Thermal loading is that of Case T1 .1. The steam generator This case uses the intended heat end is free to rotate and free radial movement is allowed. exchanger configuration and performance The RV nozzle support is fixed in rotation. ASME material combined with thermal boundary properties are used. conditions that lead to conservative
thermal g~adients, and thus conservative stresses. And it is used to ascertain the resultant stresses if the nozzle support
- does not rotate.
T1.1P Stress Thermal loading is that of Case T1 .1. The steam generator This case uses the intended heat end is free to rotate and free radial movement is allowed. exchanger configuration and performance The RV nozzle support is allowed to rotate per design. combined with thermal boundary
' ASME material properties are used. conditions that lead to conservative thermal gradients, and thus conservative stresses. This case also allows for a
: comparison to be made with Case T1 .1, Stress as to the effects of the nozzle support rotation .
Table 1.7.C-2 . Description of Postulated Annealing Conditions Evaluated With the Full 3-D Model for the. Palisades Reactor Vessel Annealing (Page 2 of 6)
TAR 3/25/96 1.7-30
•• Case
T1 .1 F
T1.2
T1.2
. Table 1.7~C-2
TAR 3/25/96
Type Run
Stress
Thermal
Str~ss
• Description/
Characteristics
Thermal loading is that of Case T1 .1. The steam generator end is fixed in rotation but free radial movement is allowed. The. RV nozzle support is allowed to rotate per design. ASME material properties are used~· This case is intended to be the most representative of the intended anneal.
Same as Case T1 .1, Thermal except that the heat exchanger temperatures are slightly greater (for example, the upper guard zone temperature during the hold period is 30°F greater than Case T1 . 1, Thermal)
Same as Case T1 .1 •. Stress except that the thermal loading was that of Case T1 .2, Thermal, and the steam
' ' '
generator end is fixed but free radial movement is allowed. The RV nozzle support is fixed in rotation.
Intent of Evaluation '
I This case uses the intended heat exchanger configuration and performance combined with thermal boundary conditions that lead to conservative thermal gradients, and thus conservafo1e stresses. This case also allows for a comparison to be made with Case T1.' P, Stress as to the e.ffects of the steam generator being fixed.
This case is intended to show that there is significant stress margin in the annealing operation even if the !
temperatures in. the anneal zone were allowed to go up to 950°F.
This case is intended to show that there :is significant stress margin in the annealing operation even if the temperatures in the anneal zone were allowed to go up to 950°F. .
Description of Postulated Annealing Conditions Evaluated With the Full 3-D Model for the Palisades Reactor Vessel Annealing (Page 3 of 6)
t.7-31
Case Type Run I
TC1 Thermal
T2 Thermal
T2 Stress
T3 Thermal ..
Table 1.7.C-2
TAR 3/25/96
• • .1
Description/ , Intent of Evaluation '
Characteristics
I 1
Same as Case T1 .1, Thermal except that a varying heat This case evaluates the effect of exchanger temperature distribution was applied which resulted in azimuthal temperature gradients in the a smaller net heat energy. The heat exchanger thus varied in heat exchanger on the temperature temperature both axially and azimuthally. The azimuthal distribution of the reactor vessel. temperature gradient was 50°F, consistent with results from the heat exchanger Proof of Principle Test.
Same as Case T1 .1, Thermal except that a higher ambient air This case was intended to create a iow temperature was used, the RV insulation properties were 33% heat loss bounding condition. better, and it was assumed that there was no internal PCS piping convection to the steam generators and primary coolant pumps.
Same as Case T1 .1, Stress except the thermal loading was that This case was intended to create a low of Case T2, Thermal, and the steam generator end is fixed but heat loss bounding condition. free radial movement is allowe'd. The RV nozzle support is fixed in rotation.
An artificial axial gradient was imposed at the annealing zone This case was intended to create a 1
ends. bounding case for RV wall bending .
Description of Postulated Annealing Conditions Evaluated With the Full 3-D Model for the Palisades Reactor Vessel Annealing (Page 4 of 6)
1i.7-32
• • ~ !
Case Type Rur:i Description/ Intent of Evaluation '
I Characteristics I
T3 Stress Same as Case T1 .1, Stress except the thermal loading was that This case was intended to create a of Case T3, and the steam generator end is fixed but free radial bounding case for RV wall bending. movement is allowed. The RV nozzle support is fixed in rotation.
T4 Thermal The temperatures of Case T1 .1, Thermal at 36 hours were This case was intended on creating a adjusted. The finite element node temperatures for the nodes bounding axial temperature gradient between the vessel'l.D. and the nozzle extension above the case across the RV nozzles to piping centerline were halved. The temperature difference evaluate the limiting condition on
' between the top and bottom of the nozzle is about 470°F. primary coolant loop pipe bending.
T4 Stress Same as Case T1 .1, Stress except that the forced axial This case was intended on creating ·a temperature gradient of Case T4, Thermal was imposed across bounding axial temperature gradient the RV nozzle region and the steam generator end is fixed but case across the RV nozzles and free radial movement is allowed. The RV nozzle support is fixed determine the resultant stresses. in rotation .
Table 1.7.C-2 . Description of Postulated Annealing Conditions Evaluated With the Full 3-D Model for the Palisades Reactor Vessel Annealing (Page 5 of 6)
TAR 3/25/96 t.7-33
• ••
Case Type Run Description/ Intent of Evaluation
T6 Stress
'
Table 1. 7 .C-2
TAR 3/25/96
Characteristics
A 140 kips force and 1.4 x 107 in-lbs moment (to account This case was intended to determine if the for the base of the steam generator support being 100 reactor vessel head ovalization stays in the inches below the nozzle) were applied at the steam elastic region. Thus if the friction force at generator location and the nozzle support is fixed. The the steam generator support prevents a cross-over leg was assumed present in one case, and in return of the reactor vessel displacement another case it was assumed to be absent. during cooldown a mechanical jacking of
the steam generator will return the reactor vessel to its original shape.
Description of Postulated Annealing Conditions Evaluated With the Full 3-D Model for the Palisades Reactor Vessel Annealing (Page 6 of 6)
1.7-34
-•--Period
Heatup
Landing
• Hold
Cool down
Table 1.7.C-3
• TAR 3/25/96
Time RV Inside Surface Comments (hours) Temperature
(Of)
0 75
13 420
21 631
30 850
32 885
34 895,
36 900
40 900
204 900
213 < 800 Heat exchanger surfaces are assumed to cool down at 100°F/hr
360 < 200
RV .. Temperature Response in the Annealing Zone Used for All 2-0 and 3-0 Analyses Except Where Noted in T~bles 1.7.C-1 and 1.7.C-2
1.7-35
-·-
•
•
Region . Thermal Cases (Of)
T1 T2 T3 T4 TS T6 T7
. RV Flange 555 550 715 550 550 715 605
'
RV Bottom Head 855 855 855 850 885 850 880
RV Nozzle Extension 285 285 315 285 285. 310 495
PCS Piping - Midloop 135 135 140 135 135 140 255
Steam Generator
Table 1.7.D-1
TAR-3/25/96
105 105 105 105 105 105 165
Maximum Temperatures at Key Node Locations/Regions Established Using The 2-D Model for the Palisades Reactor Vessel Annealing
1.7-36
• ·,·· • I Radial Displacement Vertical Displacement
The displacements are relative to the initial condition. . Strain relative to initial strain condition. Case 89 was analyzed only from 0 to 40 hours. Maximums would have occurred later. Case 82 was not performed because temperature differences between Cases T1 and T2 were insignificant. Reported values are in the cladding. Base metal values are all less. The RV beltline and RV bottom head displacements are downward, the RV flange displacements are upward At approximately the 609 foot elevation or at the bottom of the top 1 /3 of the RV beltline.
' Peak Stress Intensities, Peak Strains, and Peak Residual Deformations Established Using the 2-D Model for the Palisades Reactor Vessel Annealing ·
1.7-37
-•--Item
-
Strains averaged through the thickness
Strains at the surface (Equivalent Linear Distribution)
Local ·strains at any point
Note: (1)
Table 1.7.D-3
-·- TAR 3/25/96
Strain Strain Strain 111
(creep) (total) (allowable)
(at end of (at end of cooldown) heatup)
- . .. .. . .. . ·- .. . -- .. .. - - w
0.04% 0.09% 1%
0.05% 0.20% 2%
0.08% 0.26% 5%
ASME Code Case N-47-32 used for comparison only.
Inelastic Analysis Strains and Allowables Determined Using the 2-D Model for the Palisades Reactor Vessel Annealing
Notes: (1) These_ are estimates based on review of nodal temperatures in the model at 204 hours. Cases T3 and T4 have been omitted because they represent artificially imposed thermal gradients and they do not represent the entire annealing process particularly the highest temperature point at 204 hours. Case TC1 has been omitted becaus,e it does not represent the entire annealing process particularly the highest temperature point at 204 hours.
(2) Maximum temperature at bottom of RV flange. The value in parenthesis is at the top of the flange.
(3) Lower temperatures are observed due to smaller net thermal energy being imposed on the reactor vessel. · ·
Table 1.7.D-4 Maximum Temperatures at Key Node Locations/Regions Established Using The 3-D Model for the Palisades Reactor Vessel Annealing
(1) The RV beltline and RV bottom head displacements are downward, the RV flange displacements are upward. (2) · These maximum stress intensities are at specific through-wall cuts at the locations specified and exclude '
. cladding. The hours represent when these values occurred (i.e. 30 hrs, 36 hrs, or 204 hrs into the annealing process). · ·
(3) Differences between the stress intensities at 36 and 204 hours are minimal. (4) - Represents an artificially generated thermal loading and is independent of time. Displacements at 204 hours
cannot be presented to compare against the other cases. · -(5) , The value of 0.25 inch has been added to the 3-D model results at t=204 hours to conservatively remove the
effect of the RV support structure rising due to thermal expansion.:
Table 1.7.0-5 Maximum Displacements, and Maximum Stress Intensities Established Using the 3-0 Model for the Palisades Reactor Vessel Annealing
TAR 3/25/96 1'..7-40
' 't
i I
COMPONENT /STRUCTURE MAXIMUM COMMENTS TEMP.(°F) 111
RV Flange 779 This is from Case T1 .2, 3-0. Actually the temperature varies from 779°F to 645°F from the bottom to top of the RV
I flange. It is due to the hotter temperature in the anneal zone. .
RV Beltline: Temperatures based on the hottest case (Case T2, 3-0). 1.0. Surface during soak 900 Similar results seen in Case T1, 2-0 (Figure 1. 7 .0-1 ). The 0.0 .. surface during soak 875 1.0. surface temperature was the analytical target.
RV Bottom Head 926 This is from Case T1 .2, 3-0.
RV Nozzle Support Pads 563 This is from the hotte~t case (Case T2, 3-0). Temperature at bottom of pad is approximately 500°F. This is within the design temperature of 650°F thus potential degradation of the dry lubricant is precluded.
Vessel Support Steel 420 This is from Case T1 .2, 3-0. The temperature at the bottom of the supports is assumed as 100°F. Since the RV nozzle support pads,are within the design temperature this would be within design as well.
RV Seal Ledge 645 This is from Case T1 .2, 3-0. RV Core Stabilizing Lugs 1025 Based on an alternate calculation. Highest temperature is at
the tip of the lug nearest the heat exchanger. Temperature at reactor vessel is 900°F.
Notes: (1) These are estimates based _on review of temperature contour plots and nod.al temperatures in model.
Table 1.7.D-6
TAR 3/25/96
Maximum Temperatures at Key Locations Defined by the 3-D Model and Alternate Calculations for the Palisades Reactor Vessel Annealing (Page 1 of 2)
1.7-41
MAXIMUM I COMPONENT /STRUCTURE TEMP. (°F)111 COMMENTS
RV Core Support Lugs 900 This is from the hottest case (Case T2, 3-0). I
I
I RV Surveillance Capsule Holder Support 924 Based on an alternate calculation assuming 900°F at the RV Bracket wall interface. . I RV Outlet Nozzle Extension 612 This is from the hottest case (Case T2, 3-0).
Steam Generator 164 Based on an alternate calculation. The temperatures in the steam generators remains well below the normal operating temperatures.
Primary Coolant Pump 164 Based on an alternate calculation. The temperatures in the pumps remain well below the normal operating temperatures.
PCS Piping: Within Biological Shield Wall @ Nozzle These are from the hottest case (Case T2, 3-0). The Extension temperatures beyond this point are less than the design Inlet. 661 temperature (650°F). Thus the PCS pipe and PCS pipe Outlet 612 insulation temperatures will be within the design basis.
Outside of Biological Shield Wall @ This is from the hottest case (Case T2, 3-0). This Mid loop demonstrates that the temperatures of the PCS flow and: Inlet 204 temperature measurement instrumentation, and PCS piping Outlet 115 connections are within normal operating temperatures.
Notes: ( 1) These are estimates based on review of temperature contour plots and nodal temperatures in model.
Table 1. 7 .D-6
TAR 3/25/96
· Maximum Temperatures.at Key Locations Defined'by the 3-D Model and Alternate Calculation for the Palisad~s Reactor Vessel Annealing (Page 2 of 2)
1·.7-42
-.,_
MAX. MAX./ I COMPONENT /STRUCTURE STRESS ALLOW. COMMENTS
INTENSITY (%)12)(3)
(ksi)', 1121
Anywhere in Reactor Vessel 58.8 82 This is from Case T1 .2. 3-D at 36 hours. It is a max. j ( 46.8) ( 65 ) stress intensity and it is occurring at an edge of the RIV
nozzle support. It is related to the finite element mod~I in that the nozzle support is not allowed to rotate.
RV Flange 28.8 40 This is from Case T1 .2, 3-D at 36 hours. It is occurring ( 27.6) ( 38) at the cladding/base metal interface.
RV Inlet Nozzle Extension 40.5 141 73 This is from the Case T3, 3-D. It is occurring on the ( 28.4) ( 51 ) O.D. surface.
RV Inlet Nozzle Transition 47.5 141 66 This is from the Case T3, 3-D. It is occurring on the ( 41.1) ( 57) O.D. surface.
RV Outlet Nozzle Extension 42.6 141 77 This from the Case T3, 3-D. It is occurring on the 0.D. ( 29.5 ) ( 53) surface.
RV Outlet Nozzle Transition 50.3 70 This is from the Case T3, 3-D. It is occurring on the
Note: (1)
(2)
(3)
(4)
Table 1.7.D-7
TAR 3/25/96
( 46.8) ( 65) O.D. surface. '
These are estimates taken from stress intensity contour plots and from specific through-wall cuts in the 3-D analyses. · Other alternate calculations were also used. Cladding has been excluded. , Value from Case T1 .1 F, a more representative annealing case, is shown in parenthesis.
For the reactor vessel a Sm of 24 ksi from Code Case N-557 is used. This is the Sm value for 900°F. Sm values for lower temperatures are greater than 24 ksi thus the 3Sm will be higher. For the nozzle extensions, a Sm of 18.4 ks.i from ASME Section Ill is used. This is the Sm value for 650°F. Sm values for lower temperatures are greater than 18.4 ksi thus the 3Sm will be higher. Max./Allow. = (Max. Stress lntensity/3Sm) x 100. Case T 4 not included due to plasticity. ':
Maximum Stress Intensities Defiried Using 3-D Model and Alternate Calculations for the Palisades Reactor Vessel Annealing (Page 1 of 4)
RV Upper Shell to Intermediate 26.8 37 This is from Case T1 .1 F and T1 .1 P, 3-D at 30 hours. It Shell Transition Region ( 26.8 ) ( 37) is occurring at the cladding/base metal interface.
RV Bottom Head Transition 22.6 31 This is from Case T1 .1, T1 .1 F, and T1 .2, 3-D at 30
RV Flow Skirt
RV Seal Ledge
Note: (1)
(2)
(3)
Table 1.7.D-7
TAR 3/25/96
( 22.6 ) ( 31 ) hours. It is occurring at the cladding/base metal interface.
50.4 n/a This is from Case T3, 3-D. This is at the top tip of the ( 43.0) RV flow skirt. This is an internal attachment thus Code
Case N-55 7 is not applicable. The maximum strain is 1.9% and it is localized.
43.4 60 This is from an alternate calculation. This is a maximum stress intensity and is at the tip of the RV seal ledge.• At the flange/seal ledge interface the stress intensity is 24.5 ksi.
These are estimates taken from stress intensity contour plots and from specific through-wall cuts in the 3-D analyses. Other alternate calculations were also used. Cladding has been excluded. Value from Case T1 .1 F, a more representative annealing case, is shown in parenthesis. For.the reactor vessel a Sm of 24 ksi from Code Case N-557 is used. Thi.sis the Sm value for 9Q0°F. Sm values for lower temperatures are greater than 24 ksi thus the 3Sm will be higher. For the nozzle extensions, a Sm of 18.4 ksi from ASME Section Ill is used. This is the Sm value for 650°F. Sm values for lower temperatures are greater than 18.4 ksi thus the 3Sm will be higher. Max./ Allow. = (Max. Stress lntensity/3Sm1 x 100.
Maximum Stress Intensities Defined Using 3-D Model and Alternate Calculations for the Palisades Reactor Vessel Annealing (Page 2 of 4)
t.7-44
MAX.
.. ··.
COMPONENT /STRUCTURE STRESS INTENSITY
(ksi)111121
MAX./ ALLOW.
(%)12)(3) COMMENTS
RV Core Stabilizing Lugs <38.7 <54 This is from an alternate calculation. These lugs see / smaller stress intensities than the RV core support lugs because they are located opposite the lower guard zo~e of the heat exchanger and will see smaller axial thermal gradients.
( <31.1) ( <43)
RV Core Support Lugs @ attachment to reactor vessel
38.7 ( 31.1 )
54 This is from Case T3, 3-0. ( 43 )
RV Surveillance Capsule Holder Support Bracket Weld at RV Wall
40.4 56 This is·from an alternate calculation.
Steam Generator Inlet Nozzle 6,5 This is· from an· alternate calculation. This is a conservative estimate since the steam generator end pipe is assumed fixed in the vertical direction and in all rotations when in actuality it has some flexibility.
Note: (1)
(2) (3)
Table 1.7.D-7
TAR 3/25/96
These are estimates taken from stress intensity contour plots and from specific through-wall cuts in the 3-D analyses. Other alternate calculations were also used. Cladding has been excluded. · Value from Case T1 .1 F, a more representative annealing case, is shown in parenthesis. For the reactor vessel a Sm of 24 ksi from Code Case N-557 is used. This is the Sm value for 900°F. Sm values for lower temperatures are grea~er than 24 ksi thus the 3Sm will be higher. For the nozzle extensions, a Sm of 18.4 ksi
·from ASME Section Ill is used. This is the Sm value for 650°F. Sm values for lower temperatures are greater than 18.4 ksi thus the 3Sm will be higher. Max./ Allow. = (Max. Stress lntensity/3Sm) x 100.
Maximum Stress Intensities Defined Using 3-D Model and Alternate Calculations for the . Palisades Reactor Vessel Annealing (Page 3 of 4)
Pr,imary Coolant Pump Discharge <6.5 --- This is an estimate based on the stea~ generator inletj Nozzle nozzle calculation given above. The pump is on a longer
piping leg than the steam generator thus the effects of the annealing operation are less.
' PCS Piping <42.6 <77 This is an estimate based on the maximum stress ( <29.5) ( <53) intensity at the RV outlet nozzle extension from Case ·
T3, 3-D. Effects on the piping beyond this point are reduced substantially
PCS Piping (assuming contact 51.6 93 This is an alternate calculation in which the pump was with the lateral support) assumed to be in contact with a lateral support. This.is
not expected to occur.
RV Outlet Nozzle Cpincident With 31.6 88 From Case T1 .1 F, 3-D at 204 hours. This is a total Outer RV Shell
Note: ( 1)
(2) (3)
Table 1.7.D-7
TAR 3/25/96
stress intensity within the limits of reinforcement instead of the average through wall stress. The allowable is 1.5 Sm per NB-3227.5, where Sm of 24 ksi at 900°F is used for conservatism .
. These are estimates taken from stress intensity contour plots and from specific through-wall cuts in the 3-D analyses. Other alternate calculations were also used. Cladding has been excluded. Value from Case T1 .1 F, a more representative annealing case, is shown in parenthesis. For the reactor vessel a Sm of 24 ksi from Code Case·N-557 is used. This is the Sm value for 900°F~ Sm values for lower temperatures are greater than 24 ksi thus the 3Sm will be higher. For the nozzle extensions, a Sm of 18.4 ksi from ASME Section Ill is used. This is the Sm value for 650°F. Sm values for lower temperatures are greater than 18.4 ksi thus the 3Sm will be higher. Max./AUow. = (Max. Stress lntensity/3Sm) x 100.
· Maximum Stress Intensities Defined Using 3-D Model and Alternate Calculations for the Palisades Reactor Vessel Annealing (Page 4 of 4)
1.7-46
-MAX. I
TYPE TEMP. COMMENTS GRADIENT
(Of)
Reactor Vessel:
Through-Wall (in RV Beltline): '
During Heatup 100 This is from Case T1, 2-D. It is shown in Figure 1.7.D-2. The maximum thermal gradient appears at the middle of the RV beltline ..
During Soak 50 This is from Case T1, 2-D. It is shown in Figure 1.7.D-2. The maximum thermal gradient appears at the top of the RV beltline.
During Cooldown 70 This is from Case T1, 2-D. It is shown in Figure 1. 7 .D-2. The maximum thermal gradient appears at the middle of the RV beltline.
Azimuthal: 50 This is from Case T5, 3-D where a forced RV azimuthal temperature distribution was imposed on the RV l.D. surface. In Case TC1, 3-D where a 50°F azimuthal temperature distribution was imposed on the heat exchanger only a 25°F temperature was observed on the RV wall (Figure 1. 7 .D-5).
Axial: 470 This· is from Case T 4, 3-D where a forced large axial temperature distribution was imposed on the RV l.D. surface (Figure 1. 7 .D-6).
Table 1.7.D-8 Thermal Gradients Defined Using 2-D and 3-D Models for the Palisades Reactor Vessel Annealing
TAR 3/25/96 1.7-47
••• COMPONENT /STRUCTURE MAX.
·RADIAL DISPLACE.
(IN)
RV Flange (Top) 0.55
RV Bottom Head 0.054
Notes: (1) Displacement is in the downward direction.
• MAX.
VERTICAL DISPLACE.
(IN)
0.95
2.23111
• I
COMMENTS
This is from Case T1 .2, 3-D at 204 hours. These thermal displacements are larger than the design temperature growths. A structural integrity analysis of the monitor tube and piping in accordance with ASME Code Section Ill has shown this addition13I growth to be acceptable.
These are from Case T1 .2, 3-D at 204 hours. The vertical displacement is approximately 1.25 inches longer than the normal operating conditions. The worst case effect of this increased displacement is that the 4-inch thick block insulation · below the reactor vessel is crushed and ~he stainless steel floor covering is deformed 1.25 inches. This condition is acceptable in that operability of the floor and the cooling coils in the floor are not affected.
Table 1.7.D-9 Maximum Displacements at Key Locations Defined Using the 3-D Model (Page 1 of 2)
TAR 3/25/96 1:.7-48
• COMPONENTfflTRUCTURE
Steam Generator End of Pipe
Primary Coolant Pump End of Pipe
Crossover Leg Pipe
MAX. RADIAL
DISPLACE. (IN)
1.12
0.81
0.92
MAX. VERTICAL DISPLACE.
(IN)
0.00
0.00
0.00
·~ !
COMMENTS
These are from Case T2, 3-D at 204 hours. The maximum radial displacement is less than the normal operating condition radial displacement of 1. 125 inches. This displacement is less than the travel limits of Steam Generator A of 1.99 inches and Steam Generator B of 1.41 inches. This is an acceptable condition.
These are from Case T2, 3-D at 204 hours. This maximum radial displacement is less than the normal operating condition radial displacement.
These are from Case T2, 3-D at 204 hou_rs. These are less than that experienced during normal operating conditions. These are · acceptable.
Table 1.7.D-9 Maximum Displacements at Key Locations Defined Using 3-D Model (Page 2 of 2)
TAR 3/25/96 1.7-49
• • . \
Administrative Lin;iits I
Category Administrative Limits Reason for Selection I Heatup Average RV annealing
; Heatup rate is important in that the maximum RV strains and maximur
zone heatup of 25°F/hr RV nozzle stresses occur at the end of heatup and at the beginning ofj from ambient up to the hold period. This specified heatup rate was analyzed in both the 2,..D 850°F " and 3-D analysis and shown to be acceptable. The faster heatup rate[
early in the heatup was shown in the 2-D analysis to result in hotter 40°F/hr maximum temperatures outside the annealing zone and a reduction of the. axial heatup rate from thermal gradient across the RV nozzles. A faster heatup rate earlier ambient up to 600°F results in a slower heat up rate later in order to meet the average rate!
requirement. A slowdown in this heatup rate towards the end of hea~up is better in that the thermal enen::iv is allowed to distribute more. i
I
Time at Temperature Hold at 850°F to The 2-D and 3-D analysis results have shown that this target time and 900°F for 168 hours temperature combination yields acceptable material recovery. !
Cooldown Average surface This cooldown rate was predicted in both the 2-D and 3-D analysis arid temperature cooldown shown to be acceptable. . I rate in the RV annealing zone I
I cooldown of ~25°F/hr ' to 210°F I
RV Bottom Displacement Range The maximum RV bottom displacement predicte.d by the 3-D analysis iis Displacement (to be determined prior 2.23 inches. A more realistic displacement range will be determined I
to annealinQ) prior to annealinQ. ! Steam Generator ~ 1.375 inches The maximum steam generator end displacement predicted by the 3-0 Displacement analysis is 1.12 inches. ! PCS Piping. Displacement Range The displacements predicted for the piping in the 3-D analysis were I
I
Displacement shown to be acceptable. A more realistic displacement range will be I determined prior to annealing. !
' i
Table 1. 7 .D-10 · Administrative Limits Based on Analysis Results (Page 1 of 2)
TAR 3/25/96 1 7-50
• Administrative Limit
Category
Axial Temperature Gradients Across RV Nozzles Near End Of Heatuo Azimuthal Temperature Gradient Across .RV Flow Skirt
Administrative Limit
:!:470°F
:!:50°F
Reason for Selection
The bending stresses in the PCS piping can be minimized by minimizing the axial thermal gradient across the RV nozzles. The Case T4, Stress of the 3-D analysis showed that this limit is acceptable. The most critical azimuthal temperature variation is around the RV flow skirt area.· Case T3, Stress of the 3-D analysis ·~hawed that t.his limit is acceptable.
Through-Wall ::!: 100°F These thermal gradients were predicted in the 2-D and 3-D Temperature analysis and were acceptable. The thermal gradient across Gradient Across RV the RV beltline wall during the hold period is the most crucial Beltline Wall During .in that it ensures that at the target annealing temperature of Heatup 850°F on the RV l.D. surface the rest of the RV beltline is
11-~~'--~~~~~+--~~~~~~~----1
Through-Wall :!:50°F above 800°F. Temperature Gradient Across RV Beltline Wall During the Hold Period Through-Wall Temperature Gradient Across HV Beltline Wall During the Cooldown Period
:!:70°F
Table 1.7.0-10 Administrative Limits Based on Analysis R.esults (Page 2 of 2)
TAR 3/25/96
I I I r-51 I
I I
_. ___ :_~-
• r'
RV Top Cover Insulation---. (Included In ThennaJ Analysis Only)
I 111 I 111 t-ift -1 111 I 111 1 111 1 111 1 111 I 111 I 111 1 111 ~
I
------l------: . I
I I I I I
-------+-------
II I I II I I
- - - - - - t+----l~l-I-+ ~ 111 I II I I II I I II I I II I I Ill I 111 I II I I
tt---t-1++ -1
Inside View at t = 40 Hours, Case TC1, 3-D Analysis
/
1.7-60
·-·
- - ~-·
•
A = 180.244 DEG·F B = 274.961 DEG·F C = 369.678 DEG·F D = 464.394 DEG·F E = 559.111 DEG-F F = 653.828 DEG·F G = 7 48.544 DEG·F H = 843.261 DEG·F I = 937.978 DEG·F
I I
----t-,-~- ~.,
,
I I
I ' I
Figure 1.7.D-6
TAR 3/25/96
, ,
----·--- ---------- --~ ------ --- -- -----
370 + .275/2 = 323 DEG·F
AVERAGE ..._..___...-748}843/2=795
Cross Section of RV Outlet Nozzle for Case T4, 3-D Analysis
The Thermal Annealing Operating Plan portion of the Thermal Annealing Report (TAR) describes the equipment evaluations, operating conditions and analyses used to support the implementation of RV thermal annealing at the Palisades Nuclear Plant. -It also addresses the heating method, equipment, procedures, ALARA principles, and limiting parameters to be used during the annealing operation. Implementation of this plan will result in the restoration of RV beltline material properties to allow operation through at least March 2011. If the reactor vessel is not annealed, the Palisades limiting beltline axial weld (Heat W5214) will reach the NRC's pressurized thermal shock (PTS) screening criteria in 10 CFR Part 50.61 shortly after the end of cycle 14 (1999) based on the current SER. Detailed information is provided in each of the TAR sections in accordance with the format and content delineated in draft Regulatory Guide DG-1027. A summary of the key material from each TAR Thermal Annealing Operating Plan Section is included below. --·-='---=--'=-
The final TAR submittal, to be submitted after the Marble Hill Annealing Demonstration Program (ADP) will meet the requirements and intent of Reg Guide 1.162 though the format may remain per draft Reg Guide DG-1027.
1.10.A General Considerations (TAR Section 1.1)
The predicted recovery of the limiting RV beltline axial weld material RT NOT is 90% using the transition temperature recovery model of NUREG CR-6327 and the target annealing parameters of 850-900°F for 168 hours. The minimum predicted recovery of RT NoT and the Charpy upper shelf energy (USE) for any b'eltline material is 88%. The material closest to its PTS screening.criteria limit follQwing t_he ~nneal will have a po~t anneal RT NOT of 56°F. The post anneal RT NOT values for all the beltline materials are provided in Table 1.1.C-3. The lowest projected post anneal Charpy USE for the beltline material is 72 ft-lb. The post anneal Charpy USE values for all beltline materials are provided in Table 1.1.C-4.
The lateral shift method was used to predict the reembrittlement response of the RV beltline material. The material closest to the PTS screening criteria has a projected RT NOT of 212°F in March 2011. This material is the beltline axial weld (Heat W5214)-. RT NOT projections for all beltline materials are provided in Table 1.·1. C-3. The lowest projected Charpy USE in March 2011 is 55 ft-lb. Projected Charpy USE values for all beltline materials are-provided inTable 1.1.C-4. Projected values for Charpy USE and RT NOT indicate that the Palisades reactor vessel will maintain adequate toughness through the current end of license (EOL) and beyond to at least March 2011.
A supplemental RV material surveillance program will be used to support the annealing recovery and reembrittlement projections. Since the Palisades RV beltline weld
TAR 3/25/96 · 1.10-1
re. material is not represented in the original surveillance program, additional weld material whie11-is-ereeil31e-witl1-resJ;>eet-to-1-Q-GFR-58:6~-t'las-13een-aeeee-t0-the-slolFVeillanee-
Program. The most limiting base metal from a RT Nor perspective is represented in the original surveillance program. The supplemental surveillance program will provide the necessary information to complete the Fracture Toughness Recovery and Reembrittlement Assurance Program. -
1.10.B Description of the Reactor Vessel (TAR Section 1.2)
The Palisades RV beltline is constructed of manganese molybdenum steel plate purchased to ASME SA-3028 with modified nickel content. The RV beltline minimum wall thickness is 8.5 inches. The RV beltline incorporates portions of six plates, six axial welds, and a full circumferential weld. The most limiting with respect to RT Nor chemistry condition for the RV base metal is 0.24% copper and 0.52% nickel. The most limiting weld chemistry is 0.212% copper and 1.02% nickel. ·The internal surfaces of the reactor vessel that are in contact with the reactor coolant are clad with 308 stainless steel which is· attached to the reactor vessel with a layer of 309 stainless steel for a total nominal thickness of 0.25 inch.
There are six RV nozzles fabricated from A-508 Class 2 forgings; two 42 inch l.D. hot ·leg nozzles and four 30 inch l.D. cold leg nozzles. The reactor vessel is supported by three pads, two of which are located under inlet nozzles with the other located under an
· ·outlet nozzle. The support pads are bolted to sole plates that can slide on a set of plates mounted to steel I-beam supports, which permit radial thermal movement of the reactor vessel. Internal attachments to the reactor vessel include six surveillance holder assemblies, nine core support lugs, six core stabilizing lugs and a f.low skirt used for directing th~ prim~ry coolant flow into th~ reactor core. __
The RV beltline region has undergone 10 and 20 year inservice inspections conducted in compliance with the requirements of the ASME Code, Section XI, and the recommendations of NRC Regulatory Guide 1.150. The recorded indications were identified as plate segregates and embedded fabrication flaws which are all within the allowable limits of Section XI.
1.10.C Equipment, Components and Structures Affected by Thermal Annealing (TAR Section 1.3)
The-following equipment, structures and components other than the reactor vessel were evaluated to determine if they could be affected either thermally or mechanically by the annealing operation:
Primary coolant system (PCS) loop piping and insulation Steam generators
Reactor vessel insulation Biological shield concrete, liner and insulation 0-Ring leakage monitor tubes Reactor cavity floor Reactor cavity access tube Reactor containment building equipment hatch Instrumentation and cables
The mechanical and thermal conditions imposed on the PCS piping, primary coolant pumps and steam generators will be less severe than that experienced during normal operation except for the RV nozzle extensio!"ls. The high bending stress at the nozzle extension location (nozzle to pipe transition) caused by axial temperature gradients across the RV nozzles has been quantified by the thermal and stress analysis to be well within the ASME Code Case N-557 allowable limits.
The mechanical loading on the RV supports during the annealing will not be significantly different from normal operating loads since the nozzle support pads are free to slide during heatup and cooldown. The temperature at the RV noizle support pads will be higher than the normal operating temperature, however, it will still be within the design conditions.
Insulation on the RV beltline consists of four inch thick reflective insulation panels comprised of 304 stainless steel outer skins enclosing layers of aluminum foil and an inner 304 stainless steel sheet metal divider. This insulation ·will be exposed to tel!lperatur~s_ higher th_an the desig_n temperature of the reactor vessel. Additional oxidation of the insulating materials may occur thereby degrading the reflective -properties and increasing the radiation heat loss. However, the increased surface oxidation will be confined to only three of the nine layers of alum irium foil resulting in a negligible effect for continuing operation of the plant after annealing. The RV insulation will not undergo any structural degradation.
The reactor cavity is surrounded by c:f cylindrical biological shield wall that is composed of 7 to 8 feet of concrete. The concrete is lined with 5/16 inch and % inch thick carbon steel plates on the surface adjacent to the reactor vessel. An approximate 3 foot air gap exists between the reactor vessel and the biological shield wall liner. An embedded cooling system is encased in the first 10 inches of concrete adjacent to the reactor vessel below the RV ·nozzles for the purpose of removing the heat from the biological shield wall concrete during normal operation as well as during the annealing operation. This region of concrete is non-structural and is not taken credit for in the s_tructural support design. Supplemental cooling within the reactor cavity annulus will be provided to assist in removing the added heat load due to annealing in order to maintain the
TAR 3/25/96 1.10-3
f_1 _ biological shield wall concrete_less than_2~0°F. At this temperatur~, the net e~ect on ---- --r-,!--- - -tl'le strer-igtt-l-of--the eoncrete w1ll-be-negllg1ble, -based-on an-evaluation-comparing the----- --- --- -- -
effect of continued strength gain in the concrete due to curing beyond the initial 28-day design cure time to the effect of the reduction in strength from exposure to high temperature and radiation. The biological shield wall concrete will still be capable of performing its design basis function following the anneal.
The temperature of the reactor containment building concrete during the anneal operation will not be affected by the heat loss from the heating system ducting and the heat exchanger assembly since the resulting heat load on the reactor containment building cooling system will be less than during normal plant operation. As a precaution, areas adjacent to the heating system ducting and heat exchanger will be monitored to address any localized heating effects. Supplemental protective measures will be implemented as necessary such as duct insulation or increased standoff distance.
Nuclear instrumentation will be maintained below their environmental temperature qualification limit. Instrumentation within the PCS piping for flow and temperature measurement will be at lower temperatures than during normal operation and will not be adversely affected.
To accommodate the dry annealing operation, the RV upper guide structure (UGS) and . core support barrel (CSB) will be removed and stored with the UGS inside the CSB in the refueling cavity with the cavity drained. This integrated structure will be located near the normal CSB storage area. Temporary lead and steel shielding will be erected to minimize the radiation exposure from this dry storage. The support frame design for this added shielding_ will distribute the shielqing load over the floor area without exceeding the floor load capacity ..
The annealing parameters were established based on Palisades specific materials and their projected recoveries using NUREG/CR-6327 and a 20 and 30 finite element thermal and stress analysis of the process. These annealing parameters have been categorized as either administrative limits or as limiting parameters. Administrative limits are those bounds that have been identified to maintain a prudent engineering margin such that the limiting parameters in Section 1.8 will not be exceeded. Even if
. administrative limits are reached, the analyses in Section 1.7 show margin to limiting parameters exists. When an administrative limit is approached or crossed, action will be taken to return within the limits or an engineering evaluation to justify the deviation will be performed if correction is not feasible. These annealing parameters are also consistent with ASME Code Case· N~557 de·aling with thermal annealing. .
TAR 3/25/96 1.10-4
......
l.\ _ The target inner diameter annealing temperature and time for the Palisades RV anneal ~- ~. - --will-be-85C::l,,900 °F-tor-168-hours~--"fhe-m inimum-annealing-tem perature-and-time-for-a11---- ---- ---
the RV beltline region is 800°F for 48 hours. These minimum conditions will produce adequate fracture toughness recovery.
The maximum reactor vessel temperature will not exceed 940°F nor will the maximum vessel temperature exceed 900°F for greater than 300 hours. These limits are intended to preclude significant creep and other forms of elevated temperature metallurgical degradation. , -
The heatup rate will be an average of ~25°F/hour up to 850°F with a slowdown near the end of the heatup phase. This heatup rate coupled with a slowdown near the end of the heatup allows the thermal energy to distribute more evenly throughout the reactor vessel and PCS, thus reducing the stresses. The heat exchanger-temperature will be controlled to maintain the desired cooldown rate. The cooldown rate of the RV beltline inner diameter surface will be ~25°F/hr down to 210°F, below which no cooldown limit is required.
To minimize the bending stresses in the PCS piping, the axial thermal gradient across the RV nozzles will be procedurally controlled to less than 470°F on the l.D. of the reactor vessel near the end of heatup. The most critical azimuthal thermal gradient in terms of structural integrity is around the RV flow skirt and this will be maintained below
. 50°F. The through-wall thermal gradient in the RV beltline region will be maintained less than 100°F through the end of the heatup period, less than 50°F during the 168 hour hold period and less than 70°F during the cooldown.
1.10.E Annealing Method, Instrumentation and Procedures (TAR Section 1.5)
~- High velocity gases from remote burners will be. directed via ductwork into a multi-zone heat exchanger suspended inside the reactor vessel. The heat exchanger, in turn, delivers the required heat to the RV wall primarily by radiati_on heat transfer. The heating process is controlled to achieve the target annealing temperature in the annealing zone of the reactor vessel. The heat exchanger design and heating process both function to minimize thermal gradients and resultant stresses. The hot gases dd not come in contact with any radioactive surface inside the reactor vessel.
Temporary temperature and displacement sensors will be installed to monitor and control the heating process via pre-approved procedures !n accordance with the limits of the thermal and stress analyses and material recovery requirements. Internal temperature measurements will be taken by direct contacnemperature sensors placed at 60 locations on the RV inside diameter, 2 locations inside each inlet and outlet
TAR 3/25/96 1.10-5
.,
• _J~--nozzle for a total of 72 locations. External sensors will be installed with the use of remote tooling to-gain access-to·the-3/4 inch gap-between-the RV 0:-D: wall and the RV· insulation. External temperature sensors will be placed at 17 locations, including on the biological shield wall liner and on a RV support structure. Direct contact linear displacement sensors will be used to monitor equipment displacement at 7 external locations on the reactor vessel bottom and PCS piping. Redundant sensors will be placed at critical locations so that loss of a single signal will not have an adverse effect on the process control. Redundancy of information will also be achieved by placing sensors so as to take advantage of geometric symmetry. Calibration and accuracy requirements, sa.mpling frequency and data acquisition methods are defined for each temperature and displacement sensor in order to support the goals of the monitoring . program.
Utilizing the sensor configuration described above, the goals of the monitoring program are to:
• document the time temperature history of the RV wall within the annealing zone to confirm reactor vessel material recovery.
• measure the RV axial and/or circumferential temperature gradients and associated temperatures at the flange, nozzles, shell and the flow skirt region for acceptable temperatures and resulting stresses and strains.
• measure the displacement of the RV bottom and PCS piping, and PCS piping temperatures for acceptable resulting stresses and strains and equipment support boundary restraints.
. . • measure the temperature of the RV support structure to monitor the.RV
support· structure temperature and the concrete temperature in the vicinity of the supports.
• measure the temperature of the reactor cavity biological shield wall liner to monitor maximum concrete temperatures to ensure the strength of the concrete is retained.
A summary of the planned sequence for the annealing operation is presented below:
-•- Assemble the heat exchanger inside containment • Install external instrumentation on the reactor vessel and PCS· piping • Defuel the reactor • Remove the CSB and place into its storage location • Construct the temporary shielding for the RV internals • Air lift the UGS and place into the CSB
TAR 3/25/96 1.10-6
• -1: .. ~ ... • Perform pre-anneal inspections •- Brain-the refl:Jeling cavity to-the RV-flange level- - - -· - - - - -- -- - - -- --- - ----- ---• Lower the heat exchanger into the reactor vessel while lowering the water level • . Connect duct sections to the heat exchanger • Perform a low temperature dry out of the reactor vessel • Heat reactor vessel to achieve annealing temperature • Maintain temperature and soak for annealing period • Cooldown reactor vessel using heat exchanger for heat removal • Remove heat exchanger while flooding reactor vessel • Flood refueling cavity • Perform post-anneal inspections • Dismantle and remove temporary shielding structure from refueling cavity • Remove UGS from CSB • Perform fit up tests and install wall surveillance capsule holders • ·Move the CSB into the reactor vessel and perform fit up test • Refuel the reactor vessel
A method of high velocity combustion technology will be used to deliver the required heat to the reactor vessel via insulated hot air ducts. Radiant heat transfer from a heat exchanger placed inside the- reactor vessel will be utilized. The gases are contained within the heat exchanger and ductwork and do not come in contact with. the RV.
The multi-zoned heat exchanger is comprised of 5 axial chambers designed to provide _a homogenous temp~rature distributionwithin the reactor vessel beltline and to control
axial temperature gradients beyond the annealing band. Since the fully assembled heat exchanger unit will not fit through the Palisades equipment hatch, the heat exchanger will be assembled inside containment. The heat exchanger is shown in Figure 1.6.A.1-2. Ductwork co-nnecting the heat exchanger inlet with the burners and providing the outlet exhaust path from the heat exchanger. to outside containment will be 1 O inch diameter or greater insulated stainless steel pipe. ·
The heat will be supplied by propane fired burners and blowers located outside containment. There is one complete. gas train for each heat exchanger zone, plus an available spare. The fuel source will be liquid propane delivered by road tankers, each
--- --- - -- --- -with a-cai;,acity-of up to 19,000-gallons.- --A-f1Jel-docking -station will provide docking-for-- - -- -- --- ---up to two road tankers.
Instrumentation and controls are supplied to address equipment operation, to avoid adverse effects from undesirable therm·a1 gradients, and for process safety. The combustion control system is comprised of appropriate safety devices in accordance
TAR 3/25/96 1.10-7
with industry standards such as the American Gas Association Code and the National Fire Protection Association. The safety features initiate automatic fuel shutoff for the following conditions:
• loss of flame sensed by an ultraviolet flame sensing eye
• low combustion air pressure
• high and low fuel pressure
The control trains will be manned 24 hours per day after the burners are ignited. In addition, periodic "job walks" will be conducted to verify the heating system is operating safely. The control train will be operated in a high excess air condition to prevent gas concentrations which are flammable or explosive from developing in the heat exchanger or piping.
1.10.G Thermal and Stress Analysis (TAR Section 1. 7)
Detailed Palisades-specific thermal and stress analyses were performed to demonstrate the acceptability of the annealing operation. The AN SYS computer program was used to develop and run two dimensional (20) and three dimensional (30) finite element models. . .
The 20 model was used to identify a heating distribution that met the annealing requirements and to perform parametric studies of variations in h~at exchanger and RV emissivities, heat exchanger design configurations, RV wall thickness, heatup rates, material properties of the cladding, steam generator friction effects and e~uipm_eot boundary conditions.- It was also used to determine the effects of residual cladding interfacial stresses and to quantify the displacements, the residual deformations, and the effects of creep. The 30 model was used to quantify the thermal loadings, the thermal stress gradients, and displacements. The 30 model consisted of 15,87 4 nodes and 3,068 elements. Both the 20 and 30 analyses were conducted considering conservative conditions that would bound the actual annealing operation. Extreme bounding cases were imposed on the 30 analysis model to develop permissible thermal gradients and temperature profiles.
An analysis was also conducted to show the effect of convection above the heat
_.,,:
exchanger. This effect was determined to be n~gligiqle compared to radiation and- - -··- - ·---·- · conduction. - -- - - - -- - ·
The analyses showed that the maximum peak stress intensity anywhere in the reactor vessel is at a very localized area on the RV nozzle support pad and could be as high as 59 ksi. This stress is caused by the assumption that the nozzle support does not
TAR 3/25/96 1.10-8 ' . ..; '-~. . '
rotate. By design, the nozzle support does rotate which eliminates the localized high stress in the nozzle support pad. The RV outlet nozzle transition has a maximum calculated sfress of 47 ksi and the RV outlet nozzle extension has a maximum calculated stress of 43 ksi. These bounding values satisfy the Code Case N-557 allowable limits. Primary and secondary stresses elsewhere are all well below these values and thus fall within the Code Case N-557 allowable limit.
The maximum radial displacement of the reactor vessel could be as high as 0. 70 inch in the beltline region. The maximum radial displa,cement of the steam generator may be up to 1.12 inch. Both values are within acceptable limits.
The maximum vertical displacement in the reactor vessel could be up to 2.23 inches and occurs at the bottom head. This value is approximately 1.25 inches more than the thermal growth experienced during the normal plant heatup. The worst case effect is that the block insulation below the reactor vessel will be crushed and the stainless steel floor covering will be ·deformed up to 1.25 inches. This condition can be tolerated without affecting design basis conditions.
The peak residual deformation is 0.008 inch immediately below the RV nozzle area. This value and smaller residual deformations elsewhere in the reactor vessel are within the required clearances for the reactor internals and tolerances for the reactor vessel head and internal seating surfaces.
The largest total stress occurs at the end of the heatup peri.od and is within the stress allowables. The. maximum stress occurs just below the RV nozzles. The largest strains due to creep occur at the beginning of the cooldown and were verified to be negligible.
Limiting parameters are identified as the governing parameters to assure an effective and acceptable anneal of the Palisades reactor vessel without further analysis and testing. Maintaining the process within these limits establishes the basis for certification that the annealing operation was conducted in accordance with the approved analyses that have been documented in the TAR. These limiting parameters are also consistent with ASME Code Case N-557 dealing with'thermal annealing.
Adequate material property recovery can be achieved with an annealing temperature of 800°F for 48 hours whic_h ',,Vi~l_pe_~mi_t ~_ontin_ued ~~f~_qpe_ration of the Palisades.reactor.
- - -- - vessello afleasfMarcti 2011. However, the target time and temperature to be achieved for the inner diameter of the RV beltline will be one week (168 hours) at 850-9000F. To avoid elevated temperature metallurgical degradation such as significant creep, the reactor vessel cannot exceed 900°F for greater than 300 hours, 850°F for
TAR 3/25/96 1.10-9
greater than 1000 hours, or maximum temperature of 940°F. In addition, a maximum temperature limit for the biological shield wall concrete has been established as 250°F.
Criteria for the reactor vessel and PCS piping has been established to maintain the primary and secondary stresses below the 3Sm limit. These limits· are intended to avoid harmful permanent set and the potential for ductile flaw growth.
Limiting steam generator radial displacement criteria has been established to prevent the steam generators from hitting hard stops as a result of horizontal sliding movement. The limits are 1.99 inches for steam generator A and 1.41 inches for steam generator B.
1.10.1 ALARA Considerations (TAR Section 1.9)
As low as reasonably achievable (ALARA) dose rates to personnel working on the annealing project will be attained by applying such principles as:
• designing tooling and equipment to facilitate ease of assembly, disassembly and repair
• use of long-handled underwater tools
• providing special shielding where appropriate, for example surrounding the RV internals while in dry storage
• assembling as much of the annealing equipment outside of containment as . possible, and
• using remote tooling to install external temperature measurement devices
Personnel performing major site activities in support of the annealing project will undergo training and, when appropriate, include the use of qualified mock-ups on equipment and tooling with which they will be working. Industry e?<perience gained from the annealing demonstration project at Marble Hill will be applied to Palisades. Experienced personnel familiar with performing work in radiation areas and implementing work practices to minimize radiation exposure will be utilized as much as possible.
Exposure to personnel who must be in the area during movement and storage of the RV internals will be minimized by use of temporary shielding to protect against the effects of streaming. Airborne radioactivity will also be controlled during the annealing process through the use of HEPA filters.· The design of the reactor vessel top cover
TAR 3/25/96 1.10-10
•
•
(RVTC) precludes escape of airborne activity into containment from the primary system during the annealing operation. ·
It is the intent to remove virtually all equipment brought to the Palisades site for the annealing program at the completion of the effort. Either the equipment will be free released because it never entered a contaminated area, or it will be decontaminated on site to free release or low specific activity (LSA) levels. Equipment requiring decontamination includes the heat exchanger, ducting, internals storage and shielding material, and temporary instrumentation. Wherever possible, this equipment will be made of non-porous material and/or coated to ease the decontamination effort.
Initial anneali_ng projected dose estimates are approximately 200 rem excluding other outage scope tasks and support.
1.10.J Conclusion
The Thermal Annealing Operating Plan will ensure that the annealing process is adequately controlled and monitored in order to maintain the reactor vessel safe to operate following the anneal. It will also ensure that the reactor vessel material properties will support operation through the end of the plant's operating license .
TAR 3/25/96 1. 10-11
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•
•
ATTACHMENT 3
CONSUMERS POWER COMPANY PALISADES PLANT
DOCKET 50-255
THERMAL ANNEALING REPORT
DESCRIPTION OF ERRATA MADE TO. OTHER SECTIONS
BASED ON SECTION 1. 7 TECHNICAL REVIEW
7 Pages
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Section 1.2
Paragraph 1.2.E.1.1 Nozzle Extensions
Changed second paragraph, last sentence to read:
The loadings at the nozzle extensions due to thermal annealing are developed from the thermal and structural analyses in Section 1. T and are justified by either comparison with the design loadings or by stress analysis in accordance with ASME Code Case N-557.
Paragraph 1.2.E.1.2 0-ring Leakage Monitor Tubes
Changed second paragraph to read:
The effects of the thermal annealing on the monitor tubes are loads due to the thermal displacements of the reactor vessel and interaction with the attached piping system. Resolution of the loads is not expected to be an issue since the monitor tubes and attached piping are expected to have sufficient flexibility to react to the imposed loads. The thermal displacements of the reactor vessel; particularly the reactor vessel flange, due to the annealing are established in the thermal and structural analysis in Section 1. 7 and are used to evaluate the monitor tubes and attached piping in accordance with Section Ill of the ASME .Boiler and Pressure Vessel Code.
Piiragraph 1.2..E.2.1. . Cor~ Supporj Lugs and .Core Stabilizi11g Lugs
Changed second paragraph to read:
Six (6) core stabilizing lugs of SB-166 Ni-Cr-Fe Alloy 600 bar are weld attached to the interior wall at the bottom section of the lower shell course. The lugs are spaced around the reactor vessel for the purpose of provid,ing lateral support for the bottom of the core support barrel. At each lug, four Ni-Cr-Fe Alloy X-750 bolts are used to clamp custom machined Ni-Cr-Fe Alloy X-750 Shim pads to the lugs. The bolts are trapped by small dowel pins. The shims were machined to provide the required clearance fit with mating pads on the as-bui.ld core support
-._.-barrel (CSB) based on-field measurements. In -the original design there was no · consideration of a differential thermal loading between the lugs and the bolts that . could be caused by an annealing process because normal operation produces an isothermal situation and the thermal coefficients are essentially identical. Section 1. 7 addresses this consideration .
1
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•
•
Changed third paragraph as follows:
Changed "Ni-Cr-Fe Alloy 600 material" to "Ni-Cr-Fe material" at three places.
Changed fourth paragraph to read:
The core stabilizing lugs are not included in the thermal model in Section 1. 7 since the coefficients of thermal expansion for the Ni"'."Cr-Fe Alloy 600 lugs and the carbon manganese molybdenum steel reactor vessel shell material are very similar throughout the temperature range and the average annealing heatup and cooldown rates are low. The thermal stresses in the core stabilizing lugs are small and comparable with those of the core support lugs. The thermal stresses in the core support lugs at the RV wall attachment due to annealing are evaluated in the RV stress analysis in Section 1.7, since they are attached to the flow skirt.
There are six (6) surveillance holder tubes mounted vertically on the inside wall of the reactor vessel below the primary nozzles for the purpose of positioning surveillance capsules on the interior wall of the core region shell. The surveillance holders are attached to the wall by one hundred and eight SB-166 Ni-Cr-Fe Alloy 600 bracket supports which are weld attached to the RV wall. Each bracket is fillet welded to a SB-166 Ni-Cr-Fe Alloy 600 rectangular box beam that is 135 inches long and 0.12 inch thick. There are ·eighteen brackets per box beam, nine on each side.
Changed second paragraph to read:
The effect of the thermal annealing temperatures on the surveillance holders is thermal stress in the bracket welds at the RV wall interface, in the fillet welds at the bracket box beam interface, and in the box beam. Since the coefficients of thermal expansion for the carbon manganese molybdenum steel vessel shell material and the Ni-Cr-Fe Alloy 600 support bracket material are very similar over the entire annealing temperature range, the thermal stress at the RV wall -interface during annealing will be acceptably small. Even though the support brackets are welded to the austenitic stainless steel cladding, the base metal dictates the thermal expansion through the bond with the thin cladding. The thermal stresses for the bracket welds at the RVwall interface, the fillet welds at the bracket to box beam interface, and the box beam are evaluated in Section 1.7 .
2
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Section 1.3
Paragraph 1.3.B.1 Primary Coolant System Piping
Changed the third paragraph to read:
The thermal deflections on the PCS loop piping are developed from the analytical results in Section 1.7, Thermal and Stress Analysis. The maximum stresses occur at the· RV nozzle connections and are justified by stress analysis in accordance with the ASME Code Case N-557.
Paragraph 1.3.B.8 Biological Shield Insulation
Changed the third paragraph, second sentence, to read:
The RV thermal growth below the RV supports during annealing will be up to 1.25 inches greater than the thermal growth during the normal plant heatup.
Changed the third paragraph, sixth sentence, to read:
Therefore, the worst case effect is that the 4-inch thickness of the block insulation will be crushed and the stainless steel floor covering ·will be deformed up to 1.25 inches.
Changed the first paragraph, last sentence, to read:
These thermal deflections are within the capabilities of the leakoff piping .
3
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Section 1.4
Paragraph 1.4.D Temperature Gradients
Changed the first paragraph as follows: After the seventh sentence added the following sentence:
The largest stresses due to the axial temperature gradient will happen at the end of heatup when the RV beltline region is reaching its prescribed temperature of approximately 900°F but the flange is still considerably cooler.
Changed the last two sentences to the following~·
From the bounding thermal and stress analyses described in Section 1. 7 it was determined that an axial temperature gradient across the RV nozzles of~ 470°F near the end of heatup (administrative limit) between monitoring zones Land M as defined in Section 1.5 could be tolerated with acceptable margins to the Section 1.8 limiting stress parameter. The analyses predicting the expected conditions, however, predict the temperature gradient wi.11 be approximately 280°F near the end of heatup .
Table 1.4.A-:1 under RV Axial Temperature Gradient
Under "Limit" column, changed to read:
~470°F near the end of heatup .
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Section 1.6
Section 1.6.A.1.1 Heat Exchanger
Changed the last paragraph as follows:
There are no specific code requirements which directly apply to the heat exchanger and ducting design. In the absence of a directly applicable construction code, the design of the heat exchanger will generally be in accordance with ASME Ill, Code Case N-47-33. (Note that there is now a newly issued ASME Ill Subsection NH, Class 1 Components in Elevated Temperature Service, 1995 Edition, which is essentially identical to ASME Ill Code Case N-47-33.) The fabrication of the heat exchanger will generally be in accordance with ASME Section VIII, Division 1. However, ASME Section VIII, Division 1 code requirements for certain weld joint configurations and examinations, third party inspection, data reports and stamping will not apply. '
Section 1.6.A.1.2 Ducting
In the fourth paragraph, first sentence, changed:
"ANSI 831.3" to "ASME 831.3"
Changed fifth paragraph to. read as follows:
The ducting will be brought into containment in sections. Joints between sections will be made with mechanical couplings, bolted flanges, or they may be entirely welded. A combination of coupling types may be employed since the design and assembly requirements differ depending on where the joints are being made. For example, ALARA concerns may result in the use of mechanical couplings at locations in the vicinity of the cavity, while bolted flanges or welded connections may be desirable overall. The type of connections used will be analyzed and justified as part of the design analysis report generated for the overall piping system .
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Section 1.6.A.3 Controls and Instrumentation Including Redundancy
Changed first paragraph, first two sentences, as follows:
Section 1.5,B. contains a qescription of the method of instrumentation used during the annealing, that which is directly applicable to control of the annealing process, arid describes the redundancy of measurements. Section 1.5.C. describes the instrumentation accuracy and reliability. Section 2.1 describes the overall monitoring contingency approach.
Section 1.6.A.5.3 Reactor Vessel Drainage/Purge System
Section 1.6.A.3, Controls and Instrumentation Including Redundancy . . . . . . . . . 1.6-17
Section 1.6.A.5.3, Reactor Vessel Drainage I Purge System ................ 1.6-19
Section 2.1, Monitoring the Annealing Process Table 2.1.A-3 under Temperature Gradient, Axial ......................... 2.1-6
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thermal loadings on the nozzles extensions associated with annealing are enveloped by the design loads even though the maximum annealing temperature is higher. However, the maximum bending moment loadings at the RV nozzle extensions exceed the design thermal loadings. The loadings at the nozzle extensions due to thermal annealing are developed from the thermal and structural analyses in Section 1. 7 and are justified by either comparison with the design loadings or by stress analysis in accordance with ASME Code Case N-557.
The primary coolant system piping is discussed in Sections 1.3.A.1 and 1.3.B.1.
1.2.E.1.2 0-ring Leakage Monitor Tubes
The 3/4 inch schedule 80 0-ring leakage monitor tubes at the vessel flange are connected to a piping system which pipes any leakage to the drain tank. The monitor tubes are ASME SB-166 Ni-Cr-Fe Alloy 600 and are welded into the RV flange penetrations by Ni-Cr-Fe alloy partial penetration welds at the flange mating surface. The monitor tubes exit the penetrations on the exterior of the RV flange 16° 40' to either side of the 0° axis.
The effects of the thermal annealing on the monitor tubes are loads due to the thermal displacements of the reactor vessel and interaction with the attached piping system. Resolution of the loads is not expected to be an issue since the monitor tubes and attached piping are expected to have sufficient flexibility to react to the imposed loads. The thermal displacements of the reactor vessel, particularly the reactor vessel flange, due to the annealing are established in the thermal and structural analysis in Section 1. 7 and are used to evaluate the monitor tubes and attached piping in accordance with Section Ill of the ASME Boiler and Pressure Vessel Code.
The attached leakoff piping system.is discussed in Sections 1.3.A.9 and 1.3.B.9.
1.2.E.1.3 Reactor Vessel' Support' Pads ·
The reactor vessel is supported by the RV support pads on the underside of three (3) of the RV nozzles 120 degrees apart. The support pads, which are bolted to a sole plates and rest on the plates of the RV supports, can slide to accommodate the thermal growth at the support pads as the reactor vessel heats up and cools down. The support pads are SA-508, Class 2 low alloy steel forgings which are welded to low alloy steel weld build-up preparations on the nozzles. The bearing surface on the bottom of each support pad is 18 inches x 48 inches.
The nozzle support pads will be subjected to temperatures which are less than the RV ... qe$ign temperature _of 650°F quring the thermal annealing _operatio_n!?. Furthermore, the
loading of the pad bearing surfaces during annealing will be less than the design loading since the RV head, reactor internals, fuel and water volume will not .be supported by the reactor vessel during annealing. Therefore, there are no detrimental effects of thermal annealing on the support pads .
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· the interior wall at the bottom section of the lower shell course. The lugs are spaced around the reactor vessel for the purpose of providing lateral support for the bottom of the core support barrel. At each lug, four Ni-Cr-Fe Alloy X-750 bolts are used to clamp custom machined Ni-Cr-Fe Alloy X-750 shim pads to the lugs. The bolts are trapped by small dowel pins. The shims were machined to provide the required clearance fit with· mating pads on the as-build core support barrel (CSB) based on field measurements. In the original design there was no consideration of a differential thermal loading between the lugs and the bolts that could be caused by an annealing process because normal operation produces an isothermal situation and the thermal coefficients are · essentially identical. Section 1. 7 addresses this consideration.
The 900°F maximum annealing temperature is well below the heat treatment temperature of the Ni-Cr-Fe material. Therefore, no degradation of the mechanical properties due to the heating is anticipated. There will be a small degree of grain boundary sensitization when holding the Ni-Cr-Fe material at 900°F for 168 hours. I However, sensitized Ni-Cr-Fe material is not a concern in the normal PWR · 1 environment. . ·
The core stabilizing lugs are not included in the thermal model in Section 1. 7 since the coefficients of thermal expansion for the Ni-Cr-Fe Alloy 600 lugs and the carbon manganese molybdenum steel reactor vessel.shell material are very similar throughout the temperature range and the average annealing heatup and cooldown rates are low.· The thermal stresses in the core stabilizing lugs are small and comparable with those of the core support lugs. The thermal stresses in the core support lugs at the RV wall attachment due to annealing are evaluated in the RV stress analysis in Seetion 1.7, since they are attached to the flow skirt .
1.2.E.2.2 Flow Skirt
The flow skirt is a welded assembly of SB-168 Ni-Cr-Fe Alloy 600.plate in the bottom head region of-the Palisades reactor vessel for the purpose of directing the primary coolant flow into the reactor core. The flow skirt is supported in the bottom head by the nine core support lugs by attachment welds which were performed at the site.
The main effect of the thermal annealing operations on the flow skirt are expected to be stresses at the attachment points resulting from the thermal displacements. The thermal stresses in the flow skirt are evaluated in the reactor vessel stress analysis in Section 1. 7.
The SB-168 Ni-Cr-Fe Alloy 600 material will be slightly sensitized at temperature during annealing. However, the sensitized material is not a concern in the normal PWR environment.
1.2.E.2.3 Surveillance Holder Assemblies
There are six (6) surveillance holder tubes mounted vertically on the inside wall of the . reactor vessel below the primary nozzles for the purpose of positioning surveillance capsules on the interior wall of the core region shell. The surveillance holders are attached to the wall by one hundred and eight SB-166 Ni-Cr-Fe Alloy 600 bracket
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supports which are weld attached to the RV wall. Each bracket is fillet welded to a I SB-166 Ni-Cr-Fe Alloy 600 rectangular box beam that is 135 inches long and 0.12 inch I thick. There are eighteen brackets per box beam, nine on each side. . I
I The effect of the thermal annealing temperatures on the surveillance holders is thermal stress in the bracket welds at the RV wall. interface, in the fillet welds at the bracket box I beam interface, and in the box beam. Since the coefficients of thermal expansion for ·. I the carbon manganese molybdenum steel vessel shell material and .the Ni-Cr-Fe Alloy 600 support bracket material are very similar over the entire annealing temperature I range, the thermal stress at the RV wall interface during annealing will be acceptably small. Even though the support brackets are welded to the austenitic stainless steel cladding, the base metal dictates the thermal expansion through the bond with the thin cladding. The thermal stresses for the bracket welds at the RV wall interface, the fillet welds at the bracket to box beam interface, and the box beam are evalua.ted in Section 1.7
The Ni-Cr-Fe Alloy 600 bracket material may be slightly sensitized at temperature during annealing. However, the sensitized material is not a concern in the normal PWR environment.
1.2.E.3 Other Equipment, Components, Structures and Instrumentation
Equipment, components, structures and instrumentation in and around the reactor cavity other than RV attachments are described along with tl:le expected effects of the thermal annealing in Section 1.3. In addition to the previously identified sections addressing the nozzle extensions, the RV support pads, the leakage monitor tube leakoff piping, and the seal ledge, Section 1 ;3 also includes discussions of the effects of thermal annealing on RV insulation, primary coolant system piping, RV sliding supports, cavity seal drip pan and drain lines, primary coolant piping insulation, the biological shield and the biological shield insulation. Moreover, Section 1.3.E addresses the effects of the heat' of ·annealing on instrumentation such as ne.utron detectors and the primary coolant flow and temperature measurement instrumentation.
1.2.F REFERENCES
Combustion Engineering, "Palisades General Reactor Vessel Arrangement", Drawing Number SE-2005507-322-001, Revision 2. Windsor, CT
Fenech, R. A. 1994, "Docket 50-255 License DPR-20, Palisades Plant - Response to the January 9, 1994 NRC Request for Additional Information - 1 O CFR 50.61 Screening Criterion", Washington, D. C.
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Peter, P.A., Lippincott, E.P., Wrights, G.N., Madeyski, A., 1994, "Analysis_ of Capsule W-110 form the Consumers Power Company Palisades Reactor Vessel Radiation Surveillance Program", WCAP-14014, Pittsburgh, PA.
Porter, N. J. (CE), 1984a, "(Palisades Plant) Reactor Vessel Weld Documentation," PCE-7747, letter to J. B. Taskey (CPCo), Windsor, CT
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Porter, N. J. (CE), 1984b, "Palisades Vessel Weld Documentation," P-CE-7752, letter to J. B. Taskey (CPCo), Covert, Ml ·
RSIC Data Library Collection DLC-76, 1987, "SAILOR, Coupled, Self-Shielded, 47-Neutron, 20-Gamma-Ray, P3, Cross-Section Library for Light Water Reactors".
Section Ill of the ASME Boiler and Pressure Vessel Code, Appendix G "Protection Against Nonductile Failure".
Wrights, G. N., 1994, "Palisades Cycle 11 Reactor Vessel Inner Radius Fast Neutron Flux Data", Westinghouse Letter SE/REA-169/94 to R. Snuggerud, Pittsburgh, PA.
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seal. The drip pan is not connected to the RV seal ledge and does not, therefore, interfere with vertical and radial displacements due to the thermal expansion of the reactor vessel.
Two 2-inch diameter stainless steel drain lines extend down the reactor cavity annulus from the drip pan to the reactor cavity floor where each runs into the floor drain recess. The line on the opposite side of the reactor vessel from the floor drain makes a 90° tum at the floor and is held off the floor by small stanchion supports as the pipe traverses the floor to the floor drain.
1.3.A.12 Reactor Cavity Floor Cooling Coils
A bank of cooling coils approximately 15 feet square is located under the reactor vessel beneath the cavity floor insulation. The purpose of the cooling coils is to limit the cavity floor concrete temperature by supplying cooling water flow.
1.3.8 Effects on Equipment, Components and Structures
The effects on the above discussed equipment, components and structures as a result of annealing are presented in the following sections.
1.3.8.1 Primary Coolant System Piping
The PCS piping, at the attachments to the RV nozzles, will be affected· by the heat transfer from the reactor vessel during the annealing operation. The degree of the thermal loading is dependent on the temperature profile of the RV nozzles and the piping. With the thermal barriers in place during annealing the temperature of the pipe will remain well below the range where changes in the carbon steel material properties would occur. The pipe metal temperature also remains below the creep range. The temperatures are quantified in the thermal analysis in Section 1.7.
The PCS piping will also be loaded during the annealing heat up and cooldown due to thermal expansion and contraction of the reactor vessel and the piping itself. The compressive thermal expansion loads on the piping during heatup arid tensile loads during· cooldown remain less than the design thermal loads since the thermal expansion due to the local heating of the reactor vessel during annealing is less than the thermal expansion due to heat up of the entire PCS during normal operation. However, the bending moments at the RV nozzle connections due to the axial thermal gradient in the RV upper (nozzle) shell during annealing exceed the thermal bending moments during normal operation.
The thermal deflections on the PCS loop piping are developed from the analytical results in Section 1.7, Thermal and Stress Analyses. The maximum stresses occur at the RV nozzle connections and are justified by stress analysis in accordance with the ASME Code Case N-557.
1.3.8.2 Steam Generators
The steam generators should not be heated up during the annealing operation because of their distance from the reactor vessel. The temperature will remain well below normal operating temperature and there will be no significant thermal gradients at the steam generators. The steam generator inlet nozzles will be
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The floor insulation beneath the recess under the RV bottom head will probably be subjected to a mechanical _loading due to the greater vertical thermal growth of the reactor vessel during thermal annealing. The RV thermal growth below the RV supports during annealing will be up to 1.25 inches greater than the thermal growth during the normal plant heatup. Without sufficient clearance between the RV bottom head and the cavity floor to accommodate the increased thermal growth, the floor insulation will be loaded in compression by contact with the reactor vessel. Unibestos insulation has a compressive strength of only 1600 lb/ft2 (11 psi). This compressive strength is comparable to that of soil and is only one four-thousandth that of steel. Therefore, the worst case effect is· that the 4-inch thickness of the block insulation will be crushed and the stainless steel floor covering will be deformed up to 1.25 inches. However, the cooling coils will be protected by the crushing of the insulation and the resistance of the steel plate over the coils. The crushed insulation will cause a local reduction in the effectiveness of the floor insulation due to the decrease in thickness. However, the heat capacity of the cooling coils will accommodate the IQcal hot spot.
Additional insulation around the lower section of the reactor vessel, with the opening of the horizontal hinged panels and the use of supplemental cooling will reduce the temperatures at the cavity wall based upon the thermal analysis in Section 1.3.C. With supplemental cooling air flow through the hinged panels, the temperatures of the horizontal panels and the cavity wall insulation can be maintained within the normal operating range. Mineral wool blanket insulation is capable of effectively withstanding continuous temperatures up to 1000°F without deterioration. Therefore, the temperatures during both normal operation and annealing are not considered detrimental to the insulation.
The RV .bottom head will likely be in contact with the floor insulation and sheet metal covering of the reactor cavity during the thermal annealing operations .. However, since the Unibestos insulation is capable of withstanding continuous contact with hot surfaces up to 1200°F without deterioration, this is an acceptable condition.
·The-insulated cover over the access tube opening in the reactor cavity will be,removed for the thermal annealing.
There will be thermal pipe reactions at the 0-ring leakoff piping conneCtions due to the thermal expansion of the reactor vessel at the RV flange elevation. The effect of the thermal loads at
·the piping connections are discussed in more detail in Section 1.2.E.1. The thermal deflections at the RV flange are identified in Section 1.7. These thermal deflections are within the capabilities of the leakoff piping.
The valves and the pressure and level instrumentation outside the biological shield will not be ----- - -subjected to the effects of the elevated annealing temperatures.
• 1.3.B.10 Biological Shield Liner
The carbon steel biological shield liner plates are affected by the convection and radiation heat transfer across the annulus between the insulation and the liner. The
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the reactor vessel and PCS piping to reach a safe temperature for termination of the annealing operation. Thus cooling of the Palisades reactor vessel involves a continuous decrease in the heat exchanger temperature, with a resultant decrease in the RV temperature by a thermal radiation process. The heat exchanger temperature will be controlled to maintain the desired cooldown rate. The cooldown rate of the RV beltline inner diameter surface will be ~25°F/hr (administrative limit) down to 210°F. After 210°F has been reached the stresses are sufficiently low to not cause further residual stresses therefore a controlled cooldown limit beyond 210°F is not necessary.
1.4.D Temperature Gradients
Section 1. 7 indicates that the stresses created by the annealing process are dictated by the temperature distribution or gradients within the reaCtor vessel. The most critical temperature gradient is that running axially across the Palisades RV nozzles. · This axial temperature gradient results in variable radial thermal displacements along the length of the reactor vessel cylindrical wall. This axial deformation has been identified as the "Coke Bottle" effect and was seen in a past study (Server, 1985) as well as in the analyses documented in Section 1.7. The variable axial deformation results in the bending of the RV cylindrical shell. This bending causes higher stresses at the upper shell to intermediate shell transition and to a lesser degree at the intermediate shell to bottom head transition. The most significant effect of the RV rotation is the production of a bending of the PCS piping particularly at the RV nozzle to pipe transition (nozzle extension). The largest stresses due to the axial temperature gradient will happen at the end of heatup when the RV beltline region is reaching its prescribed temperature of approximately 900 D F but the flange is still considerably cooler. For the Palisades RV annealing the "coke bottle" effect is less pronounced due to the larger heating zone than was used in the previous studies, however the bending is still present. The means of minimizing the bending stresses is by minimizing the axial temperature gradient across the RV nozzles. From the bounding thermal and stress analyses described in Section 1. 7 it was determined that an axial temperature gradient across the RV nozzles of ~470°F near the end of heatu·p (administrative limit) between zones Land M, as defined in Section 1.5 could be tolerated with acceptable margins to the Section 1.8 limiting stress parameter. The analyses predicting the expected conditions, however, predict the temperature gradient will be approximately 280°F near the end of heatup.
Azimuthal temperature variations are expected to arise from variability in the heat exchanger wall temperatures and from variations in the effectiveness of the RV insulation. These are expected to be small since the heat exchanger, by the nature of its design, has shown during proof of principle testing to produce uniform heat circumferentially. Also, the RV insulation performance was measured at several locations following the fall 1995 outage and was found to be uniform circumferentially. The most critical azimuthal gradient in terms of structural integrity is that around the RV flow skirt. This region has been identified since the thin perforated plate of the RV flow skirt is not as stiff as the RV shell to which it is attached, and as such could undergo significant permanent deformation ifthis azimuthal temperature gradient is not controlled. Other regions in the reactor vessel being thick solid sections can accommodate larger azimuthal temperature gradients. As described in Section 1.7, an azimuthal temperature variation at the
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• • • Parameter Limit1 Comments
RV Heatup Rate Avg. ~25°F/hr heatup rate from Based on RV beltline ID temperatures ambient to 850 ° F
RV Heatup Rate 40°F/hr maximum heatup rate-from Based on RV beltline ID temperatures ambient to 600°F
RV Time at Temperature Hold at 850°F to 900°F for 168 Based on RV beltline ID temperatures hours
RV Beltline Through-Wall ~100°F/hr Heatup Based on RV beltline ID and OD Temperature Gradient ~50°F/hr Hold temperatures
' ~70°F/hr Cooldown
RV Azimuthal Temperature Gradient ~50°F Based on maximum to minimum OD temperatures near the flow skirt elevation
RV Axial Temperature Gradient ~470°F near the end of heatup Nozzle region temperatures on RV ID, between Zones Land Mas defined in Section 2.1
RV Bottom Displacement Displac~ment range<2> If the cold clearance at the RV bottom
is greater than 2 inches then this limit will be eliminated -
Steam Generator Displacement ~1.375 inches Based on PCS loop piping radial displacement monitoring
Table 1.4.A-1 'Administrative Limits For the Palisades Reactor Vessel Anneal (Page 1 of 2)
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exchanger joints fashioned during assembly inside containment will be structural joints, meaning that they will be required to transmit pressure and/or dead loads. In addition to leak tightness and structural integrity, joint design must consider factors such as ease of assembly and disassembly, schedule impact, and ALARA. Bolted and gasketed mechanical connections will be used wherever they prove acceptable. For the heat exchanger, all seal joints will require seal welding to be performed during the assembly process. As a result, overall joint design will combine the berJefits of bolted connections (rapid assembly and strength) with the benefits of seal welding (leak tightness).
Based on the analysis described in Section 1.7, the peak required heat loads, zone inlet temperatures, and flow rates were estimated for each zone of the heat exchanger. The peak requirements are not expected to exceed those shown in Table 1.6.A.1-2. Based on the expected operating temperatures, the exchanger will need to be constructed from stainless steel. Grade 304 stainless will be used for the majority of the exchanger, although certain higher temperature components may be constructed from Grade 310.
There are _no specific code requirements which directly apply to the heat exchanger and ducting design. In the absence of a directly applicable construction code, the design of the heat exchanger will generally be in accordance with ASME Ill, Code Case N-47-33. (Note that there is now a newly issued ASME Ill Subsection NH, Class 1 Components in Elevated Temperature Service, 1995 Edition, which is essentially identical to ASME Ill Code Case N-47-33.) The fabrication of the heat exchanger will generally be in accordance with ASME Section VIII, Division 1. However, ASME Section VIII, Division 1 code requirements for certain weld joint configurations and examinations, third party inspection, data reports and stamping will not apply.
1.6.A.1.2 Ducting
The inlet ducting will c_arry th!3 hot air.and products of combu_stion from the burners to the heat exchanger, and the outlet ducting will return those gasses to the outside environment. The routing of the inlet and outlet ducts inside containment is described in Section 1.6.A.2. Based on this routing and on the performance parameters described above, 1 O inch diameter duct will be used. If, as a result of the final heat exchanger design or slight variations in the duct routing, flow losses become excessive, then larger duct sizes can be accommodated.
The selection of ducting material is dependent on the peak operating temperature of the duct. Grade 304 stainless will be used for all of the exhaust duct runs and the majority of the inlet duct, although certain higher temperature components may be constructed from Grade 310 stainless or refractory lined pipe.
The supports for the ducting inside containment will be pipe stands resting on the operating deck floor wherever possible. A beam or structural platform across the reactor cavity will be used if required to support the last length of the horizontal duct run. Support of the vertical ducts from the heat exchanger will be provided by the RVTC and lift rig. Support of the ducting outside of containment may
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potentially require structural ties into the spent fuel pool building, in addition to floor supports. Thermal expansion of the duct will be accounted by appropriately designed pipe hangers, saddles, movable supports and strategically placed bends in the duct routing.
In the absence of a directly applicable construction code, the design and fabrication of the ducting will generally be in accordance with ASME 831.3. A complete thermal and structural analysis will be performed to ensure the integrity of the ducting. This analysis will account for the operating temperatures of the duct, and all existing loads on the duct, including dead weight of the duct and insulation, flow induced vibration loads, reactions due to duct supports, and thermal expansion loads.
The ducting will be brought into containment in sections. Joints between sections will be made with mechanical couplings, bolted flanges, or they may be entirely welded. A combination of coupling types may be employed since the design and assembly requirements differ depending on where the joints are being made. For example, ALARA concerns may result in the use of mechanical couplings at locations in the vicinity of the cavity, while bolted flanges may be desirable overall. The type of connections used will be analyzed and justified as part of the design analysis report generated for the overall piping system.
To minimize thermal losses within the ductwork, and to provide protection of personnel and adjacent equipment, the ductwork will be well insulated. Depending on the peak operating temperature of the particular section of duct, and on the proximity to adjacent equipment or personnel routes, additional protective measures will be used. To prevent adverse effects on adjacent structures, spacing (standoffs) and additional insulation between the ductwork and adjacent structures will be used if required. To provide additional personnel protection, temporary barriers may be erected to prevent personnel contact with the ductwork insulation. Ductwork insulation will be mechanically protected from abrasion as required. This may be accomplished by encasing the insulation with high temperature cloth or metallic foil, or by jacketing the insulation with sheet metal. Consideration shall be given to maximize decontamination ability of the ductwork insulation. Refer to Section 1.9.D, for discussion of waste and decontamination issues .
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From the burner platform, the inlet ducts will enter the east side of the spent fuel building. The lines will enter the building, run west to the hatch, then roughly s·outhward into the containment equipment hatch. After entering the hatch, the lines will run across the operating deck, then downward toward the reactor vessel. The exhaust ducting will be routed back up to the operating deck, then out through the equipment hatch. From there the lines will turn upward and westward, and will exit through the west wall of the spent fuel building, preferably below .the bridge crane rail.
1.6.A.3 Controls and Instrumentation Including Redundancy
Section 1.5.B contains a description of the method of instrumentation used during the annealing, including that which is directly applicable to control of the annealing process, and describes the redundancy of measurements. Section 1.5.C describes the I instrumentation accuracy and reliability. Section 2.1 describes the overall monitoring 1 I contingency approach. Section 1.6.A.1.3 describes the gas control trains themselves. · 1
Finally, Section 1.6.A.4 describes the temperature data acquisition equipment which will be used during the annealing. Refer to those sections for a description of the control and instrumentation hardware.
Section 1.5.D.1 includes a description of the overall control methodology that will be used during the annealing process. Refer to .that section for a description of how the control and instrumentation hardware is used during the anneal.
1.6.A.4 Equipment for Measuring and Recording Temperatures and Temperature Profiles
To meet the data display and recording needs for the annealing, a computerized data acquisition system will be employed. The system will provide real time monitoring of all measured temperatures and all calculated gradients, ramp rates, etc. as required. Data will be obtaineo at a rate appropriate to fulfill the objectives described in Sections 1.5 and 2.1. In addition to data storage and logging, extensive display and operator interface features exist, such as alarming and reporting functions, which aid the operator during the annealing. To the extent that the frequency of measurement required for analysis and record keeping differs from the frequency provided for operator feedback and process control, the display of the data may be updated at an interval different than that at which it is stored.
As with all of the annealing· equipment, adequate spares will be supplied such that a · failure of a particular data acquisition system component can be quickly rectified. For protection during loss of power, the data acquisition system will be connected to an uninterruptable power supply, so that continued operation can be maintained while changeover to backup power takes place. The data acquisition system will be calibrated in accordance with Quality Assurance requirements .
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legs are connected to the polar crane main hook using a main link block and a saddle pin assembly. The lift rig will be removed from the RVTC after heat exchanger assembly installation.
The lift rig and support frame components are designed for the heat exchanger equipment and integral shielding lift load. The RVTC, lift rig and/or components will be load tested in accordance with the requirements of the ANSl/ASME NQA-2b-1987, Part 2.15 to provide for safe handling of the annealing heat exchanger assembly. The total lift height of the heat exchanger including the lift rig and RVTC is approximately 43 Feet 9 inches which is within the 46 feet 9 inch Palisades polar crane lift height available.
1.6.A.5.3 Reactor Vessel Drainage/Purge System
An auxiliary RV drainage system will be used to drain the reactor vessel from below the RV nozzles. This system will route any water in the RV bottom to the plant drain system in the refueling cavity fuel transfer pit. The RV drainage system consists of a submersible pump within a pipe integrated into the heat exchanger assembly. After the reactor vessel is drained, the piping used will be reconfigured to allow connectionito a filtered ventilation system. The reactor vessel will then be heated to approximately 250°F and held to dry. Any steam from residual moisture in the reactor vessel will be removed. The reactor vessel may be vented in this manner throughout the annealing process. •
1.6.A.5.4 Nozzle Thermal Barriers
The nozzle thermal barriers are designed to restrict the amount of heat transfer from the .. heat exchanger inside the reactor vessel through the PCS piping during the annealing. The nozzle thermal barriers will be remotely installed 'in and removed from the inlet and outlet nozzles of the Palisades reactor vessel. Without the thermal
· barriers in the nozzles, the six RV nozzle openings provide a significant escape path for the generated heat, the mechanism for escape being the free convective heat transfer down the pipes. Since free convection depends on mass transfer, the nozzle thermal barriers will significantly reduce the amount of air flow from the reactor vessel into the PCS piping. An additional consideration is to reduce the heat transfer via radiation from the heat exchanger and conduction through the nozzle thermal barriers which in turn could set up convection in the air on the backside of the barrier. Because the nozzle thermal barriers will be installed while the reactor vessel is flooded, the
.. _design will_p~.rmit dr~in_age of water thr()ugh the barrier when the reactor vessel is drained and reflooded. ·· · · · ·
Prior to Annealing Properties of Palisades RV Beltline Material (Best-Estimate) ......................... . Prior to Annealing Properties of Palisades RV Beltline Material (Assumed RV Properties for Analyses) ......... . Projected RT Nor Annealing Response of Palisades RV Ma.terials ................................... . Projected CvUSE Annealing Response of Palisades RV Materials ................................... . Operating History of the Palisades RV Cycles 1 Through 11 Chemical Composition of the Original Surveillance Materials for Palisades Reactor Vessel ...................... . Transition Temperature and USE Results for Palisades RV Surveillance Program Base Metal (LT) Plate No. 03803-1,
PAGE
1.1-11
1.1-12
1.1-13
1.1-14 1.1-15
1.1-16
Heat No. C-1279-3 from CVGRAPH . . . . . . . . . . . . . . . . . 1.1-17 Transition Temperature and USE Results for Palisades RV Surveillance Program Base Metal (TL) Plate No. 03803-1, Heat No. C-1279-3fromCVGRAPH ................. 1.1-18 Transition Temperature and USE Results for Palisades RV Surveillance Program Weld Metal Heat 3277 f.rom CVGRAPH 1.1-19 Transition Temperature and USE Results for Palisades RV Surveillance Program HAZ Material from CVGRAPH . . . . . . . 1 . 1-20 Transition Temperature _and USE Results for Palisades RV Surveillance Program SRM Material from CVGRAPH . . . . . . . 1 . 1-21
Palisades Reactor Vessel Beltline Materials ... ,\. . . . . . . . . 1.2-18 Comparison of Best Estir!Jate Reactor Vessel Weld Chemistry and Surveillance Weld Chemistry for Palisades ......... . Palisades Reactor Vessel Recordable Radiographic Indications Palisades Reactor Vessel Recorded Weld Repairs ........ . Fast Neutron Fluence (E > 1.0 MeV) at the Reactor Vessel Clad/BaseMetal Interface . _ .•. _ ...... -...... ,. ....... . Angle Beam Indication Results for Palisades Reactor Vessel
1.3.C.3-1 Maximum Neutron and Gamma Ray Dose Rates During Power Operation Incident on the Palisades Primary Biological Shield ............................. · ..
1.3.C.3-2 Maximum Radiation Exposure as a Function of Radius into the Palisades Primary Biological Shield .................. .
1.3.C.3-3 Relative Axial Distribution of Neutron and Gamma Ray Dose Within the Palisades Biological Shield ................ .
1.3.C.9-1 Results of Reactor Cavity Model Using Corrected Insulation and Liner Temperatures from lnsula~ion Characterization Study (Nominal Air Flow of 5000 CFM) .............. .
1.3.C.9-2 Results of Reactor Cavity Model Using Corrected Insulation and Reflected Liner Temperatures from the Reactor Vessel Characterization Study (Nominal Air Flow of 5000 CFM) ~ ...
1.3.D-1 Piping Design Temperatures and Pressures for Palisades PCS 1.3.D-2 Piping Nozzles in Palisades PCS .......... ; ........ . 1.3.D.9.1-1 As-Left Radiographic Indications on Palisades PCS Piping .. . 1.3.D.9.1-2 Recorded Weld Repairs for Palisades PCS Piping ........ . 1.3.D.9.2-1 Weld lnservice Examination Results on Palisades PCS Piping,
Line PCS-42-RCL-1 H ........................... . 1.3.D.9.2-2 Weld lnservice Examination Results for Palisades PCS Piping, ·
PAGE
1.3-40
1.3-42
1.3-44
1.3-45
1.3-46 1.3-47 1.3-48 1.3-51 1.3-53
1.3-54 .
Line PCS-42.:RCL-2H ........................ ~ . . . 1.3-55 1.3.D.9.2-3 Weld. lnservice Examination Results for Palisades PCS Piping,
Description of Postulated Annealing Conditions Evaluated with 2-D Model for the Palisades Reactor Vessel Annealing 1. 7-24. Description of Postulated Annealing Conditions Evaluated with 2-D Model for the Palisades Reactor Vessel Annealing 1.7-29 RV Temperature Response in the Annealing Zone Used for All 2-D and 3-D Analyses . . . . . . . . . . . . . . . . . . . . . . . . . 1. 7 .35 Maximum Temperatures at Key Node Locations/Regions Established Using the 2-D Model for the. Palisades Reactor Vessel Annealing . . . . . . . . . . . . . . . . . . . . . . . . . 1. 7-36 Peak Stress Intensities, Peak Strains, and Peak Residual Deformations Established Using the 2-D Model for the Palisades Reactor Vessel Annealing . . . . . . . . . . . . . . . . . . 1. 7-37
xv
TABLE
1.7.D-3
1.7.D-4
1.7.D-5
1.7.D-6
1.7.D-7
1.7.D-8 1.7.D-9
1.7.D-10
1.8.A-1-
2.1.A-1
2.1.A-2
2.1.A-3
2.1.A-4
THERMAL ANNEALING REPORT
LIST OF TABLES
PAGE
Inelastic Analysis Strains and Allowables Determined Using the 2-D Model for the Reactor Vessel Annealing . . . . . . . . . 1. 7-38 Maximum Temperatures at Key Locations Defined by the 3-D Model and Alternate Calculations for the Palisades Reactor Vessel Annealing . . . . . . . . . . . . . . . . . . . . . . . . . 1. 7-39 Maximum Stress Intensities Defined Using 3-D Model and Alternate Calculations for the Palisades Reactor Vessel Annealing .. · .................... ·. . . . . . . . . . . . . 1. 7-40 Thermal Gradients Defined Using 2-D and 3-D Models forthe Palisades Reactor Vessel Annealing .......... ~ . . . . . . . 1. 7-41 Maximum Displacements at Key Locations Defined Using the 3-D Model ....... ·. . . . . . . . . . . . . . . . . . . . . . . . . . . . 1. 7-43 Administrative Limits Based on Analysis Results . . . . . . . . . 1. 7-47 Maximum Displacements at Key Locations Defined Using. the 3-D Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1. 7-48 Administrative Limits Based on Analysis Results ......... · 1. 7-50
. .
Limiting Parameters for the Anneal Without Need for Further Analysis and Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.8-4
