ML18065A395

Forwards Preliminary Thermal Annealing Rept,Thermal Annealing Operating Plan,Section 1.3, Equipment,Components & Structures Affected by Thermal Annealing.

ML18065A395
Person / Time
Site:	Palisades
Issue date:	01/12/1996
From:	Smedley R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9601220078
Download: ML18065A395 (83)

Text

.....I consumers Power*

POWERiNii MICHlliAN"S PROliREll Palisades Nuclear Plant: 27780 Blue Star Memorial Highway, Covert, Ml 49043 January 1 2, 1 996 U S Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

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DOCKET 50-255 - LICENSE DPR PALISADES PLANT PRELIMINARY THERMAL ANNEALING REPORT, THERMAL ANNEALING OPERATING PLAN, SECTION 1.3, EQUIPMENT, COMPONENTS AND STRUCTURES AFFECTED BY THERMAL ANNEALING . . .

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 plan to submit the final Thermal Annealing Report (TAR) in the third quarter of 1 996 after the results of the Marble Hill reactor vessel annealing demonstration have been evaluated. The TAR will include the information recommended in Draft Regulatory Guide DG-1027, Format and Content of Application For Approvaf For Thermal Annealing -of Reactor Pressure Vessels. To permit NRC review of the TAR to begin before the Marble Hill results are known, we will make a series of submittals of preliminary TAR sections as they are developed. This letter provides the fourth of those submittals.

The attachment to this letter contains the Thermal Annealing Operating Plan Section 1.3, Equipment, Components and Structures affected by Thermal Annealing. This section is presented in the format recommended by Section C.1 of DG-1027.

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( 9601220078 960112 PDR ADOCK 05000255 P PDR A CMS' ENER6YCOMPANY

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• 2

SUMMARY

OF COMMITMENTS This letter contains no new commitments and no revisions to existing commitments.

~?rd~S~~?j Manager, Licensing CC Administrator, Region Ill, USNRC Project Manager, NRR, USNRC NRC Resident Inspector - Palisades Attachment

ATTACHMENT CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255

• THERMAL ANNEALING REPORT SECTION 1 THERMAL ANNEALING OPERATING PLAN SECTION 1.3 EQUIPMENT, COMPONENTS AND STRUCTURES AFFECTED BY THERMAL ANNEALING

• 80 Pages

• 1.3 1.3.A EQUIPMENT, COMPONENTS AND STRUCTURES AFFECTED BY THERMAL ANNEALING Description of Equipment, Components and Structures The equipment, structures and components that will be affected, either thermally or mechanically, during the annealing operation are described in the following sections. Note that the effects on the reactor vessel and reactor vessel nozzles as well as internal attachments such as the flow skirt, capsule holders, core support lugs, core stabilizing lugs and cladding are addressed in Sections 1 . 2 and 1 . 7.

1.3.A.1 Primary Coolant System Loop Piping The clad carbon steel primary coolant system (PCS) loop piping is welded to the reactor vessel (RV) at the RV primary nozzle extensions. The cold leg piping runs between the four RV inlet nozzles and the primary coolant pump (PCP) discharges and between the PCP suctions and the steam generator outlet nozzles. The cold leg piping between the RV inlet nozzles and the PCP's consists of an elbow, which is connected to the RV inlet nozzle extension, and a straight run, which is connected to the PCP discharge nozzle. The cold leg piping between the PCP suctions and the steam generator outlet nozzles consists of an elbow connected to the steam generator nozzle transition piece, a straight run, another elbow, and another straight run which is connected to the PCP suction elbow transition piece.

The cold leg pipe sections have a 30-inch nominal l.D. and a 2 3/4-inch nominal base metal wall thickness. The elbows have a 30-inch nominal l.D. and a 3-inch

• nominal wall base metal thickness.

The hot leg piping runs from the two RV outlet nozzles to the steam generators.

Each hot leg piping run consists of a straight section of pipe, which is connected to the RV outlet nozzle extension, and an elbow, which connects to the steam generator inlet nozzle. The hot leg pipe sections have a 42-inch nominal l.D. and a 3-3/4-inch nominal base metal wall thickness. The elbows have a 42-inch nominal I. D. and a 4-1 /16-inch wall thickness.

The piping material is SA 516 Grade 70 carbon steel, mill clad on the l.D. per SA 264 (roll bonded) with 1 /4-inch minimum thickness austenitic (304L) stainless steel. The piping is welded to the A 508-64, Class 1 carbon steel nozzle extensions on each of the six RV primary nozzles. The joints in the PCS piping and at the RV primary nozzles were clad internally with two passes of weld deposited austenitic (309/308) stainless steel. A more detailed description and discussion of the piping is included in Section 1.3.D.

A plan view and elevation view of the PCS, with piping component numbers indicated, is presented in Figures 1.3.A.1-1 and 1.3.A.1-2, respectively. The run lengths of the sections of hot and cold leg piping are presented below:

• TAR 1/12/96 1.3-1

• Leg Hot Cold Section RV to SG 2/1 Pump 1 8/28 to SG 1/2 Section Number 673-01 &

673-04 673-05+

Run Length (Pipe + Elbow = Total) 136 1/8 in + 381/2in =

14 ft 7 in 57 9/16 in + 67 1/4 in +

674-07 & 47 1/8 in =

Pump 1A/2A to SG 1/2 674-05 + 14 ft 4 in 674-07 Cold Pump 2A/1 A to RV 673-08 & 165 9/16 in + 25 3/4 in =

673-11 15ft11 in Cold Pump 28/1 8 to RV ., 674-01 & 143 3/4 in + 47 1/8 in =

674-04 15ft11 in The two Palisades steam generators are joined to the reactor vessel by the hot leg piping in the PCS loop arrangement. The steam generators are separated from the reactor vessel by approximately 14 feet of piping and the biological shield.

Therefore, the steam generators will not be subjected to the elevated annealing temperatures. However, the steam generator inlet nozzles will be loaded as a result of the thermal displacements of the reactor vessel and piping during the annealing operations .

1.3.A.3 Primary Coolant Pumps The four primary coolant pumps are separated from the reactor vessel by approximately 16 feet of piping and the biological shield. Like the steam generators, the primary coolant pumps are not subjected to the elevated annealing temperatures. However, the pump discharge nozzles will be loaded by the thermal displacements of the reactor vessel and piping.

1.3.A.4 Reactor Vessel Supports The RV supports provide the load path from the reactor vessel to the structural concrete. The Palisades reactor vessel has three support assemblies under supports pads on RV nozzles at 120° intervals around the reactor vessel. The supports and associated structures are located within the reactor cavity and are uninsulated. The main load carrying members of the supports are the structural steel beams and columns that are built into the structural concrete of the biological shield. The nozzle support pads are bolted into the ASTM A536-65T, Grade 65-45-12, ductile cast iron sole plates in the expansion plate assemblies. The bottom components of the expansion plate assemblies are the expansion plates themselves, which are pads of Merriman Alloy 424 (ASTM 822 Alloy E) bronze with Type II lubricant or equivalent. These ride on carbon steel shim plates which are mounted on support beams. The expansion plate assemblies allow radial translation and rotation at the RV nozzle support pads to accommodate thermal expansion and contraction of the reactor vessel during normal operation as well as annealing.

TAR1/11/96 1.3-2

1.3.A.5 Reactor Vessel Insulation The RV thermal insulation is a reflective insulation system consisting of a network of panels with 304 stainless steel sheet outer skins enclosing layers of aluminum foil and a 304 stainless steel sheet metal divider. The RV insulation panels are typically four inches thick and are held together by truss head bolts in tube spokes through the thickness of the panels. The bolts are unthreaded, but are held in place by welded washers.

The insulation is mounted on the reactor vessel from a steel channel support ring supported by stanchions which rest on top of the RV nozzles. The RV insulation is assembled in a cylindrical shell arrangement extending above the support ring to the cavity seal ledge and below the support ring to the lower shell. The RV shell insulation terminates at the lower shell, but is joined there by the horizontal run of the biological shield insulation which extends to the reactor cavity wall. The RV shell insulation panels are attached to the support ring and to each other by sheet metal screws. The insulation is installed to leave a 3/4-inch nominal gap around the reactor vessel O.D. to allow for radial thermal growth of the reactor vessel during heat up. The RV nozzles are also covered with 4-inch thick sections of reflective insulation enclosing the nozzles except for the support pads, which are uninsulated. The reflective insulation acts as a combination of dead air space, radiation reflectors and convection barriers to minimize the heat loss from the reactor vessel.

1.3.A.6 Primary Coolant System Loop Insulation The primary coolant piping insulation within the biological shield is also reflective insulation with aluminum foil layers enclosed in panels of 304 stainless steel sheet metal. The reflective insulation on the piping is typically three inches thick and forms a cylindrical shell around each of the PCS piping loop runs, extending from the RV nozzles through the biological shield penetrations.

Much of the PCS piping insulation outside of the biological shield has been replaced by Transco Thermal-Wrap fiberglass insulation. This fiberglass insulation is in blankets of fiberglass cloth which wrap around the piping and are attached with velcro strips. The piping with Thermal-Wrap insulation presently includes sections of the hot legs between the biological shield and the steam generators and the crossover legs between the steam generators and the primary coolant pumps.

1.3.A.7 Biological Shield The biological shield is the reinforced concrete and steel liner surrounding the

, reactor cavity. A detailed description and discussion of the biological shield is included in Section 1.3.C.

1.3.A.8 Biological Shield Insulation The biological shield insulation system, illustrated on Figure 1.3.C-1 as Region 1, utilizes two types of insulating material covered with austenitic stainless steel sheet metal and/or aluminum foil. The hinged panels extending horizontally from the RV lower shell insulation and the panels covering the lower part of the biological shield wall are filled with mineral wool insulation. The bottom covering of the hinged TAR 1/11 /96 1.3-3

panels is stainless steel sheet metal, and the top covering consists of wire mesh and aluminum foil. The insulation on the lower reactor cavity wall from the floor up 10 feet to the 601 foot elevation (where the hinged horizontal panels are mounted) is 4 inch thick mineral wool insulation fixed against the cavity liner. The wall insulation surface toward the interior of the reactor cavity is covered with stainless steel sheet metal.

The floor under the reactor vessel is insulated for the most part by four 2-inch thick layers (total thickness 8 inches) of Unibestos insulation. There are, however, three recessed areas in the floor with reduced thicknesses of Unibestos block insulation.

Since the reactor vessel extends down below the floor level (Elevation 591 feet 3 inches), there is a 4 inch deep recess approximately 4 feet square around the apex of the bottom head. The floor insulation thickness beneath this recess is 4 inches.

There is a second recess in the floor insulation at the opening of the access tube inside the reactor cavity. This recess is approximately 30 inches square and is 4 inches deep to permit an insulated cover to be mounted over the access tube's interior opening. The floor insulation thickness beneath this recess is also 4 inches.

The third recess is a cutout in the insulation around the floor drain. The cutout is 18 inches square with a depth of 4 inches. The insulation thickness at the floor drain is 4 inches. Ail of the floor insulation is covered by austenitic stainless steel floor plate.

The removable insulation cover which is placed over the access tube opening inside the reactor cavity is insulated with a 10-inch thickness of Nukon fiberglass insulation .

• 1.3.A.9 0-Ring Leakage Monitor Tube Leakoff Piping The 0-ring leakage monitor tubes are 3/4-inch schedule 80 piping which extend out of the RV flange penetrations. They are connected to a leakoff piping system through which leakage past the 0-ring gaskets during reactor operation can flow to the drain tank. The system uses a pressure switch to detect inner seal leakage and a level switch to detect outer seal leakage, with valves to isolate sections of the system. The valves and the instrumentation are located outside the biological shield.

1.3.A.10 Biological Shield Liner The biological shield liner forms a metallic barrier on the wall of the biological shield in the reactor cavity adjacent to the reactor vessel. The biological shield liner is weld constructed from carbon steel plate material. The plate thickness is 5/1 6 inch from the 599-1 /2 foot elevation up to 624 foot 2 inch level opposite the RV flange.

The plate thickness is 1 /2 inch from the 599-1 /2 foot level down to the reactor cavity floor. The liner has an 11 foot inner radius. Therefore, the inner surface of the liner is about 32 inches from the outer surface of the RV shell insulation. There are openings in the biological shield liner for the neutron detector wells, RV support structures and primary coolant piping penetrations.

1.3.A.11 Cavity Seal Drip Pan and Drain Lines

• A stainless steel sheet metal drip pan is attached to the refueling cavity liner extension plate material in order to collect minor leakage past the reactor cavity TAR1/11/96 1.3-4

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 reac~or vessel from the floor drain makes a 90° turn 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 1 5 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.B ________ Ef.f.ects__QO E_guipm~_r1t, C_Q_ITIP9r:l~_nt§ ~11cJ S_tr!J!?ture§__

The effects on the above discussed equipment, components and structures as a result of annealing are presented in the following sections.

1.3.B.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 and 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 Section Ill of the ASME Boiler and Pressure Vessel Code.

1.3.B.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 TAR 1/11/96 1.3-5

mechanically loaded as a result of the thermal expansion of the reactor vessel and piping. The thermal expansion loading from the PCS piping will cause the steam generators to translate on the steam generator sliding supports as during normal operation. However, the thermal pipe reactions (including bending) due to the local heating of the reactor vessel are less than the design thermal piping reactions. The analytical results in Section 1 . 7 conclude that the horizontal movement at the steam generators during annealing is less than the horizontal movement during normal operation. The stress at the steam generator end during annealing was determined to be minimal as identified in Section 1 . 7. The loadings on the steam generator nozzles are developed from the analytical results and compared to the design loads due to thermal expansion in Section 1 . 7.

1.3.B.3 Primary Coolant Pumps The primary coolant pumps should not be heated up during the annealing operation because of their distance from the reactor vessel. The pump temperatures will remain well below normal operating temperatures and there will be no significant thermal gradients. The primary coolant pump disc-harge nozzles will also be loaded due to the restrained thermal displacement of the reactor vessel and piping. The pump support is designed to accomodate the primary coolant pump displacement.

The loading of the discharge nozzles during annealing (including bending) is less than the design loads. The resulting horizontal deflection and maximum stress at the primary coolant pump discharge nozzle is similar to that cited for the steam generator nozzle in Section 1.3.8.2.

1.3.B.4 Reactor Vessel Supports The RV supports are in intimate contact with the RV nozzle support pads. They are uninsulated and exposed to the reactor cavity. Therefore, they will be subjected to temperatures during annealing that are higher than their normal.

operating temperature as the result of conduction heat transfer through the RV nozzles plus radiation and convection within the cavity. The temperatures are quantified by the thermal analysis in Section 1. 7.

The mechanical loading on the RV supports during the annealing heatup and cooldown will not be significantly different from the loading during normal heatup and cooldown since the nozzle support pads are free to slide as the reactor vessel heats up and cools down.

The effect of the elevated annealing temperatures on the lubricant is a specific concern for later operation of the reactor vessel. The RV support expansion plates are bearing plates of Merriman Alloy 424 (ASTM 8-22 Alloy E) bronze with Type II dry lubricant in trepanned holes to provide solid film lubrication between the expansion plates and the mating sliding support sole plates. This bearing plate system is among a class of bearing systems commonly known by the Merriman trade name Lubrite. The expansion plates have a design temperature of 650°F like the reactor vessel. In the case of the RV sliding supports, this is the maximum temperature for the dry lubricant. At temperatures above 650°F, the lubricant binders can degrade and cause the lubricant in the trepanned holes to crumble into a fine powder. The powdered lubricant could then be postulated to migrate out of the bearing, thus increasing the coefficient of friction between the expansion plates and the sole plates. However, the temperature at the bottom of the support pads TAR 1/11/96 1.3-6

will be less than 650°F during the annealing process as indicated in Section 1. 7.

Therefore, degradation of the solid lubricant at temperatures above 650°F is not an issue.

In addition to the lubricant concern, the bronze expansion plates were specifically sized to *carry the bearing load resulting from the weight of the reactor vessel, reactor internals and fuel, fully assembled RV head, water volume, etc. at 650°F.

However, the thermal annealing will be performed without the weight of the RV head, reactor internals, fuel and water volume. The heat exchanger and RV top cover which add load to the RV supports during thermal annealing are lightweight by comparison. Therefore, the bearing load of the bronze expansion plates is not an issue.

1.3.B.5 Reactor Vessel Insulation The reflective insulation around the RV annealing zone will be subjected to the

.. rn~~ir_:nu_m_anD~_C!Jing temperature profile at the inside surface of the panels. The temperature will decr-ease-fhrougn tlle4::.incl1-thicl<n-ess of theq:Janels~- Tlie-* .... ---

insulation panels will flex to accommodate the thermal gradients with no degradation of the insulating properties. Due to its forgiving construction, the RV insulation will maintain its shape and thickness during the thermal annealing operations. The construction of the RV insulation panels with the layers of aluminum foil and stainless steel sheet metal pinned in between the inner and outer stainless steel sheet metal skins will accommodate the large thermal gradients through the 4-inch thickness. Each layer of material within the panels is free to

• • grow and contract radially as its temperature changes. The construction also permits the layers to slide over one another to accommodate changes in curvature.

Therefore, no destructive forces due to the large through-thickness temperature gradients will develop within the panels during the annealing heatup and cooldown.

The smaller and more gradual axial and circumferential thermal gradients in the insulation will be accommodated by the local flexibility at the joints between the panels within the insulation network. Insulation temperatures at selected locations are quantified in Section 1.3.C.

The effect of the annealing on the effectiveness of the thermal insulation will be deterioration of the insulating properties of the reflective insulation within the reactor cavity. Additional oxidation of the stainless steel and aluminum surfaces may occur at the maximum annealing temperature, thereby further degrading the reflective properties and increasing the radiation heat loss. ASTM Standard C 667 for reflective insulation systems operating at temperatures above ambient air states that the hot surface temperature for aluminum should be limited to 750°F due to increased emittance resulting from surface oxidation. The limit for 300 series stainless steel is 1200°F. With the 900°F maximum annealing temperature on the surface of the RV shell, the temperature on the inside surface of the RV insulation panels will be approximately 895°F. Assuming a linear temperature decrease through the thickness of the panels to approximately 365°F at the O.D., the temperature of only three of the nine layers of aluminum foil will exceed 750°F.

Since the efficiency of the radiant heat transfer across the surfaces of these three layers of aluminum foil is only a very small contributor to the overall effective conductivity of the reflective insulation, the effect of increased surface oxidation is considered negligible for continuing operation of the plant after annealing. The melting point of aluminum is 1220°F which is well above the annealing TAR 1 /11 /96 1.3-7

temperatures. Therefore, no degradation of the aluminum foil that would increase the convective heat transfer through the in*sulation is expected .

At the annealing temperature, the reactor vessel will grow radially approximately 0.71 inch. The effect of the thermal growth on the 3/4-inch nominal cold gap between the reactor vessel and the insulation will be to reduce the gap during the annealing heatup. However, any interference between the reactor vessel and the insulation will be minimal so that the sheet metal insulation panels may deflect or slightly deform (less than 1 /4-inch deflection) to accommodate the strain. The temporary temperature sensors which will be mounted on the O.D. of the reactor vessel during the annealing process will be small relative to the cold gap width in order to minimize the interference due to wedging between the reactor vessel and the insulation. In addition, the temperature sensors will be sheathed in stainless steel so that the sensors will not be crushed, thereby changing their characteristics should interference develop. The interference will be relieved as the thermal equilibrium is obtained during the annealing operations. Although the insulation temperature will initially lag the temperature of the RV shell during the annealing heatup, the insulation will eventually heat up and grow thermally away from the vessel O.D. to reopen the gap. There will be no permanent degradation of the insulation properties as a result of the interference during the annealing heatup.

1.3.B.6 Primary Coolant System Piping Insulation The maximum temperatures of the PCS piping reflective insulation within the biological shield is within the 650°F design value for the piping. The fiberglass insulation on the PCS piping is all outside the biological shield and will be subjected to even lower temperatures. Therefore, there will be no deterioration of the piping insulation properties as a result of thermal annealing operations. The temperatures of the PCS piping are quantified in Section 1 . 7.

1.3.B.7 Biological Shield The biological shield heat load during annealing will be higher than during normal operation. Therefore, acceptable temperature limits for the structural concrete of the biological shield are established for the annealing. The temperature limits will be maintained during the annealing operations by means of supplemental cooling and bottom head insulation. The details of the biological shield considerations are provided in Section 1.3.C, Description of Biological Shield.

1.3.B.8 Biological Shield Insulation The biological shield insulation temperature could be higher than normal operating temperature depending upon the location in the reactor cavity. Two locations will see the greatest mechanical effect of the annealing temperatures; the hinged horizontal panels of the biological shield insulation adjacent to the RV shell, and the Unibestos cavity floor insulation in the recess around the RV bottom head.

The hinged horizontal panels are mounted between the RV shell insulation and the cavity wall insulation. The insulation structure with the attachment to the outer skin of the RV reflective insulation is very flexible so that any mechanical effect is regarded as not detrimental.

TAR 1/11/96 1.3-8

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 0.5 inch to 0. 7 inch 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 1 600 lb/ft 2 (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 0. 7 inch. 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 local 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 wali 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 1 000°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.

1.3.B.9 0-Ring Leakage Monitor Tube Leakoff Piping 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 less than the design radial and vertical thermal growth of the RV flange.

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 TAR 1/11/96 1.3-9

liner plates will heat up and cool down with the reactor vessel during the annealing operations. Supplemental cooling in the reactor cavity will maintain the cavity air temperature at 170°F, which is below normal operation limits. This will control surface temperatures at the cavity liner so that the thermal flexing during thermal annealing operations will not be an issue.

1.3.B.11 Cavity Seal Drip Pan and Drain Lines The drip pan loading during the annealing should be comparable to normal operation. Even with higher temperatures, the drip pan is free to move radially since it is supported only by the refueling cavity liner extension. Therefore, no degradation will result from thermal annealing.

With supplementary air cooling the thermal growth of the drain lines will remain within the bounds of normal operation. Therefore, the thermal loads on the piping and the drip pan due to thermal displacements will not be detrimental to either.

1.3.B.12 Reactor Cavity Floor Cooling Coils- -

The cooling coils under the reactor cavity floor will be subjected to a higher thermal load during thermal annealing than during normal reactor operation in the local area of the 4 ft. x 4 ft. recess of the floor insulation under the RV bottom head. The RV bottom ,head will reach a higher maximum temperature during the thermal annealing than during normal operations. In addition, degraded insulation due to increased thermal expansion of the reactor vessel, as discussed in section 1.3.B.8, will

• reduce the local effectiveness of the floor insulation. However, supplementary thermal insulation will be applied to the RV bottom head and to the floor around the 4 inch deep recess in order to restrict the area affected by the higher heat load.

The 170°F cavity air outside the supplementary insulation will maintain the balance of the cavity floor at temperatures much less than the 533 ° F air temperature during normal operation. As a result, the overall cooling capacity of the 15 ft. x 15 ft. bank of cooling coils is sufficient to cool the concrete of the reactor cavity floor so that the 250 ° F upper bound temperature limit for the concrete will not be exceeded during the thermal annealing process.

1.3.C Description of Biological Shield 1.3.C.1 Dimensions The reactor cavity arrangement for Palisades consists of a carbon steel lined circular cylinder which extends from the base mat at elevation 583' - 6" to the RV flange at elevation 624' - 6". The reactor cavity biological shield wall has an inside diameter of 22 feet and a thickness that varies from 7 feet to 8 feet. At elevation 614' - 2.5" (Hot Legs) and 614' - 8.5" (Cold Legs), the bioshield wall is split by the PCS piping penetrations. The resulting piers extend up to the RV flange. Figure 1.3.C-1 provides a section view of the reactor cavity.

1.3.C.2 Materials r-;; '\ The bioshield wall below the PCS piping penetrations is composed of two distinct radial regions. The first 1O" is referred to. as "sacrificial" concrete. This region contains the embedded Shield Cooling System (SCS) coils which are designed to TAR 1 /11 /96 1.3-10

remove the heat from the reactor vessel. The region between this sacrificial concrete and the outer shield wall is the normally reinforced structural concrete which provides strength for the RV supports and dynamic effects associated with pipe rupture and seismic loads. The bioshield wall above the PCS piping penetrations does not contain embedded shield cooling system coils and has no sacrificial concrete. The concrete was designed to the requirements of ACI 318-63, "Building Code Requirements for Reinforced Concrete".

1.3.C.3 Radiation Exposure of the Biological Shield Over the course of the reactor operating history, the portions of the primary biological shield located adjacent to the reactor vessel experience exposure to both neutron and gamma radiation. During reactor power operation, the shield is exposed to neutron and gamma irradiation resulting from fissioning in the reactor core as well as to secondary gamma irradiation resulting from neutron interactions with structural materials within and external to the reactor core. During shutdown

-i;>e!'.iG>ds.,--the-biological sbield.is_also__exp.ose.d_to additLon.al_ga_111m~ r(lg_i_atiQ.Q re_sutti_ng _

from activation of structural components within the system. These activation-induced gamma ray dose rates during periods of reactor shutdown are much lower (approximately three orders of magnitude) than the levels experienced during power operation and, therefore, can be neglected when computing the total integrated exposure of the primary biological shield.

For periods of power operation, calculations of neutron and gamma ray dose rates incident on the primary biological shield were performed using two-dimensional discrete ordinates techniques in r,8 geometry. All of the transport calculations used the BUGLE-93 cross-section library. The BUGLE-93 library is a 67-group coupled neutron/gamma-ray ENDF/B-VI data set produced specifically for light-water reactor applications. In the Palisades analysis, anisotropic scattering was treated with a P3 expansion of the scattering cross-sections and the angular discretization was modeled with an S16 order of angular quadrature.

The core power distributions utilized in the transport calculations were generated with the normalized nodal cycle energy database. The aforementioned core power distributions were provided in terms of average pin power within the outer assemblies, axial power shapes, and the beginning-of-cycle and end-of-cycle assembly burnups. From the fuel burnup for each assembly, an average fission spectrum for that assembly was calculated using ENDF/B-VI fission spectra for each fissionable isotope. Using the fission spectrum for each assembly, the assembly power generation obtained from the cycle burnup and the power shape within the assembly from relative pin powers, a 67 group neutron/gamma ray fission source was calculated for each mesh point in the transport model of the reactor core. Secondary gamma ray production in structural materials, internal as well as external to the reactor core, were calculated from the neutron flux solution via the coupling included in the cross-section data base.

Results of the radiation exposure evaluations for the Palisades primary biological shield are provided in Tables 1.3.C.3-1 through 1.3.C.3-3. The information provided in Table 1.3.C.3-1 represents the axial maximum neutron and gamma ray

.. \ dose rates at several azimuthal locations along the biological shield inner radius .

The inner radius is the bounding location as evidenced by the data on Table 1.3.C.3-2. Data are provided for each operating cycle to date. Along with the TAR 1/11 /96 1 .3-11

radiation levels, the reactor operating history in terms of effective full power days per fuel cycle are also provided. In Table 1.3.C.3-2, the integrated neutron and gamma ray dose rates through the end of cycle 11 are provided as a function of depth into the concrete biological shield wall. Again, the information in Table 1.3.C.3-2 is applicable to the core midplane elevation and represents the axial maximum in the exposure distribution. In Table 1.3.C.3-3, a relative axial distribution of neutron and gamma ray dose is provided so that the midplane data from Table 1.3.C.3-2 can be adjusted to axial elevations spanning +/- 6.5 feet relative to the midplane of the reactor core. Thus, for a given elevation the integrated dose, D(z), is determined by:

D(z) = D(Midplane) x F(z) where:

D(z) = Dose at axial elevation z D(Midplane) = Dose at the midplane elevation from Table 1.3.C.3-2 F(z) = Relative axial factor from Table 1.3.C.3-3 The relative axial distribution function represents a typical traverse derived from neutron flux measurements obtained in the Palisades reactor cavity during fuel cycles 8 and 9.

• 1.3.C.4 Reactor Vessel and Biological Shield Insulation The RV thermal insulation is a reflective insulation system consisting of a network of panels. A description of the RV insulation is provided in section 1.3.A.5. The biological shield insulation consists of hinged panels that extend horizontally from the lower RV insulation and panels that cover the lower part of the biological shield wall .. A description of the biological shield insulation is provided in section 1.3.A.8.

The horizontal hinged panels prevent hot air from migrating up the annular region between the RV insulation and the biological shield wall during normal operations.

1.3.C.5 Cooling Provisions The shield cooling system (SCS) is a closed-loop system consisting of two full-capacity sets of cooling coils, two full capacity pumps, a heat exchanger, a surge tank, and associated piping and valves. The heat is rejected to the component cooling water system through the SCS heat exchanger. The design parameters for the SCS are:

• Design heat removal capability (per loop): 180,000 Btu/hr

• Coolant pumps: 125 GPM each

• • SCS Heat Exchanger Design Duty: 200,000 Btu/hr The SCS provides cooling coils at the cavity floor below the insulation and the biological shield walls up to elevation 614'. The SCS consists of 3/4" pipes embedded three inches and seven inches into the bioshield wall concrete with a TAR 1/11 /96 1.3-12

circumferential spacing of approximately nine inches. In order to provide cooling for the main RV supports, the cooling coils are spaced at closer intervals to remove the heat that may be conducted down the main RV supports. A bank of cooling coils approximately 15 feet square is located beneath the cavity floor insulation.

The area above elevation 614' to the RV flange is not cooled by the SCS; heat is removed by heat transfer through the piers.

Based on the insulation and configuration of the SCS piping, the reactor cavity can be divided into three separate regions as depicted in Figure 1.3.C-1.

Region 1: Insulation panels attached to the cavity liner and cooling coils embedded three inches and seven inches into the bioshield wall.

Region 2: Insulation panels attached to the reactor vessel and cooling coils embedded three inches and seven inches into the bioshield wall.

Horizontal insulation panels separate the air spaces of Regions 1 and 2.

Region 3: Insulation panels attached to the RV wall. There are no cooling coils in the concrete region above elevation 614' (piimaiy coolant piping penetrations).

1.3.C.6 Concrete Properties The proposed annealing operation will place a significant source of heat in the vicinity of the biological shield wall. The high temperature of the annealing operation and its duration may expose portions of the biological shield wall to higher than normal temperatures.

Different concrete types and mixes respond to high temperatures in different ways.

Accordingly, six significant concrete properties have been considered to evaluate the potential impact of high temperature exposure on the Palisades biological shield wall concrete performance. These are as follows:

Aggregate Type Cement Type Moisture Conditions State of Loading (during high temperature exposure)

Compressive Strength Age of Concrete Of these six concrete properties, the two most important with respect to the concrete's resistance to degradation when exposed to high temperatures are aggregate type and moisture conditions. Each of these properties is discussed briefly with respect to the Palisades design in the paragraphs that follow.

Aggregate Type Industry literature indicates that the loss of compressive strength of concrete upon exposure to high temperatures is considerably lower with aggregates that do not contain silica. Dolomite and limestone aggregates do not contain silica and are therefore considered to have the least effect on concrete strength degradation TAR 1/11/96 1 .3-13

compared to other types of aggregate. This is further verified by Portland Cement Association (PCA) test results where it was concluded that "Specimens made of carbonate aggregate ... retained more than 75% of their original strengths at temperatures up to 1200°F when heated without load and tested hot. The corresponding temperature for the siliceous aggregate was about 800°F."

Petrographic analysis of aggregate used in the Palisades plant concrete indicates the aggregate is predominately dolomite and limestone. Petrographic analysis included the observation that "the well interlocked crystals and lack of intercrystalline pores should result in high resistance to physical deterioration."

Therefore, it is concluded that the biological shield wall concrete is made with one of .the more heat resistant of the commonly-used concrete aggregates and will represent a conservative situation (compared to use of siliceous aggregate types) when considering research data consisting of concrete containing any other or unknown aggregate types.

Cement Type The Palisades biological shield wall cement type is ASTM C-150 Type II. This concrete type is adequate for continued strength gain over time, particularly with

---

kleal-cu.-ing-conditions ..- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - 1 Moisture Conditions The Palisades biological shield wall moisture conditions appear to have been favorable over time to support continued strength gains over the last 27 years.

These conditions can be summarized as follows and point to the ability of the biological shield wall concrete to maintain moisture for continued hydration:

Moist Concrete that is continually in a moist environment is better able to maintain continued hydration of cement to support the long term strength gains that are achievable.

Lined The fact that the biological shield wall is lined on the interior face helps to seal in moisture. This aids in mciintaining proper hydration to support the long term strength gains that are achievable.

It is a well-accepted axiom in the concrete industry that mass concrete never thoroughly dries, that it remains moist or damp except for the first few inches near the surface. Research has indicated and logic dictates that mass concrete that is sealed on one side by a steel liner must be considered to be in the moist condition when considering the effects of high temperature exposure.

The biological shield wall concrete is eight feet thick and sealed on the interior by a steel liner. For the purpose of evaluating the effects of higher temperatures on concrete performance, the biological shield wall concrete is considered to be sealed to moisture in its current state. The liner provides total sealing to moisture loss and, coincidentally, this most critical region near the liner is the same area that is exposed to the elevated temperature being considered. This represents the most conservative situation when considering research data on high temperature TAR 1/11 /96 1.3-14

exposure of concrete samples consisting of unsealed concrete or both sealed and

• unsealed concrete (sealed or unsealed to moisture loss).

State of Loading Various literature exists which considers the effects of high temperature exposure to concrete in different states of loading. A literature search conducted by the PCA's Construction Technology Laboratory (CTL) considered research for concrete heated in both the stressed (loaded) and unstressed (unloaded) conditions. One particular conservatism noted in the CTL report is that the observations contained therein are based on concrete heated in the unloaded condition. This represents the conservative condition since more strength degradation occurs at moderately elevated temperatures for unloaded concrete versus loaded concrete. This is a conservatism for the biological shield wall concrete when considering the CTL observations since it is, at a minimum, loaded with the dead loads.

Compressive Strength The Palisades biological shield wall compressive strength requirement at 28 days has been set at 5000 psi per Paiisades Specification Number 5935-C-30. Actual in-place compressive strength at 28 days is higher than the required 5000 psi as

_____verified_b-.,crepresentafot_e testing_fmm_as-=pJac_ed_c_on_crete__i_QtJ:be biolog~_q_I shield wall as follows:

Compressive Strength 28-Day 90-Day 6366 6579 6226 6720 5625 6579 5873 6720 6084 Unknown 6137 Unknown 6084 Unknown 6066 Unknown 5022 6084 5129 5978 5098 6508 5270 6119 It is a well-established fact that Portland cement concrete will continue to gain strength at a decreasing rate with increasing time as long as favorable humidity conditions exist for continued hydration of cement. As previously mentioned, the biological shield wall concrete is mass concrete which in itself is normally considered to never dry out. Furthermore, concrete exposed to the highest temperatures during the annealing operation will be that nearest the liner. This region represents ideal moisture conditions since it is considered to be totally sealed against moisture loss by the seal liner. For the biological shield wall concrete, therefore, ideal moisture conditions have existed for continued strength gain over the last 27 years .

TAR1/11/96 1 .3-15

Age Of Concrete Research indicates that the longer concrete is cured before exposure to high temperatures, the less significant will be the deterioration of the structural properties of the concrete.

1.3.C.7 Design Temperature Limits A design temperature limit for the biological shield wall is not specifically defined for the Palisades plant. For purposes of discussion clarity, a listing of temperatures and their particular licensing or codes and standards significance to Palisades is provided:

150°F The typical design and normal upper temperature limit for concrete quoted in ACI 349 below which degradation of strength or stiffness is not expected to occur.

165°F The estimated concrete biological shield wall temperature above which the shield cooling system shall be operating per Palisades Technical Specifications.

1_8_0~F_____ the __ tem~raJ_u_r:._e of the bioJQgical shield wall structura_I_ __

concrete that Bechtel Palisades Plant design criteria says shall not be exceeded.

• 200°F The temperature of the biological shield wall non-structural (sacrificial) concrete that Bechtel Palisades Plant design criteria says shall not be exceeded. More recent nuclear power concrete design criteria also define 200°F as a long term operating temperature limit for local, loaded concrete areas such as penetrations and other surfaces.

250°F The current temperature limit for the Palisades biological shield wall concrete inner surface which, if measured by existing instrumentation, is acceptable given that it exists only locally for lightly loaded concrete near the inner surface.

350°F The temperature limit used in newer nuclear power concrete associated with local, short term accident loads such as jet impingement etc.

Concrete temperature is not a design basis value. However, the design basis load allowables must be met at these higher temperatures and resultant increased (decreased) loads and capacities must be reconciled to the design basis, see Section 1.3.C.8.

1.3.C.8 Justification for Exceeding Normal Temperature Limit of 150°F The proposed annealing operation will raise portions of the biological shield wall above the normal temperature limit of 150°F. This requires an evaluation of the effects of the annealing temperature transient on the integrity/strength of the

'biological shield wall concrete and its ability to continue to perform its design TAR 1 /11 /96 1 .3-16

functions. The evaluation is presented in two parts considering the two aging effects exhibited in concrete strength variation over time and temperature:

• Concrete strength improves with age (curing) under the right environmental conditions.

• Concrete strength degrades when exposed to high temperatures for significant periods of time.

The key to the evaluation of the biological shield wall concrete is that both of these aging effects can be quantified based on knowledge of the actual biological shield wall concrete properties from Section 1.3.C.6 and the information derived from industry studies of concrete behavior under conditions similar to the proposed annealing operation listed in Section 1.3.G. A conservative estimate of the net effect of the annealing operation is derived by taking a conservatively low expected increase in the original 28-day concrete strength (due to curing since original pour) and comparing this to a conservatively high decrease in concrete strength due to exposure to high temperatures.

Concrete Compressive Strength Improvement

---

A-conser-vative-quantification--oLtbeJong_term_increase_in_c_o_ncrn_te_c_omp_rn_s_s_i~e strength due to curing is presented as follows:

1. Evaluate as-placed concrete compressive strength results.

• An evaluation of original 28-day concrete compressive strength results yields an acceptable level of confidence for a resulting concrete compressive strength when the biological shield wall concrete was originally placed.

Representative as-placed results indicate appreciably higher original concrete compressive strengths than required by design specifications.

2. Consider long-term concrete compressive strength gain from as-placed time (28 days) to the present.

Consideration is given to the ideal long-term curing conditions provided by the moist, sealed concrete. The Portland Cement Association (PCA) and U.S.

Bureau of Reclamation literature form the basis for the compressive strength gain analysis.

These sources contain specific information related to continued long-term strength gain of concrete. PCA information indicates that concrete moist-cured for 180 days will develop approximately 129% of the 28-day moist-cured strength. This same PCA source also indicates that continuously moist-cured concrete with Type II Portland cement (biological shield wall concrete is specified to have Type II Portland cement in accordance with Specification Number 5935-C-30) that achieves approximately 4200 psi at 28 days will achieve approximately 6400 psi at five years, representing an increase of 152% of the 28-day compressive strength. Bureau of Reclamation information indicates that continuously moist-cured concrete that achieves approximately 4500 psi at 28 days will achieve approximately 5700 psi at TAR 1/11/96 1 .3-17

180 days, representing an increase of 127% of the 28-day compressive strength .

The above information is conservative from a time standpoint when compared to biological shield wall concrete which has been in place and gaining strength in a moist condition for the last 27 years. These sources indicate 127% and 129% increase from 28-day to 180-day compressive strength, and a 152%

increase from 28-day to 5-year compressive strength.

The following table summarizes these concrete strength gains with time:

% Strength Increase Source Time of Testing over 28-Day Strength Bureau of Reclamation 180 Days 127%

PCA 180 Days 129%

PCA 5 Years 152%

A conservative value of 140% will be assumed for increase in compressive strength for the biological shield wall concrete from the 28-day strnngth to the present-day strength (approximately 27 years). It is obvious from the strength

_______ incrnas_e_tr_end_Vllitb_timJLi_n the_abov~J:abl~_that a _140% strength increase

• factor is very conservative for the 27-year time periocr:-- - - ~----------------

Concrete Strength Degradation At High Temperature

• A conservative quantification of the decrease in concrete compressive strength due to exposure to high temperatures is presented as follows:

1. Define worst case concrete temperature during annealing transient.

A temporary ventilation system will be installed to support the annealing operation and limit the extent of the concrete temperature transient. This system will consist of an air exhaust fan to provide forced downward air flow through the reactor cavity seal and nozzle penetrations to the annular region between the reactor vessel and the biological shield wall and exhausting through the reactor cavity access tube.

Section 1 .3.C.9 presents the results of the reactor cavity thermal analysis.

This analysis included consideration of the proposed temporary forced air circulation system configuration and confirms the maximum expected cavity seal liner plate temperature anywhere in the vicinity of the inner biological shield wall surface will not exceed 250°F. This temperature, therefore~ will serve as the conservative upper bound for consideration of the maximum possible degradation in the concrete compressive strength that could reasonably be expected to occur during or as a result of the annealing transient.

2. Define the worst-case effects of high temperature exposure on concrete

. compressive strength. This assessment was completed in three steps as follows:

TAR 1 /11 /96 1.3-18

a . Consider the existing aggregate type

• b.

The aggregate used for the Palisades biological shield wall concrete is carbonate based. This type of concrete has relatively good resistance to higher temperature degradation effects.

Consider the moisture content Moisture content of the concrete plays an important part in determining the extent of compressive strength degradation at higher temperatures. Dry concrete generally maintains its compressive strength at high temperatures better than moist concrete. For conservatism, this evaluation considered the concrete to be moist to conservatively maximize the estimate of potential compressive strength degradation.

c. Compare existing concrete properties and proposed annealing temperature limit to available data obtained through an industry literature search.

Test data for similar concrete types exposed to high temperatures provides an estimate for the maximum degrndation in conciete compiessive strength that could reasonably be expected to occur during or as a result of the

_________a_n_n_e_a_l_ing_Qrocess. SQecifically, a conservative decrease of 25% in compressive strength is assumed for unloaded concrete which is considered sealed against moisture loss and exposed to sustained temperatures of 250°F. Most degradation occurs in the first cycle (for cycled temperatures exposure) or during the first 28 days for sustained temperature exposure.

Research has concluded that restrained concrete resulted in an increase while unrestrained concrete resulted in a decrease in concrete compressive strength upon exposure to high temperatures. Biological shield wall concrete under consideration (near the liner) falls somewhere in between each of these ideal conditions. The steel liner will have some restraining effect due to an arching effect. This restraint, as well as restraint provided by the biological shield wall concrete structure itself is neglected in this analysis, thus further addin,g to the conservatism of the 25 % decrease.

It is critical to note that all concrete tested in the industry research results utilized in this evaluation was relatively new concrete, nominally cured for periods ranging from 7 days to 260 days prior to exposure to elevated temperatures. It should be noted further that the longer concrete is cured before exposure to elevated temperatures, the less significant will be the deterioration of the structural properties of the concrete. This adds further conservatism to this analysis when consideration is given to the significant difference between the test specimen curing times of 7 days to 260 days (prior to exposure to elevated temperatures) as compared to the biological shield wall concrete curing time of 27 years.

3. Evaluate the Impact of Exposure to Radiation Total neutron and gamma radiation exposure in units of rads at the biological shield wall concrete inner face are indicated in Table 1 .3.C.3-2. The maximum TAR1/11/96 1.3-19

exposures are at the 0 degree azimuthal angle. For the neutron radiation exposure, units of rads can be converted to neutrons per square centimeter (n/cm 2 ) by dividing by a factor of 3.33E-9. These maximum exposures translate as follows:

Total Neutron Exposure Total Gamma Exposure 9.19E + 9 rad (2. 76E + 18 n/cm 2 ) 1.89E + 9 rad These are worst-case exposures at the biological shield wall concrete inner surface nearest the reactor vessel. The exposures decrease as the depth into biological shield wall concrete (away from the reactor vessel) increases, as indicated in Table 1.3.C.3-2. For example, at 11.6 inches depth into the biological shield wall concrete, exposures decrease to 4.36E + 8 rad (neutron exposure) and 6.03E + 8 rad (gamma exposure).

Various sources exist which discuss effects of neutron and gamma radiation on concrete. The NUMARC Class 1 Structures License Renewal Industry Report indicates that concrete does not begin to experience a loss in compressive strength until exposure exceeds 1E+19 n/cm 2

• The same report indicates that test data implies that gamma radiation dose in excess of 1 E + 10 rads reduces

_ _ _ _ _ _str_engtb_am:La_table in the regort shows no effect on concrete compressive strength until gamma radiation exceeds 5.5E + 9 rads.

Industry research indicates that concrete exposed to 1 E + 1 8 n/cm 2 undergoes

• no significant change in compressive strength. This research also indicates that *some tests resulted in small damage at 2.4E + 21 n/cm 2 at a temperature between 400°F and 1000°F. Other research indicates that the effects of irradiation are small up to integrated neutron fluxes of 1E+19 n/cm 2

• Based on the above worst-case exposures of biological shield wall concrete to neutron and gamma radiation and comparison to industry testing and research, it is concluded that radiation exposures to this point in time have no measurable effect on compressive strength of biological shield wall concrete.

Estimate Post Annealing Concrete Compressive Strength Long-term strength gain of biological shield wall concrete has been conservatively determined to be 140%, which represents an increase from the recorded 28-day compressive strength to the present-day compressive strength approximately 27 years later.

A loss in compressive strength of biological shield wall concrete has conservatively been determined to be 25%, which represents a decrease in compressive strength of present-day biological shield wall concrete when exposed to 250°F during the

• reactor vessel thermal annealing process. Another way of expressing this is to say that biological shield wall concrete will maintain 75% of its present-day compressive strength after exposure to temperatures up to 250°F.

The following analysis determines the minimum 28-day compressive strength required that, when increased due to the strength-gaining effects of continued (27 years) curing under favorable moisture conditions, will still exceed the required TAR 1/11 /96 1.3-20

design strength of 5000 psi even after the strength decrease due to exposure to 250°F:

Assume: y = 28-day compressive strength.

Then: 1.40Y = present-day compressive strength

. 75(1.40Y) = 1.05Y = present-day compressive strength of biological shield wall concrete as affected by exposure to 250 ° F.

In order to maintain required design assumptions of a minimum of 5000 psi compressive strength, the present-day strength of biological shield wall concrete as affected by exposure to 250°F must be greater than 5000 psi.

Therefore: 1.05Y > 5000 y > 4762 psi This means that the 28-day compressive strength of biological shield wall concrete must theoretically be at least 4762 psi. Review of the previous as-placed 28-day

---

compr:essiv.e-strengths_foLbiofo_gicaLsbieJd_w_aJLc_oncrete indicates that the average ------

of the available representative 28-day compressive strengths is 5742 psi, with all individual test results being greater than 5000 psi. This represents an adequate (and conservative) margin of nearly 1000 psi over the requirement of 4762 psi established above.

Conclusions A comparison of the worst-case high-temperature exposure effects on concrete strength (reduction of concrete compressive strength) to the minimum expected concrete compressive strength gain over time was completed to determine a net effect on concrete strength. The results of this comparison are that there is no significant net effect that will alter the ability of the biological shield wall concrete to perform its design basis function. In other words, upon completion of the thermal annealing process, the biological shield wall concrete will still have compressive strength in excess of the minimum required 5000 psi.

It is also concluded that integrated radiation exposures to this point in time had no measurable effect on the compressive strength of the biological shield wall concrete.

During the annealing operation, temperature sensing instrumentation will be placed in reactor cavity regions 2 and 3 (as defined by Figure 1.3.C-1) to monitor the cavity steel liner temperature. This temperature monitoring will provide confirmation of the temperature predictions and will also confirm that the biological shield wall concrete temperature did not exceed 250 ° F.

TAR1/11/96 1.3-21

1.3.C.9 Biological Shield Wall Thermal Analysis The purpose of this analysis is to determine the heat transfer from the reactor vessel and connected nozzles/pipin'g insulation during the annealing process to assess the impact on the biological shield wall concrete. The analysis uses a two dimensional (radial and vertical) steady state thermal model which balances heat gains from the reactor vessel and attached coolant piping with heat losses to the surrounding environment, including the shield cooling system. The model also includes the cooling effects of annulus air flow created by the temporary forced air circulation system. This system draws containment air down through the refueling seal ring gap and the primary coolant piping biological shield wall penetrations, through the annulus between the biological shield wall liner and the reactor vessel, and through the reactor cavity access hatch located below the reactor vessel.

This thermal analysis serves as the basis for the stated maximum expected concrete temperature considered in the concrete evaluation of Section 1.3.C.8.

Assumptions The following assumptions were made in the development and use of the model used for this analysis.

1. The reactor vessel, air gap, insulation, annulus air gap, and biological shield wall can all be modeled as concentric cylinders.

• 2. Emissivities of the insulation, reactor vessel, and bioshield liner surfaces adjacent to air gaps are based on observations taken during the 1995 refueling outage. The emissivity of the reflective insulation is based on these observations and the sensitivity analysis done to characterize the insulation.

3. The interior surface of the Transco insulation is assumed to have the same emissivity as the exterior surface.

4. The maximum shield cooling system water temperature during the annealing process is 90°F.

5. The maximum containment temperature during the annealing process is 100°F.

6. The cavity is cooled by a fan which pulls air through the seal ring gap and the gaps in the primary coolant loop piping penetrations. This forced circulation air cooling is based on fan suction through the reactor cavity access tube. A range of air flows are evaluated in the analysis.

7. Sufficient horizontal insulated access panels (convection barriers) will be opened to accomodate air circulation during annealing.

8. The bottom head of the reactor vessel is insulated with 4" Nukon insulation or equivalent.

9. All openings into the reactor cavity, other than those mentioned above, are sealed.

TAR 1/11 /96 1.3-22

10. A vertical hot gap of 0.5" (minimum) exists between the reactor cavity seal gutter and the RV head flange to provide the required minimum flow area through the reactor cavity seal during the annealing process.

11. The surface emissivity for the Transco insulation is chosen as 0.5. The surface emissivity for the biological shield wall liner is chosen as 0.9. Both are intended to be conservatively high.

Analysis Conservatism

1. The liner plate temperature is assumed to be the concrete temperature. The temperature drop that may exist between the liner plate and the concrete surface is conservatively ignored.

2. The liner plate in Region 3 is assumed insulated on the concrete side. This results in a maximum liner surface temperature in this region.

3. The "worst case" reflective insulation transmittance as developed in the insulation characterization analysis is used for this analysis.

4. The convective heat transfer coefficients used for the cavity heat transfer

,_ _ _ _ _ _ _ between_the_insulatio_[Lsudace_and_tb_e_bfoJo_gi_c_aLshieJd_walUio.e.Ls_u_rfa___c_e-to~------~

air flow model are approximately 25% less than the literature suggests.

5. The RV annealing temperatures used for the analysis are the maximum

• steady state values (900°F in the beltline region) .

Thermal Model Description TSAP (Thermal System Analysis Program), Version 3A was utilized to develop a two dimensional (radial and vertical) steady state thermal model which incorporates forced air circulation through the reactor cavity annulus area. The model includes the reactor vessel wall, the air gap between the reactor vessel wall and the Transco insulation, the Transco insulation, the annulus air gap between the Transco insulation and the biological shield wall liner, the biological shield wall liner, the biological shield concrete wall, the shield cooling system embedded within the biological shield wall concrete, the p_rimary coolant loop piping and pene-trations, the bottom reactor vessel head (insulated), and the annulus forced circulation air flow described above. The general model is depicted in Figure 1.3.C.9-1.

Reactor Cavity Region Variations - Vertical The general model shown in Figure 1.3.C.9-1 incorporates the division of the reactor cavity area into three distinct regions (Region 3 - top, Region 2 - middle, and Region 1 - bottom) as defined by Figure 1.3.C-1. Accordingly, there are unique model considerations for each region. -

The model for Region 3 is broken down in the vertical plane as follows:

• The (00) level nodes start at the reactor vessel flange and extend down to the point where the vessel wall thickness decreases above the nozzles. This TAR 1/11 /96 1.3-23

region includes the air flow from the gap between the refueling seal and reactor vessel flange .

The ( 1 0 - 30) levels are equally spaced between the (00) level to the interface with Region 2. This area of Region 3 contains the nozzles and the air flow paths in the gaps between the primary coolant loop piping insulation and the penetration shield blocks.

• The heat load from the nozzles (Node 195) is modeled using the same insula-tion characteristics (Nodes 195-197) used for the vessel.

• The primary coolant loop piping (Node 600), insulation (Node 610) and air flow through the penetrations (Node 620) are conservatively modeled.

The model for Region 2 is broken down in the vertical plane into equally spaced

. nodal levels (40 - 90) between the 614. 7' elevation down to the horizontal insulation panels which separate Region 2 from Region 1 (Elevation 601.5').

The model for Region 1 consists of the (99) nodal level and represents the area below the horizontal insulation panels that separate Region 2 from Region 1 down to the cavity floor (Elevation 591' 3"). This model contains only the air flow from

_____Begion-2_and_tb.e_b_e_atJo_aJ:i from the bottom head of the reactor vessel.

Reactor Cavity Region Variations - Radial

• There are significant radial variations in the upper regions of the reactor cavity that are accounted for in the heat transfer model developed for each region. Figures 1.3.C.9-2 and 1.3.C.9-3 show the typical radial cross sections of nodes and conductors at elevations corresponding to Region 2 and 3 respectively. Note that the bioshield liner is assumed insulated in Region 3.

Reactor Cavity Air Flow A temporary forced cooling system will be installed to draw air from general areas of containment into the reactor cavity and out the reactor cavity access tube through a filtration system. This direction of flow was chosen in order to minimize potential contamination of general areas of containment that could result from blowing air up from the vessel lower head and out into containment. All flow paths such as the cavity flooding lines and drain line to the sump will be sealed to ensure all flow is from the refueling seal ring gap and the coolant loop piping penetrations.

The fan will take suction through the 30" reactor cavity access tube at elevation 592' - 3". The temporary system will provide a design nominal flow of 5,000 cfm air flow.

The air flow distribution between the reactor vessel flange and the primary coolant loop piping penetrations is calculated based on the existing open areas. The result-

• TAR1/11/96 1.3-24

ing air flow split between the refueling seal and the primary coolant loop piping penetrations is calculated to be:

Refueling seal ring gap { 1) 1,350 CFM { 27%)

Piping penetrations (6) 3,650 CFM { 73%)

Total flow 5,000 CFM (100%)

The direction of flow chosen complicates the selection of heat transfer coefficients for the liner and insulation. The air flow regimes are classified as "mixed" counter current flow since the tendency exists for the buoyancy effects to create a natural convection film along the liner and insulation surfaces while the forced flow tends to disrupt this film with all flow eventually being removed from the reactor cavity through the reactor cavity access tube. These effects are taken into account in the determination of appropriate heat transfer film coefficients.

Reactor Vessel Insulation Characterization The Transco RV insulation is characterized based on the in-plant inspections and measurements conducted during and after the 1995 refueling outage. The effectiveness factors for the insulation aie calculated based on tempeiatuie measurement data corrected for measurement bias and conservatively chosen emissivities of the reflective insulation and the biological shield wall liner Q_la_t_e_.- - * - - - - - -

The reflective insulation is modeled in the same manner as above with the effectiveness factor used to modify the Transco design data calculated as a

• function of average reactor vessel temperature and air temperature. This is determined at each vertical node in the model.

The RV bottom head is modeled as insulated with 4 inches of Nukon insulation or equivalent.

Heat Load from Nozzles The nozzles are modeled separately with convection to the air in nodes 310, 320, and 330. The nozzle temperature is an output of the reactor vessel thermal analysis discussed in Section 1. 7. Heat input from the nozzles into the surrounding concrete is calculated within the reactor cavity thermal analysis using a model subroutine based on heat transfer through the 4" of reflective insulation. The temperature drop through the insulation is handled in the same way as the temperature drop through the reactor vessel insulation with the same effectiveness factors. The film coefficient is based on the film coefficient calculated for the reactor vessel insulation wall to the air in each node.

Heat Load from Primary Coolant Loop Piping The primary coolant loop piping heat loads were calculated in a similar fashion to the RV insulation. However, all of the heat transfer from the piping is modeled as transferring directly to the air flowing through the penetration.

TAR 1/11 /96 1.3-25

Thermal Analysis Results Sensitivity studies were performed with this model to determine a nominal temperature profile in the reactor cavity as well as a maximum temperature profile of the biological shield wall liner. Parameters varied included air flow, air flow distribution, reactor vessel insulation effectiveness and variation of thermal conductivity of the biological shield wall liner to shield cooling system and biological shield wall boundary.

The results of the thermal analysis are presented in Tables 1.3.C.9-1 and 1.3.C.9-

2. These two tables demonstrate the cavity temperature ranges calculated.

Table 1.3.C.9-1 presents the cavity temperatures assuming a realistic thermal transmittance for the reflective insulation. This thermal transmittance was based on the insulation and cavity liner temperature data taken following the 1995 refueling outage. The maximum calculated effective thermal conductivity for the insulation with a vessel temperature of 900°F is 1.174 BTU-IN/HR-FT2-°F. The maximum calculated liner temperature is 188°F in Region 3. The exit air temperature is 150°F. These temperatures are considered to be a realistic assessment of temperatures that will be experienced during annealing.

Table 1.3.C.9-2 presents a conservative estimate of the cavity temperatures. The reactor vessel insulation transmittance is based on a conservative extrapolation using the temperature data taken following the 1995 refueling outage. This assessment used only the temperatures taken from the vessel insulation wall and the air. The cavity liner temperatures were calculated based on the conservative assumption that the temperature drop from the insulation surface to the air was the same as that from the air to the liner. This maximized the effective thermal conductivity calculated. When this insulation thermal transmittance was used in the cavity model with the forced circulation the maximum calculated liner temperature is 219°F in region 3. The maximum calculated effective thermal conductivity for the ins.ulation with a vessel temperature of 900°F is 1.850 BTU-IN/HR-FT2-0F. The exit air temperature is 164°F.

Additional cases were evaluated which varied the flow split between the primary coolant piping penetrations and the refueling seal gap. The resulting maximum liner temperatures were within a few degrees of the maximum liner temperatures presented in Table 1.3.C.9-2.

As can be seen from the results, the biological shield wall concrete temperature will remain well below 250°F during the thermal annealing process. Considering the conservative methods used in the analysis provides reasonable assurance that the concrete temperature analysis fully bounds the expected thermal annealing conditions.

1.3.D Description of Primary Coolant System Piping The primary coolant system piping consists of those sections of pipe that are used to interconnect the reactor vessel, two steam generators and the four primary coolant pumps. The pressurizer is connected to one of the RV outlet pipes by means of a surge line. The primary coolant system piping layout is shown in

' Figures 1.3.A.1-1 and 1.3.A.1-2. Each pipe and elbow section is fabricated per SA TAR1/11/96 1.3-26

516 Grade 70 with roll-bonded clad of stainless steel (Type 304 L) on the interior surface per SA 264. The piping was shop-fabricated and shop-welded into subassemblies to minimize the amount of field welding. The straight lengths of piping were formed by rolling the roll-bond clad plate into a cylinder and welding along the one longitudinal seam. The elbow assemblies were fabricated by forming two 180-degree segments and welding the halves along two longitudinal seams.

The pipe subassembly was formed by butt welding two or more sections of pipe and back-cladding of the weld joint. Fabrication of piping and subassemblies was done by shop personnel experienced in making large heavy wall welds. Welding procedures and operations met the requirements of the ASME B&PV Code,Section IX, 1965, W65a, see section 1.3.D.9.1. All welds were 100% radiographed and magnetic particle or liquid penetrant tested to the acceptance criteria of the ASME B&PV Code, Section Ill, Class A, 1965, W65a. All primary coolant piping penetrations were attached in accordance with the requirements of the ASA 831 . 1, Power Piping Code, 1955. Heat treatment of the piping was performed after all fittings were assembled, nozzle bosses welded and back-cladding was completed.

Field welds were post-weld heat-treated to the requirements of the ASME B&PV Code, Section Ill, Class A, 1965, W65a.

The Palisades PCS opeiates at nominal conditions of 2060 psia vvith a hot leg temperature of 580°F and cold leg temperature of 533°F. The design temperature

______an~d. Rressure are tabulated in Table 1.3.D-1 along with the material and nominal size of the primary coolant system piping. The Palisades PCS piping was designed in accordance with ASA 831 . 1-1 955 Code for Pressure Piping, which specifies limits on stress due to thermal expansion of the primary coolant system. These

• • expansion stresses are a function of the resultant bending and torsional moments produced by thermal expansion, as well as the geometric properties of the pipe and fittings.

The RV primary inlet and outlet nozzles were designed in accordance with the 1965 Edition of the ASME Boiler and Pressure Vessel Code, Section Ill (through Winter 1 965 Addenda).

The stress limits to. be used for the annealing process are in accordance with the design criteria of 3Sm for the primary plus *secondary stresses. Each section of the primary coolant system piping is described in detail below. The stress analysis is described in Section 1 . 7.

1.3.D.1 Reactor Vessel to Steam Generator Piping Two pipe sections (42 inches l.D.) were fabricated to connect the RV outlets to the two steam generators. Each pipe section was fabricated from SA 516 Grade 70.

Each section consists of a straight length of pipe butt welded to an elbow. Each section of piping contains nozzles described in Table 1.3.D-2.

1.3.D.2 Outlet of Primary Coolant Pumps to Reactor Vessel Four pipe sections connect the primary coolant pumps to the RV inlets. The pipe sections have 30 inches l.D. Each piping section was fabricated from SA 516

. , Grade 70. The pipe sections are shown in Figures 1.3.A.1-1 and 1.3.A.1-2. Each

,. section of piping contains nozzles described in Table 1.3.D-2.

TAR1/11/96 1 .3-27

1.3.D.3 Outlet of Steam Generators to Primary Coolant Pump Suction Piping Four pipe sections connect the steam generators to the pump suction piping. Each 30-inch l.D. pipe section was fabricated from SA 516 Grade 70. These sections each consist of a straight length butt welded to an elbow. There are no nozzles in these sections.

1.3.D.4 Primary Coolant Pump Suction Piping Four pipe sections connect the steam generator outlet piping to the pump suction elbows. The pump suction elbow was supplied as an integral part of the pump.

Each piping section was fabric.ated from SA 516 Grade 70. Each section consists of a straight length of pipe butt welded to an elbow. Each section of piping contains nozzles described in Table 1.3.D-2.

1.3.D.5 Primary Coolant System Piping Supports and Restraints The PCS piping has no individual supports on either the hot leg or cold leg pipes.

The hot legs are supported by their connections to the RV outlet nozzle and the inlet nozzle of the steam generator. The cold leg piping is supported at the RV inlet nozzle, steam generator outlet nozzle, and the primary coolant pump suction and

-~---elisehaF9e-n0z-zles....- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

1.3.D.6 Primary Coolant Pump Support

• The supports and stops for the primary coolant pumps control movement in the horizontal and vertical planes during seismic and pipe break conditions, but allow normal operating movement.

The pump and motor assembly is supported on four feet welded to the pump casing. Each support foot sits on top of a cylindrical support pad, which in turn sits on top of a circular lubricated plate in a cylindrical c*avity. The four cylindrical cavities are part of a support frame. A single large stud extends downward from each foot through an oversize hole in the floor of the cylindrical cavity, and a nut and heavy washer on the underside of the cylindrical cavity prevent uplift of the pump. Another circular lubricated plate is provided between the heavy holddown washer and the underside of the cylindrical cavity for unrestrained normal operating movement. A cylindrical spacer on the stud assures a small vertical gap between the holddown washer and the underside of the cylindrical cavity to prevent binding during normal operation movement.

Guide shims are fastened to the cylindrical cavities (sockets) in the gap between the support pad and the socket. These shims prevent motion of the pump perpendicular to the direction of normal operating movement. A block is fastened to the socket in this same gap on the side away from the reactor vessel to act as a stop in the event of a pipe break.

'To reduce system stresses resulting from vertical thermal growth of the RV nozzles, the primary coolant pumps and cold leg pipes are cold sprung. After all primary coolant piping is welded and stress relieved, the pump and support frame are raised vertically from their nominal cold position by an amount equal to one half the expected thermal growth and maintained in that position by shimming under TAR1/11/96 1.3-28

the hot leg axis causing unequal vertical displacements between the casing lugs and socket mating surfaces. In the final installed cold position the two supports farthest from the reactor vessel have a gap between the support pad and socket while the two supports nearest the reactor vessel have gaps between the anchor washer and the bottom surface of the socket. The total gap at each support point is set during initial installation and maintained by the spacer between the support pad and the anchor washer.

The primary coolant pump support configuration is depicted in Figures 1.3.D.6-1 and 1.3.D.6-2. Travel limit values for the primary coolant pumps are presented in Section 1 .4.

1.3.D.7 Steam Generator Supports The steam generator is supported at the bottom by a conical skirt welded to the lower head. At the bottom of the skirt is a flange which is bolted to a heavy cast steel sliding base. Low friction cylindrical bearings similar to those used under the reactor vessel supports are located at four locations between the sliding base casting and steel plates fastened to the support structure. These bearings are the sliding interfaces which aiiow horizontai motion of the steam generator parailel to the hot leg axis. Guide plates are welded to the top surface of the embedded

---

support-plates-tbaLtbe-bear:ings-slide-on-to-pr:evenLtbe-sliding-beaf"ings-fr:om-sliding----~

out of the sockets.

Cast openings in the sliding base act as keyways for two keys embedded in the

• support structure. Both keys prevent horizontal motion of the steam generator perpendicular to the hot leg axis. The key closest to the reactor vessel also acts as a stop for motion of the steam generator along the hot leg axis during a hot leg pipe break. Additional cast openings in the sliding base accept vertical anchor bolts embedded in the support structure. Plates with slotted holes are welded to the sliding base at the top of the cast openings. These slotted holes allow movement of the sliding base relative to the anchor bolts during any horizontal steam generator motion. Nuts at the top of the anchor bolts bear against these plates to restrain upward motion of the steam generator during seismic or pipe break conditions. During normal operation there is a gap between these nuts and the plates on the sliding base to prevent bending of the anchor bolts. This gap is set to a specified value which is verified during hot functional testing.

In the keyways, low friction expansion plates are used to minimize resistance to thermal motion if the sliding base contacts the embedded keys. The stop gap between the sliding base and the key closest to the reactor vessel must be shimmed to a specified value which is verified during hot functional testing.

Under seismic and pipe break conditions, horizontal support at the top of the steam generators is provided by two keys integrally welded to the steam generator and eight hydraulically interconnected snubber assemblies. These supports allow normal operating motion of the steam generator parallel to the hot leg. The keys restrain the steam generator in the direction perpendicular to the hot leg axis and the snubbers act in the direction parallel to the hot leg axis .

• The keys mate with keyways attached to the building primary and secondary shield wall. Clearances are required at the ends and sides of the keys to preclude binding TAR 1/11/96 1.3-29

during normal operation and any other steam generator motion parallel to the hot leg axis.

Each snubber assembly consists of a hydraulic cylinder with the piston rod end connected to the steam generator and the fixed end fastened to the building primary or secondary shield walls.

The snubber pin-to-pin length must be accurately set and verified during hot functional testing to ensure that the piston has sufficient clearance to prevent it from hitting either end of the cylinder under normal operating seismic or pipe break conditions.

The steam generator support configuration is depicted in Figures 1.3.D. 7-1 and 1.3.D. 7-2. Travel limit values for the steam generators are presented in Section 1 .4.

1.3.0.8 Reactor Vessel Supports Vertical support of the reactor vessel is provided by three equally spaced bearing assemblies. One assembly is under the loop 2 hot leg nozzle and the other two are under each of the loop 1 cold leg nozzles. A pad is welded to the underside of e-----~each_o_Lthes~e_thrne_nozzles._Tbe_bearings_are_mounted_to_tbe_bottom_of_tbe_pads------1 and the sides of the pads mate with the lateral support structure to provide horizontal support for the reactor vessel under seismic and pipe break conditions.

The bearing assemblies consist of a bearing socket bolted to the underside of the nozzle pad that mates with a sliding bearing having a convex cylindrical shape .

The bottom of the sliding bearing is flat and mates with a plate attached to the vertical support structure. All relative horizontal movement between the reactor vessel and the building occurs at this flat surface. Both upper and lower mating surfaces of the sliding bearing are lubricated. Low friction expansioin plates are bolted to the sides of the lateral support structure to allow free radial expansion of the reactor vessel. With the use of shims, a gap is established between the nozzle support pads and the expansion plates to preclude bimfing during normal operation.

The support configuration is depicted in Figure 1 . 3. D. 8-1 . This arrangement permits radial thermal grovvth of the reactor vessel while maintaining it centered and restrained from movement resulting from seismic forces.

1.3.0.9 NOE Test Results 1.3.0.9.1 Preservice NOE Results on Primary Coolant System Piping The SA 51 6 Grade 70 plate used in the manufacture of the Palisades PCS piping was acquired by CE or by subcontractors in accordance with CE purchase specifications and ASME Code requirements. Receipt inspection was performed at which time the material certifications, heat treatment certifications, ultrasonic inspection results, dimensions (including clad thickness) and supplier identification, as applicable, were verified. The plate was further processed (e.g., formed, machined, heat treated and inspected) prior to welding. There are no records of weld repairs made to the plate. (See following discussion on weld inspection).

The lack of plate repair records does not necessarily indicate that no repairs were made.

TAR 1/11/96 1.3-30

Primary coolant piping sub-assembly fit-up, welding and inspection was performed for CE by several subcontractors. This included both 30 inch and 42 inch diameter straight pipe lengths and elbow assemblies. The subcontractors performed the welding, the in-process magnetic particle and dye-penetrant inspections, and the final radiographic examination on these sections. CE performed the fit-up and welding of the elbow to the straight pipe section, and the fit-up and welding of the nozzles to the pipe section. CE also performed the inspections of those welds and the final surface examinations (i.e., dye penetrant and magnetic particle) of the entire piping sections. Some of the subcontractor radiographs and radiographic acceptance sheets were provided to CE and, from the information provided, it was inadequate to determine the weld repair locations or orientation of the radiographic layout to kno,wn pipe locations (such as axis lines). Radiographs and radiographic acceptance sheets for NOE performed by CE were retained, but the location and orientation of the repairs or as-left indications could not be determined in some cases. The general location of the recordable as-left radiographic indications is listed in Table 1.3.D.9.1-1. The general location of the recorded weld repairs is listed in Table 1.3.D.9.1-2.

1.3.D.9.2 lnservice Inspection Results on Primary Coolant System Piping From the ASME Section XI required inservice inspections, some embedded and

---

sudace-indications_hav.e_heen_s_e_eo_in_s_o_me of the PCS Qiging welds that have been selected for inspection. These indications were all found to be acceptable in accordance with Section XI of the ASME Code. These indications are located in welds outside of the zone affected by the thermal annealing process, e.g. not in

• the pipe welds adjacent to the RV nozzles .

The inservice inspections performed included surface and volumetric examination techniques. Tables 1.3.D.9.2-1 through 1.3.D.9.2-6 include the weld number, a description of the weld, the inspections performed, the outages in which the inspections were performed, and the inspection results for the six lines of piping.

Figure 1.3.D.9.2-1 provides a description of the lines and identifies the specific welds for each line.

1.3.E Physical Descriptions of Other Equipment and Instrumentation That Could Be Affected by Thermal Annealing Other equipment and instrumentation in and around the reactor cavity that could be affected by thermal annealing unless removed or precautions otherwise taken are discussed in the following sections.

1.3.E.1 Neutron Detector Instrumentation The neutron detectors are installed in the detector wells in the biological shield.

Source range and wide range detectors ( 1 /3, 2/4) are located in windows at the inside wall of the biological shield and are subject to radiant heat transfer from the reactor vessel and the convective heat transfer in the reactor cavity annulus.

Power range detectors 5, 6, 7 and 8 are located within the biological shield and are shielded from radiant arid convective effects by the biological shield liner and an inch of concrete. The neutron detectors are used to detect neutron flux through the RV shell during power operation and need not be in place during the thermal annealing operations. However, there is some concern for damaging the TAR1/11/96 1.3-31

instrumentation during removal from the detector wells as well as the extra radiation dose that would be required to remove and replace the detectors.

Therefore, leaving the instrumentation in place during annealing is preferred if the detector temperatures can be maintained below the specified maximum temperatures for the detectors.

Source, wide range and power range detectors can withstand operating temperatures up to 300°F. Therefore, the maximum temperature limits for the detectors will not be exceeded during annealing.

The temperature vs. time profiles during annealing will be used to update the EQ lifetimes of the source and wide range detectors.

1.3.E.2 Primary Coolant Flow 'and Temperature Measurement Instrumentation The taps in the PCS piping for the flow and temperature measurement instrumentation are located outside the biological shield and at least eight feet down the pipe from a RV primary nozzle. The temperatures at the instrumentation during the annealing operations will remain well below the normal operating tempernturns. Therefore, there is no concern for leaving .the instrumentation in place during the annealing. The thermal profile down the PCS piping is quantified in Section 1 . 7.

~~--~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~-

1.3.E.3 Reactor Cavity *Relief Tube Rupture Disc The rupture disc in the 30-inch diameter relief tube is located under the Loop 1 B primary cold leg pipe outside the bioshield. The relief tube angles downward through the concrete to intersect the reactor cavity access tube midway through the biological shield. The polyethylene rupture disc may be subjected to convective heat transfer through the relief tube and primary piping penetration. Other heat sources are created by conductive heat transfer down the cold leg pipe and along one of the RV support beams. The temperature of the rupture disc during the thermal annealing, however, will be less than its temperature during normal reactor operation since it is located at least seven feet down the pipe from the RV inlet nozzle outside the biological shield.

1.3.E.4 Electrical Cable in the Refueling Cavity and on the Operating Deck Electrical cable for the RV systems within the refueling cavity will be positioned to avoid adverse effects from hot air in the refueling cavity. Any cable leads for reactor instrumentation, control rod drive, head vent system, RV level monitoring system, etc. that remain in the cavity or on the operating deck (649 foot level) adjacent to the cavity after the RV head is removed will be remote from the reactor vessel, RV top cover, hot air vents and detector well openings. Electrical cable is typically limited to temperatures below 160°F. However, relocation of electrical cable from the refueling cavity will occur as is normal for any refueling outage.

1.3.E.5 Access Tube The access tube is a 30-inch diameter penetration running horizontally through the concrete biological shield into the reactor cavity at the reactor cavity floor level.

The access tube is lined by a cylinder of 1 /4-inch thick carbon steel plate which is TAR 1/11 /96 1.3-32

an extension of the biological shield liner. The access tube opening in the access room outside the biological shield is closed by a bolted blind flange cover. SCS piping penetrates the access tube behind the flange and passes through the access tube into the reactor cavity.

The access tube will realize a thermal load from the thermal annealing. The air temperature will be less than the upper bound temperature established for the structural concrete for the short term of the annealing. Therefore, the thermal load will have no adverse effects on the access tube liner and the surrounding structural concrete.

1.3.E.6 Reactor Containment Building Equipment Hatch The equipment hatch is the 12-foot diameter port into the reactor containment building (RCB) which provides the access for larger equipment to be delivered into the RCB from outside.

The hot air to and from the heat exchanger inside the RCB will be carried by thermally insulated ducts which are to be routed through the equipment hatch to heat the reactor vessel during the thermal annealing operations. The equipment hatch and the surrounding structural concrete of the RCB will be subjected to a

_______tbermaLload_frnm_tbe_ins_ulate_d_d_uc_t_s_.___1he ducJ_ insL!_)gition ~nd additional JJrotective measures will be sufficient to maintain the temperature of the hatch an(f the -----

surrounding concrete below the 150°F normal concrete temperature limit during the thermal annealing operations. Therefore, due to heat dissipation there will be no

• adverse effects on the hatch and the surrounding structural concrete .

1.3.E.7 Neutron Detector Well Seal Plates The seal plates over the neutron detector well openings in the refueling cavity will be removed for the thermal annealing operations. This will protect the plate seals from the elevated temperatures.

1.3.E.8 Primary Coolant System Loop Piping Connections Several piping systems have connections at branch nozzles in the PCS pipe. The piping connecting into the branch nozzles includes the surge line, the shutdown cooling sy.stem, the safety injection system, the charging system, letdown and drain lines, spray lines, and sampling lines. All of the branch nozzle connections are located outside the biological shield and at least eight feet down the PCS pipe from the RV inlet or outlet nozzle. All of the branch nozzle connections in the hot legs are closer to the steam generator than to the reactor vessel, and all of the connections in the cold legs are closer to the primary coolant pump.

The temperatures at all of the piping connections during the thermal annealing process will remain well below the normal operating temperatures. The radial displacements at the connections during annealing are comparable to or less than the radial displacements during normal operation since the radial displacements at the steam generators and primary coolant pumps during annealing are less than the radial thermal displacement during normal operation. Because the vertical displacements at the piping connections during annealing are small, and the piping systems flexible and designed for both thermal and seismic loading during normal TAR 1/11 /96 1 .3-33

operation, neither the piping systems nor branch nozzles will be overstressed due to restrained thermal displacement during the annealing operations.

1.3.F Description of Overall Containment The reactor core is to be completely unloaded and transferred to the auxiliary building spent fuel pool racks for storage during the annealing. The annealing activities will have no impact to the fuel unloading/loading, transferring and storage processes which will be performed utilizing normal administrative control procedures for fuel movement and storage. However, the annealing will have an impact on the outage activities particularly within the reactor containment building due to the equipment movement, assembly/disassembly and testing required for the equipment described in Section 1.6, Thermal Annealing Equipment. The radiological impacts to personnel during the outage are discussed in Section 1.9, ALARA Considerations.

While the heat exchanger test and actual anneal ar'e ongoing, there will be heat lost from the heating system ducting and heat exchanger assembly itself. The resultant heat load on the reactor containment building cooling system will be less than during plant operation. Locally, the heating system ducting and heat exchangei assembly impact on plant equipment and structures will be addressed by utilizing

~---~a=d=d~it~io~n=a~I grotective measures such as insulation or increased standoff distance.

Combustibles on the operating floor and other affected areas will be controlled utilizing normal administrative control procedures for hot work permits.

• The amount of working area available to prepare for and perform the annealing must be maximized. The equipment associated with the RV disassembly and maintenance activities will be removed from the reactor containment building operating floor or will be stored in alternative areas after use. Examples of this equipment might be the ISi tools/components, RV studs, nuts, washers and.

tensioners, top section of the upper guide structure (UGS) lift rig and in-core instrumentation (ICI) access platforms, core support barrel (CSB) lift rig, etc.* The plant equipment and components that interfere with or may be affected by the annealing process will be removed during annealing preparations. Examples of this equipment include the RV wall surveillance capsules, stud hole plugs, reactor cavity seal, RV guide pins and sleeves, etc. In addition, equipment to support the annealing will need to be installed. Examples of this equipment include additional CSB stora*ge pads, internals shielding components, reactor cavity supplemental cooling, etc. Special provisions will be required for the storage of the RV internals components. Activities performed in support of annealing on the reactor containment building operating fl9or and other affected areas will be controlled utilizing normal administrative control procedures for heavy loads movement and storage.

1.3.F.1 Internals Storage During a normal refueling outage, the CSB and the UGS are stored under water in*

the refueling cavity. During the RV annealing project, however, these components will be stored in the refueling cavity with the cavity drained. This requires the design of temporary shielding that can be installed after the components are moved to their storage location and prior to draining the cavity. The shielding will be removed after the cavity is refilled and before the components are replaced in the TAR1/11/96 1.3-34

reactor vessel. Normally, the components are stored on pads at opposite sides of the reactor vessel; however, for this project they will be stored together with the UGS placed inside the CSB at the approximate location of the existing CSB storage pads.

1.3.F.2 Shielding Design Basis Measurements of the contact dose rates on the components were taken in June, 1995. These results were used to determine a source term appropriate for the shielding design. Since the measurements were taken with the components stored under water, the values were corrected to account for dry storage during the annealing project where the shielding effect of the water would not be present.

For the purpose of defining a gamma energy spectrum, the source term was conservatively assumed to be comprised entirely of Co 60 with 1.173 MeV and 1.332 MeV gammas. -

The maximum contact dose rate measured on the outside surface of the CSB was 15,000 R/hr at the lower half of the CSB corresponding to the core centerline elevation. The values at the upper part of the CSB were measured at less than 100 R/hr. These values were measured with the CSB stored under water. The main contributor to the surface dose of the lower portion of the CSB is the core shroud.

The- d r:.y-dose __ rate_focthe _CSB_is_conser_vati~eb/_calc_u late_d_to__b_e_j__O_Q,_O_QO_BLh~r*~~-- -----~

The UGS had relatively low dose rates along most of its periphery (2 to 3 R/hr) except in the area of the fuel bundle alignment plate (bottom portion) where the dose rates range from 100 R/hr to a maximum reading of 1800 R/hr. The dry dose

• rate for the UGS is calculated to be 1 971 R/hr.

The dose rate target for the shielding was established as a calculated dose rate of 500 mR/hr or less in the refueling cavity work areas and 60 mR/hr or less at the 649 foot elevation. In designing the shielding for the core support barrel (CSB) and upper guide structure (UGS), a dose allocation of 20% of this value is allotted for dose contribution from other sources, i.e. the reactor vessel. The work area in the refueling cavity is defined to be at elevation 624'-6" in the cavity seal area and on top of the annealing reactor vessel top cover (RVTC). The limiting work area will be the shielding to cavity seal interface region. Another limiting work area is on the 649 foot elevation and is defined as the travel route from the equipment hatch, along the west side *of the refueling cavity, to the staging area. Shielding design will comply with Palisades administrative procedures. The final design and installation of the shielding will be completed with the goal of maintaining personnel exposures as low as reasonably achievable (ALARA).

The resulting internals shield design for the annealing project is summarized as follows:

• vertical shield on the east and south sides of the CSB/UGS storage location extending up from the refueling cavity floor;

• horizontal plate at elevation 649' covering the open areas of the refueling cavity on the north and west sides of the storage location;

• vertical shield mounted on the horizontal plate (see above);

TAR 1/11/96 1 .3-35

• horizontal top cap extending over the CSB/UGS;

• shielded enclosures for the eight UGS instrumentation stalks; and

• interface provision for airborne contamination control equipment.

1.3.F.3 Internals Storage and Shielding Configuration The process for the movement and storage of the UGS in the CSB in the refueling cavity will reduce the overall cost, schedule and risk of internals damage to Palisades. In addition, this process provides some self-shielding for the dry internals storage in the refueling cavity.

The storage and shielding of the internals is based on dry storage in the reactor cavity. It is believed this shielding approach is the best solution for the following reasons:

• Allows for dry storage of internals, thereby simplifying the design by not requiring redundant leak-tight seals.

• Shielding will be removed from the site after completion of annealing

~~~~~~~~grocess. No storage facility is regu ir_e_d_*~~~~~~~~~~~~~~~~~~~~~

• There are no major welded attachments or permanent modifications to the existing refueling cavity floor structures .

• Preparations for the RV annealing program will commence in the refueling cavity floor prior to fuel off-load. The temporary storage pads are placed on the refueling cavity floor and leveled as needed. The permanent storage pads are not modified or removed. This operation shall take place while the cavity is still dry. The temporary storage pads are required to clear the cavity seal area. The storage pads will not exceed the height of the existing CSB storage pads. This eliminates any interference from temporary structures during the fuel off-load. After the temporary CSB storage pads are in place, the UGS and CSB movement can begin, which are the same as standard outage operations. The upper guide structure is moved to its permanent pads in the cavity southw~st corner. After the fuel is unloaded, the CSB is pulled from the reactor vessel and placed into the CSB storage stand.

1.3.F.4 Shielding System Design After the CSB is placed on the pads, the shielding panels will be erected over the bottom support frame placed in the refueling cavity. The shielding panels are locked in place using the channel frame along the cavity wall and tie-brace members. The shielding panels have locating devices on the lower edge for proper fit and beveled edges to prevent streaming. The panels are designed to have allowance for the cavity draining. The support frame will be designed to distribute the shielding load over the cavity floor area without exceeding floor load capacity.

The shielding structure will be designed and installed such that no welding or direct attachment to the refueling cavity liner is required .

TAR 1/11/96 1.3-36

The shielding consists of lead sealed in water tight steel container panels to preclude any water contamination and any mixed waste disposal concerns. The shielding containers will be made from carbon steel with a protective coating for ease of decontamination and removal from site.

The shielding material selection and its thicknesses will be designed in accordance to the shielding design basis. All shielding will be of modular panel construction as much as possible for the ease of installation from the operating floor elevation 649' -0 into the flooded refueling cavity. The shielding functional design will minimize installation time and total absorbed dose to meet the ALARA requirements. The shielding panel joints are designed such as V-shaped groove vertical joints and semi-circular joints to reduce the potential for radiation streaming from highly radioactive sources to accessible areas within containment. These joints will be designed, assembled, fitted and verified prior to shipment. The streaming or "hot spots" can be further reduced by selectively attaching lead blankets to achieve dose rate goals.

The shielding wall will be placed on the sides of the CSB approximately 5 feet high above elevation 649' - O". The shielding wall around the CSB will be approximately 29'-6" high above the cavity floor and in three modular panels. The panel thicknesses will vary from top to bottom panels in lead or equivalent lead shots in

---

containeLpanels. __Tbe_horizontaLtop_cap___wilLb_e_s_te_el_plate_(p_r:__~gµivale01__J~ad) over the entire CSB flange, and cover over the entire west cavity corner. The in-core instrumentation guide tube stalks will be shielded using steel enclosures, which will be supported on the flange shielding cover and its supporting frame. The shielding

• modules and structural support frame components will be designed to allow their transfer through the equipment hatch using the trolley system that is currently in place.

1.3.F.5 UGS/CSB Lift Shielding will be provided at the operating deck to protect personnel during the movement of the UGS and CSB and prior to installation of the shielding. The shielding will be supplemented as necessary, per Section 1.9.C, to meet radiation levels near personnel of not more than the 60 mR/hour target while performing the UGS lift.

The UGS will be lifted from the UGS storage stand on the east side of the cavity and placed into the CSB as it sits on the CSB storage pads in the northwest corner of the cavity. The UGS will be lifted out of the water when placed into the CSB.

In order to minimize personnel exposure, temporary shielding will be utilized during the UGS lift.

The UGS fuel bundle alignment plate will be lifted just far enough above the operating deck to clear the CSB flange, moved laterally, then immediately lowered into the water. This operation is expected to take less than 15 minutes. When the UGS fuel bundle alignment plate is under water, the UGS can be aligned with the CSB alignment keys and lowered onto the CSB flange. Spacers on the CSB flange can be used to limit the engagement of the UGS keyways with the CSB alignment keys to provide additional interface clearances during the UGS seating operation .

The use of cameras will minimize the need for close observation of the insertion process. After the UGS is seated in the CSB and the lift rig is removed, the TAR 1 /11 /96 1.3-37

horizontal top cap will be placed over the entire UGS to reduce the radiation fields around the internals. Separate shielding will be placed over the in-core instrumentation guide tube stalks to minimize the radiation fields.

1.3.F.6 Post Annealing Lift Following the RV annealing and after the refueling cavity is flooded, the temporary shielding will be set up on the operating deck and the UGS shielding cap and instrumentation stalk shielding removed. This will allow the lift rig to be installed and the UGS to be lifted out of the CSB. The UGS will be lifted out of the water and moved east to its storage stand and lowered into the refueling cavity water.

This operation is expected to take less than 15 minutes. Critical RV dimensions are obtained in accordance with Section 2.2. After all required dimensions are checked, and the shielding wall around the CSB is removed, the CSB is placed into the reactor vessel in accordance with the fit up test of Section 2.3. This will be the first critical fit up of the internals into the reactor vessel after the annealing operations. The temporary CSB storage pads are removed from the- reactor cavity after the completion of RV head placement. The shielding will be removed from the Palisades site after the annealing program is completed. The shield module panels will be disassembled and the lead reclaimed from the panels to the extent possible. The panel containers may be cleaned free of the lead. Some small amount of the module panel may have to be dis~_o_s_e_d_o_f_a_s_L_S_A_w_a_s_te_._ _ _ _ _ _ _ _ _ _ 1 1.3.G References

• Engineering Specification for Primary Coolant Pipe and Fittings for Consumers Power Company, Specification No. 0070P-006, Rev. 2, dated 3/17/94.

Specification for Steam Generator Assemblies for Consumers Power Company,

• Specification No. 19377-PE-120, Rev. 4, dated 6/24/82.

"Hardened Concrete: Physical and Mechanical Aspects", American Concrete Institute Publication, Monograph No. 6, 1971.

"Compressive Strength of Concrete at Temperatures to 1600°F", Temperature and Concrete, American Concrete Institute Publication SP-25, 1971.

"Effects of Moisture Content on the Structural Properties of Portland Cement Concrete Exposed to Temperatures Up to 500°F", Temperature and Concrete, American Concrete Institute Publication SP-25, 1971.

Construction Testing Laboratories letter of September 13, 1995, M. G. Van Geem to R. B. Jenkins, "Literature Search on Changes in Compressive Strength due to Moderate'ly Elevated Temperature, Palisades Nuclear Plant".

"Design and Control of Concrete Mixtures," Portland Cement Association, 13th Edition, 1 992.

"Concrete Manual," U. S. Department of the Interior, 8th Edition, 1975 .

TAR 1/11 /96 1.3-38

Davis, H. S., "Effects of High Temperature Exposure on Concrete," Materials

• Research and Standards, V. 7, No.. 10, Oct. 1967, pp 452-459.

Nasser, K. W. and Lohtia, R. P. , "Mass Concrete Properties at High Temperatures," ACI Journal, No. 68-19, March 1971, pp. 180-186.

Kawahara, Abe H., Ito, T., and Haraguchi, A., "Influence Factors of Elevated Temperatures on Thermal Properties and Inelastic Behavior of Concrete", Concrete for Nuclear Reactors, American Concrete Institute Publication SP-34, 1972, pp.

847-870.

Nasser, K. W., and Chakraborty, M., "Temperature Effects on Strength and Elasticity of Concrete Containing Admixtures", Temperature Effects on Concrete, American Society for Testing and Materials STP 858, 1985, pp. 118-133.

"Class 1 Structures License Renewal Industry Report," NUMARC Report Number 90-06, June 1990.

Granata, S., and Montagnini, A., "Studies on Behavior of Concretes under Irradiation, Concrete for Nuclear Reactors, American Concrete Institute Publication SP-34, 1972, pp. 1163-1172 .

• TAR 1/11/96 1.3-39

-* Cycle Cycle Length 0° AZIMUTHAL ANGLE Neutron

[rad/hr]

Gamma

[rad/hr]

Total Neutron Total Gamma (EFPD} (rad} (rad) 1 379.4 1.37E + 5 2.39E+4 1.24E+9 2.18E+8 2 449.1 1.25E + 5 2.19E+4 1.35E+9 2.36E+8 3 349.5 1.13E + 5 2.43E+4 9.47E+8 2.04E+8 4 327.6 1.17E + 5 2.53E+4 9.18E+8 1.99E+8 5 394.6 1.12E + 5 2.42E+4 1.06E+9 2.29E+8 6 333.4 1.16E + 5 2.49E+4 9.25E+8 2.00E + 8 7 369.9 1.11E+5 2.39E+4 9.84E+8 2.12E+8 8 373.6 7.13E+4 1.62E +4 6.39E+8 1.45E+8 9 298~5 5.37E+4 1.13E +4 3.85E+8 8.09E+7 I

10 356.9 4.20E+4 9.44E+3 3.60E+8 8.09E+ 7  !

--L1~ ~4-30.4- _3.65E_+_4 _ __8. 25E_+_3_ _ 3.-2'ZE_+_8_ _ 8.5-3E_+_'Z I TOTAL I I I I 9.19E+9 I 1.89E + 9 I

• Cycle 1

Cycle Length (EFPD}

379.4 16° AZIMUTHAL ANGLE Neutron

[rad/hr]

1.34E + 5 Gamma

[rad/hr]

2.37E+4 Total Neutron (rad}

1.22E+9 Total Gamma (rad}

2.16E+8 2 449.1 1.24E + 5 2.18E+4 1.33E+9 2.34E+8 3 349.5 1.11E+5 2.40E+4 9.35E+8 2.02E+8 4 327.6 1.16E + 5 2.52E+4 9.15E+8 1.98E + 8 5 394.6 1.12E + 5 2.42E +4 1.06E + 9 2.29E+8 6 333.4 1.14E + 5 2.47E+4 9.16E+8 1.98E+8 7 369.9 1.1OE+5 2.37E+4 9.75E+8 2.10E +8 8 373.6 7.27E+4 1.61E+4 6.52E+8 1.44E+8 9 298.5 5.33E+4 1.13E +4 3.82E+8 8.07E+7 10 356.9 4.32E+4 9.51E+3 3.70E+8 8.14E+7 11 430.4 3.66E+4 8.19E+3 3.78E+8 8.46E + 7 I TOTAL I I I I 9.13E+9 I 1.88E+9 I Table 1.3.C.3-1 Maximum Neutron and Gamma Ray Dose Rates During Power Operation Incident on the Palisades Primary Biological Shield

• TAR 1/11/96 (Page 1 of 2) 1.3-40

• • Cycle Cycle Length 30° AZIMUTHAL ANGLE Neutron

[rad/hr]

Gamma

[rad/hr]

Total Neutron Total Gamma (EFPD) (rad) (rad) 1 379.4 1.25E+5 2.29E+4 1.14E + 9 2.08E+8 2 449.1 1.16E + 5 2.12E +4 1.25E+9 2.28E+8 3 349.5 9.91E+4 2.28E+4 8.31E+8 1.91E+8 4 327.6 1.06E+5 2.43E+4 8.35E+8 1.91E+8 5 394.6 1.02E+5 2.34E+4 9.67E+8 2.22E+8 6 333.4 1.03E+ 5 2.37E+4 8.24E+8 1.90E + 8 7 369.9 9.90E+4 2.27E+4 8.79E+8 2.02E+8 8 373.6 6.06E+4 1.47E+4 5.43E +8 1.32E+8 9 298.5 4.51E+4 1.05E+4 3.03E+8 7.49E + 7 10 356.9 4.05E+4 9.38E+3 3.35E+8 8.04E + 7

---

1 ---430-.4- --3.-35~+-4--- ---7-. 9 5 E-+-3-- ---3.40E-+-8-- --- -8.2-lE-+ -

I TOTAL I I I I 8.29E+9 I 1.80E+9 I

• Cycle 1

Cycle Length (EFPD) 379.4 45° AZIMUTHAL ANGLE Neutron

[rad/hr]

1.18E + 5 Gamma

[rad/hr]

2.22E+4 Total Neutron (rad) 1.07E+9 Total Gamma (rad) 2.02E +8 2 449.1 1.1OE+5 2.07E+4 1.19E + 9 2.23E+8 3 349.5 9.50E+4 2.23E+4 7.97E+8 1.87E+8 4 327.6 1.03E+5 2.39E+4 8.06E+8 1.88E+8 5 394.6 9.85E+4 2.29E+4 9.33E+8 2.-17E+8 6 333.4 9.90E+4 2.31E+4 7.92E+8 1.85E+8 7 369.9 9.54E+4 2.23E+4 8.47E + 8 1.98E +8 8 373.6 5.61E+4 1.42E+4 5.03E+8 1.28E +8 9 298.5 4.23E+4 1.01E+4 3.03E+8 7.20E + 7 10 356.9 3.92E+4 9.23E+3 3.35E+8 7.91E+7 11 430.4 3.29E+4 7.91E+3 3.40E + 8 8.17E+7 I TOTAL I I I I 7.92E + 9 I 1.76E +9 I Table 1.3.C.3-1 Maximum Neutron and Gamma Ray Dose Rates During Power Operation Incident on the Palisades Primary Biological

• TAR1/11/96 Shield (Page 2 of 2)

• 1 .3-41

TOTAL NEUTRON DOSE (Rad)

Depth Into oo 16° 30° 45° I Concrete (In) I I I I I 0.0 9.19E+9 9.13E+9 8.29E+9 7.92E+9 0.3 8.46E+9 8.42E+9 7.58E+9 7.21E+9 0.7 7.39E+9 7.37E+9 6.59E+9 6.24E + 9 1.3 6.00E+9 6.01E+9 5.33E+9 5.04E + 9 2.7 3.97E+9 4.02E+9 3.52E+9 3.35E+9 4.7 2.31E+9 2.28E+9 2.03E+9 1.88E +9

• 6.1 8.1 11.6 15.7 1.58E + 9 9.47E+8 4.36E+8 1.80E+8 1.53E+9 9.53E+8 4.32E+8 1.75E+8 1.39E+9 8.62E+8 3.97E+8 1.71E+8 1.25E+9 7.89E+8 3.54E+8 1.40E + 8 18.5 9.84E+7 9.24E+7 9.73E + 7 6.92E+ 7 19.7 6.91E+7 6.68E + 7 7.16E + 7 5.13E + 7 20.9 5.03E+7 5.00E+ 7 5.36E+ 7 3.83E+ 7 22.1 3.71E+7 3.72E+7 3.93E+7 2.86E+7 23.2 2.76E+7 2.81E+7 2.93E+ 7 2.15E+7 Table 1.3.C.3-2 Maximum Radiation Exposure as a Function of Radius into the Palisades Primary Biological Shield (Page 1 of 2)

• TAR 1/11/96 1.3-42

lI TOTAL GAMMA RAY DOSE (Rad)

Depth Into oo 16° 30° 45° Concrete (In) 0.0 .. 1.89E+9 1.88E+9 1.80E+9 1.76E+9 0.3 1.75E+9 1.74E + 9 1.66E+9 1.63E+9 0.7 1.65E+9 1.64E+9 1.56E+9 1.53E+ 9

. 1.3 1.54E+9 1.53E +9 1.45E + 9 1.42E + 9 2-:1 1~38E+9- :37ET9- :-30E--F9- --:-2/h-!:)

4.7 1.21E+9 1.20E+9 1.13E + 9 1.1OE-t-9

• 6.1 8.1 11.6 15.7 1.08E+9 8.94E+8 6.03E+8 3.46E+8 1.07E +9 8.93E+8 6.03E+8 3.46E+8 1.01E+9 8.38E+8 5.68E+8 3.33E+8 9.81E+8 8.12E+8 5.43E+8 3.05E+8 18.5 2.42E+8 2.37E+8 2.35E+8 1.99E+8 19.7 1.90E+8 1.86E +8 1.89E+8 1.60E+8 20.9 1.50E+8 1.50E+ 8 1.54E+8 1.29E+8 22.1 1.20E+8 1.21E+8 1.24E+8 1.04E+8 23.2 9.64E+ 7 9.78E+7 9.95E+ 7 8.38E+ 7 Table 1.3.C.3-2 Maximum Radiation Exposure as a Function of Radius into the Palisades Primary Biological Shield (Page 2 of 2)

• TAR 1/11 /96 1.3-43

• Distance From F(Z)

Core Midplane (ft)

+6.5 0.24

+6.0 0.35

+5.5 0.47

+5.0 0.59

+4.5 0.70

+4.0 0.78

+3.5 0.87

+3.0 0.91

+2.5 0.96

+2.0 0.96

+ 1~5 0-:97

+ 1.0 0.97

+0.5 0.97 0.0 1.00

-0.5 1.00

-1.0 0.98

-1.5 0.97

-2.0 0.97

-2.5 0.92

-3.0 0.86

-3.5 0.84

-4.0 0.73

-4.5 0.65

-5.0 0.54

-5.5 0.44

-6.0 0.32

-6.5 0.19 Table 1.3.C.3-3 Relative Axial Distribution of Neutron and Gamma Ray Dose Within the Palisades Biological Shield

• TAR 1 /11 /96 1.3-44

{

Node Liner Air RV Insulation RV Elevation Temp Temp O.D. Temp Temp (Of) (Of) (OF) (Of) 00 153 110 184 675 10 163 120 215 700 20 177 124 237 775 30 188 127 257 850 40 151 130 226 900 50 152 132 228 900 60 154 135 230 900 70 156 138 232 900 80 157

• 141 234 900 90 159 144 236 900 99 N/A 150 308 850 '

RV heat flux in Region 2 = 190 BTU/HR-FT 2 (at OD of vessel)

RV heat flux at flange in Region 3 = 108 BTU/HR-FT 2 (at OD of vessel)

RV Insulation Thermal = 1.174 BTU-IN/HR-FT 2 -°F Conductivitiy in Region 2 RV Insulation Thermal = 1.109 BTU-IN/HR-FT 2 -°F Conductivitiy in Region 3 Heat removed by 5000 CFM air flow,

• Q = 2.54 X 105 BTU/HR Table 1.3.C.9-1 Results of Reactor Cavity Model Using Corrected Insulation]

and Liner Temperatures from Insulation Characterization Study (Nominal Air Flow of 5000 CFM)

• TAR1/11/96 1.3-45

Node Liner Air RV Insulation RV Elevation Temp Temp O.D. Temp (°F) Temp (Of) (Of) (Of) 00 170 114 211 675 10 184 128 250 700 20 204 134 278 775 30 219 137 303 850 40 155 141 263 900 50 155 145 265 900 60 156 148 267 900 70 158 152 269 900 80 159 155 271 900 90 161 158 273 900 99 NIA 164 321 850 RV heat flux in Region 2 = 274 BTU/HR-FT 2 (at OD of vessel)

RV heat flux at flange in Region 3 = 153 BTU/HR-FT 2 (at OD of vessel)

RV Insulation Thermal = 1.850 BTU-IN/HR-FT 2 -°F Conductivity in Region 2 RV Insulation Thermal = 1. 742 BTU-IN/HR-FT 2 -°F I Conductivitiy in Region 3 Heat removed by 5000 CFM air flow, Q = 3.28 X 105 BTU/HR Table 1.3.C.9-2 Results of Reactor Cavity Model Using Corrected Insulation and Reflected Liner Temperatures from the Reactor Vessel Insulation Characterization Study (Nominal Air Flow of 5000 CFM)

• TAR1/11/96 1.3-46

I Nominal Design Design Description Material Schedule Size Pressure Temp.

(inches) (psia) (OF)

I Hot leg Carbon steel with roll Special 4" wallI 421D 2500 650 (reactor vessel to bonded stainless steel (3-3/4" Carbon steam generator) clad on ID with 1 /4" SS) I I

Cold leg Carbon steel with roll Special 3" wallI 301D 2500 650 (steam generator to bonded stainless steel (2-3/4" Carbon reactor vessel) clad on ID with 1/4" SS) I Surge line (hot leg Type 31 6 stainless steel 140 12 2500 700 to pressurizer) I Pressurizer spray Type 31 6 stainless steel 160 I 3 2500 650 I

Primary system Type 31 6 stainless steel 160 2 2500 650 drain lines Table 1.3.D-1 Piping Design Temperatures and Pressures for Palis~des PCS TAR1/11/96 1.3-47

• * \* ' '

J I

Item Pipe Section Nozzle(#) Nominal Base Cladding Notes Identification Pipe ~ize Material (10) 1 673-04 RV to SG1 Surge(1) i" 1 I SchedUIEl 140 Carbon Steel Forging 5/32" Type 304 SS Thermal Sleeve I

2 673-04 RV to SG1 Drain Nozzle(1) 2j' lnconel SB-166 N/A N/A Schedu le 160 1

3 673-04 RV to SG1 Temperature Measurement-1.261" OD lnconel SB-166 N/A N/A RTD(5) 4 673-04 RV to SG1 Pressure 3/!" lnconel SB-1 66 N/A N/A Measurement(4) SchedJle 1 60 I

Sample(1) I I

5 673-01 RV to SG2 Shutdown Cooling 12" Alloy Steel . 5/32" N/A Outlet(1 l SchedJle 140 Forging Type 304 SS I A-508-64-CL 1 I

6 673-01 RV to SG2 Temperature 1.260" OD lnconel SB-166 N/A N/A Measurement-RTD(5) I 7 673-01 RV to SG2 Pressure 3/h" lnconel SB-166 N/A N/A I

I Measurement(4) Schedule 160 Sample(1) I I

I 8 673-11 Pump 1A to RV Safety Injection 12" Carbon Steel 5/32" Min lnconel SDC lnlet(1) SchedJle 140 Forging Type 304 SS SB 168 Thermal I Sleeve Table 1.3.D-2 Piping Nozzles in Palisades PCS (Page 1 of 3)

I TAR1/11/96 1.3-48

Item Pipe Section Nozzle(#) Nominal/ Base Cladd ing Notes Identification Pipe Siz~ Material (ID I 9 673-11 Pump 1A to RV Charging lnlet(1) 2" I lnconel SB-166 N/A .lnconel

• Schedule 1 60 SB 168 Thermal I Sleeve 10 673-11 Pump 1A to RV Temperature Measurement-1.260" 10 lnconel SB-1 66 N/A N/A RTD(3) 11 673-08 Pump 2A to RV Safety Injection 12" I Carbon Steel 5/32" Min lnconel SDC Inlet( 1 ) Schedule 140 Forging Type 30 4 SS SB 168 Thermal I Sleeve 12 673-08 Pump 2A to RV Spray Nozzle(1) 3" I lnconel SB-1 66 N/A N/A Schedule 160 13 673-08 Pump 2A to RV Charging lnlet(1) 2" I lnconel SB-1 66 N/A lnconel Schedule 160 SB 168 Thermal Sleeve

- I 14 673-08 Pump 2A to RV Temperature 1.260" CDD lnconel SB-1 66 N/A N/A Measurement-RTD(3) I 15 674-04 Pump 18 to RV Safety Injection 12" / Carbon Steel 5/32" Min lnconel SDC lnlet(1) r Schedule 40 Forging Type 30 4 SS SB 168 Thermal Sleeve 16 674-04 Pump 1 B to RV Spray(1) lnconel SB-1 66 N/A N/A 3" I Schedule 160 Table 1.3.D-2 Piping Nozzles in Palisades PCS (Page 2 of 3)

TAR1/11/96 1.3-49

• i

• I Item Pipe Section Nozzle(#) Nominal/ Base Cladding Notes Identification Pipe Siz~ Material (ID)

I 17 674-04 Pump 1 B to RV Temperature 1.260" op lnconel SB-1 66 N/A N/A Measurement- I RTD(3) I 18 674-01 Pump ;28 to RV Safety Injection 12" I Carbon Steel 5/32" Min lnconel SDC lnlet(1 l Schedule 140 I Forging Type 304 SS SB 168 Thermal I Sleeve 19 674-01 Pump 28 to RV Temperature 1.260" do lnconel SB-166 N/A N/A Measurement-RTD(3) I I

20 674-05 674-07L to Drain(1) 2" I lnconel SB-166 Pump 1A Schedule 160 N/A N/A Pressure 3/4" 1 lnconel SB-1 66 Measurement(2) Schedule 1160 I

21 674-05 674-07 L to Drain(1) 2" I lnconel SB-166 I

Pump 2A Scheg~~; 60 N/A N/A Pressure lnconel SB-1 66 Measurement(2) Schedule J60 22 673-05 674-07R to Drain(1) 2" I lnconel SB-166 Pump 18 Schedule ~ 60 N/A N/A Pressure 3/4" I lnconel SB-166 Measurement(2) Schedule 11 60 23 673-05 674-07R to Letdown and 2" I 1

lnconel SB-166 Pump 28 Drain(1) Schedule 1160 N/A N/A Pressure 3/4" lnconel SB-166 Measurement(2) Schedule /1 60 Table 1.3.D-2 Piping Nozzles in Palisades PCS (Page 3 of 3)

TAR 1/11/96 1.3-50

• I Pipe Assembly #

673-01 673-04 I Description Elbow to Pipe Girth Seam Elbow to Pipe Girth Seam I Type of Indication Processing Marks Surface Defects, Processing Marks I

673-05(2) Elbow to Pipe Girth Seam Processing Marks 673-08 Elbow to Pipe Girth Seam Processing Marks 673-11 Elbow to Pipe Girth Seam Surface Defects, Processing Marks 674-04 Elbow to Pipe Girth Seam Surface Defects, Processing Marks 674-01 Elbow to Pipe Girth Seam Processing Marks 674-05(2) Elbow to Pipe Girth Seam Processing Marks 674-07R( 1) Elbow to Pipe Girth Seam Processing Marks 674-07R(2) Elbow to Pipe Girth Seam Processing Maiks 674-07L(1) Elbow to Pipe Girth Seam Processing Marks 674-07L(2) Elbow to Pipe Girth Seam Processing Marks 674-05(2) Safe-End to Pressure Surface Defects l.D.

Measurement & Sampling Nozzle 673-01 Nozzle to Pipe-Pressure Processing Marks, Measurement & Sampling Porosity, Slag Inclusions, Nozzle Surface Defects 673-04 Nozzle to Pipe-Pressure Porosity, Processing Measurement & Sampling Marks, Slag Inclusions, Nozzle Surface Defects 673-04 Drain Nozzle to Pipe Girth Processing Marks Seam 673-05 & 674-05 Letdown & Drain Nozzle Processing Marks, to Pipe Girth Seam Porosity 673-08 & 674-04 Spray Nozzle to Pipe Girth Slag Inclusions, Seam Processing Marks 673-01 Safe-End to Shutdown Surface Defects, Porosity Cooling Outlet Nozzle Table 1.3.D.9.1-1 As-Left Radiographic Indications on Palisades PCS Piping (Page 1 of 2)

• TAR1/11/96 1.3-51

Pipe Assembly # Description Type of Indication 673-01 Nozzle to Pipe-Shutdown Processing Marks Cooling Outlet Nozzle 673-05 & 674-05 Nozzle to Pipe-Pressure Processing Marks, Measurment Nozzle Porosity 673-08 & 673-11 Nozzle to Pipe-Charging Processing Marks, Inlet Nozzle Porosity, Surface Defects 673-04 Safe-End to Surge Nozzle Processing Marks, Assembly Porosity 673-04 Surge Nozzle Assembly Processing Marks, Slag to Pipe Inclusions 673-08, 673-11, Safe-End to Safety Porosity, Processing 674-01, 674-04 Injection Nozzle Marks 673-08, 673-11, Nozzle to Pipe-Safety Processing Marks, Slag 674-01, 674-04 Injection Nozzle Inclusions Table 1.3.D.9.1-1 As-Left Radiographic Indications on Palisades PCS Piping (Page 2 of 2)

• TAR 1/11 /96 1.3-52

I Weld Seam I Location I 3-674 Elbow To Pipe Girth (Assy. 674-01) 5-6748 Elbow To Pipe Girth (Assy. 674-05) 7-6748 Elbow To Pipe Girth (Assy. 674-07R) 3-673M,P Safe-end to Measurement and Sampling Nozzle 3-674 Elbow To Pipe Girth (Assy. 674-01) 5-6748 Elbow To Pipe Girth (Assy. 674-05) 7-7648 Elbow To Pipe Girth (Assy. 674-07R)

• 5-675 Drain Nozzle to Pipe Girth (Assy. 674-04) 7-7658 Let Down and Drain Nozzle to Pipe Girth (Assy. 673-05 or 674-05) 9-675A,8 Spray Nozzle to Pipe Girth (Assy. 673-08 or 674-04)12-675 Shutdown Cooling Nozzle to Pipe Girth (Assy. 673-01) 15-675G,H Pressure Measurement Nozzle to Pipe Girth (Assy. 673-05 or 674-05) 1-676A,8 Charging Inlet Nozzle to Pipe Girth (Assy. 673-08 or 673-11) 4-676 Safe-end to Surge Nozzle Girth (Assy. 673-04) 5-676 Surge Nozzle Assy. to Pipe Girth (Assy. 673-04) 8-676A Safe-end to Safety Injection Nozzle Girth (Assy. 673-08, 673-11, 674-01 or 674-04) 9-676A/D Safety Injection Nozzle to Pipe Girth (Assy. 673-08, 673-11, 674-01 or 674-04)

Table 1.3.D.9. 1-2 Recorded Weld Repairs for Palisades PCS Piping I*

I TAR1/11/96 1.3-53

• Weld#

Weld Description 151 Performed 151 Implemented 151 Results (Interval-Outage) 1 Nozzle to Transition UT 1-3, 1-9, 2-9 No indications Piece 2 Transition Piece to UT 1-3, 2-9 No indications Pipe 2LD Longitudinal Weld- --- Inaccessible ---

Downstream 3LU Longitudinal Weld- --- Inaccessible ---

Upstream 3 Pipe to Pipe --- Inaccessible ---

3LD Longitudinal Weld- --- Inaccessible ---

Downstream 4LU Longitudinal Weld- --- No previous ---

Upstream examinations

• 4 4LD1 4LD2 Pipe to Elbow Longitudinal Weld-Downstream Longitudinal Weld-No previous examinations No previous examinations No previous Downstream examinations 5LU-1 Longitudinal Weld VT 1-1 No indications Upstream ('71 Code) 5LU-2 Longitudinal Weld- VT 1-1 No indications Upstream ('71 Code) 5 Elbow to Transition VT 1-1 No indications Piece ('71 Code)

(New weld in 1991) 6 Transition Piece to surface, UT 2-6 Surface indications Nozzle 1-201-258 only, all acceptable to ASME Section XI Table 1.3.D.9.2-1 Weld lnservice Examination Results on Palisades PCS Piping, Line PCS-42-RCL-1 H TAR1/11/96 1.3-54

• Weld#

Weld Description ISi Performed ISi Implemented ISi Results (Interval-Outage) 1 Nozzle to Transition UT 1-3, 1-9, 2-9 No indications Piece 2 Transition Piece to UT 1-3, 2-9 No indications Pipe 2LD Longitudinal Weld- --- Inaccessible ---

Downstream 3LU Longitudinal Weld- --- Inaccessible ---

Upstream 3 Pipe to Pipe --- Inaccessible ---

3LD Longitudinal Weld- --- Inaccessible ---

Downstream 4LU Longitudinal Weld- UT 1-4 No indications Upstream

• 4 4LD-1 4LD-2 Pipe to Elbow Longitudinal Weld-Downstream Longitudinal Weld-Downstream UT UT UT 1-4 1-4 1-4 No indications No indications No indications 5LU-1 Longitudinal Weld- --- No previous ---

Upstream examinations 5LU-2 Longitudinal Weld- --- No previous ---

Upstream examinations 5 Elbow to Transition surface, new No indications Piece UT (New weld in 1991) 6 Transition Piece to surface, 2-9 Surface indications Nozzle 2-201-358 UT only, all are acceptable to ASME Section XI Table 1.3.D.9.2-2 Weld lnservice Examination Results for Palisades PCS Piping, Line PCS-42-RCL-2H

• TAR1/11/96 1 .3-55

Weld ISi ISi ISi Weld# Description Performed Implemented Results (Interval-Outage) 1 Nozzle to Transition surface, UT 2-7 UT indication Piece 1-401-258A only (embedded, (New weld in 1991) 30% DAC),

acceptable to ASME Section XI 2 Transition. Piece to surface, UT new Four UT Elbow indications only (New weld in 1991) (all embedded),

aii acceptable to ASME Section XI 2LD-1 Longitudinal Weld- --- No Previous ---

Downstream Examinations 2LD-2 Longitudinal Weld- --- No Previous ---

• 3LU-1 3LU-2 Downstream Longitudinal Weld-Upstream Longitudinal Weld-Upstream VT

('71 Code)

VT

('71 Code)

Examinations 1-1 1-1 No indications No indications 3 Elbow to Pipe UT new UT indication (New weld in 1991) only (embedded),

acceptable to ASME Section XI 3LD Longitudinal Weld- VT 1-1 No indications Downstream ('71 Code) 4LU Longitudinal Weld- --- No Previous ---

Upstream Examinations Table 1.3.D.9.2-3 Weld lnservice Examination Results for Palisades PCS Piping, Line PCS-30-RCL-1 A (Page 1 of 4)

• TAR 1/11/96 1.3-56

Weld ISi ISi ISi Weld# Description Performed Implemented Results (Interval-Outage) 4 Pipe to Elbow --- No Previous ---

Examinations 4LD-1 Longitudinal Weld- --- No Previous ---

Upstream Examinations 4LD-2 Longitudinal Weld- --- No Previous ---

Downstream Examinations 5LU-1 Longitudinal Weld- --- No Previous ---

Upstream Examinations 5LU-2 Longitudinal Weld- --- No Previous ---

Upstream Examinations 5 Elbow to Pipe --- No Previous ---

Examinations

• 5LD 6LU 6

Longitudinal Weld-Downstream Longitudinal Weld-Upstream Pipe to Transition Piece UT UT No Previous Examinations 1-4 1-4 No indications No indications 1-201-258 7 Transition Piece to Pipe UT 1-4, 2-4 No indications 8 Elbow to Pump P-50A surface, UT . 2-4 Surface indications only, all acceptable to ASME Section XI 9 Pump P-50A to Transition surface, UT 1-6, 2-4 Surface Piece indications only, all acceptable to ASME Section XI Table 1.3.D.9.2-3 Weld lnservice Examination Results for Palisades PCS Piping, Line PCS-30-RCL-1A (Page 2 of 4)

• TAR1/11/96 1.3-57

Weld ISi ISi ISi Weld# Description Performed Implemented Results (Interval-Outage) 10 Transition Piece to Pipe --- No Previous ---

Examinations 10LD Longitudinal Weld- --- No Previous ---

Downstream Examinations 11 LU Longitudinal Weld- --- No Previous ---

Downstream Examinations 11 Pipe to Pipe --- No Previous ---

Examinations 11LD Longitudinal Weld- --- No Previous ---

Downstream Examinations 12LU Longitudinal Weld- --- Inaccessible ---

Upstream 12 Pipe to Pipe --- Inaccessible ---

• 12LD 13LU Longitudinal Weld-Downstream Longitudinal Weld-Upstream Inaccessible Inaccessible 13 Pipe to Pipe Inaccessible 13LD Longitudinal Weld- --- Inaccessible ---

Downstream 14LU Longitudinal Weld- --- Inaccessible ---

Upstream 14 Pipe to Elbow --- Inaccessible ---

14LD-1 Longitudinal Weld- --- Inaccessible ---

Downstream 14LD-2 Longitudinal Weld- --- Inaccessible ---

Downstream Table 1.3.D.9.2-3 Weld lnservice Examination Results for Palisades PCS Piping, Line PCS-30-RCL-1 A (Page 3 of 4)

• TAR 1/11/96 1 .3-58

Weld ISi ISi ISi Weld# Description Performed Implemented Results (Interval-Outage) 15LU-1 Longitudinal Weld- --- Inaccessible ---

Upstream 15LU-2 Longitudinal Weld- --- Inaccessible ---

Upstream 15 Elbow to Transition UT 2-9 No indications Piece 15LD-1 Longitudinal Weld- UT 2-9 No indications Downstream

• 15LD-2 16LU-1 16LU-2 Longitudinal Weld-Downstream Longitudinal Weld-Upstream Longitudinal Weld-UT UT UT 2-9 2-9 2-9 No indications No indications No indications Upstream 16 Transition Piece to UT 1-9, 2-9 No indications Nozzle Table 1.3.D.9.2-3 Weld lnservice Examination Results for Palisades PCS Piping, Line PCS-30-RCL-1 A (Page 4 of 4)

• TAR1/11/96 1 .3-59

Weld ISi ISi ISi Weld# Description Performed Implemented Results (Interval-Outage) 1 Nozzle to Transition surface, new No indications Piece 1-401-2588 UT (New weld in 1991) 2 Transition Piece to surface, new No indications Elbow UT (New weld in 1 991) 2LD-1 Longitudinal Weld- VT 1-1 No indications

• Downstream ('71 Code) 2LD-2 Longitudinal Weld- VT 1-1 No indications Downstream ('71 Code) 3LU-1 Longitudinal Weld- --- No Previous ---

Upstream - Examinations 3LU-2 Longitudinal Weld- --- No Previous ---

Upstream Examinations 3 Elbow to Pipe surface, new No indications (New weld in 1991) UT 3LD Longitudinal Weld- --- No Previous ---

Downstream Examinations 4LU Longitudinal Weld- UT 1-4 No indications Upstream 4 Pipe to Elbow surface, 1-4 UT indication UT only (embedded),

acceptable to ASME Section XI 4LD-1 Longitudinal Weld- UT 1-4 No indications Downstream 4LD-2 Longitudinal Weld- UT 1-4 No indications Downstream Table 1.3.D.9.2-4 Weld lnservice Examination Results for* Palisades PCS Piping, Line PCS-30-RCL-1 B (Page 1 of 3)

• TAR 1/11/96 1.3-60

Weld ISi ISi ISi Weld# Description Performed Implemented Results (Interval-Outage) 5LU-1 Longitudinal Weld- --- No Previous ---

Upstream Examinations 5LU-2 Longitudinal Weld-- --- No Previous ---

Upstream Examinations 5 Elbow to Pipe --- No Previous ---

Examinations 5LD Longitudinal Weld- --- No Previous ---

Downstream Examinations 6LU Longitudinal Weld- UT 1-4 No indications Upstream 6 Pipe to Transition Elbow UT 1-9 No indications 7 Transition Piece to surface, 1-4, 2-4 Surface Elbow UT indications only, all acceptable to ASME Section XI 8 Elbow to Pump P-508 surface, 2-9 Surface UT indications only, all acceptable to ASME Section XI 9 Pump P-508 to surface, 1-9,2-7 Surface Transition Pipe UT indications only, all acceptable to ASME Section XI 10 Transition Piece to Pipe UT 1-4 No indications 10LD Longitudinal Weld- UT 1--4 No indications Downstream 11 LU Longitudinal Weld- --- Inaccessible ---

Upstream 11 Pipe to Pipe --- Inaccessible ---

Table 1.3.D.9.2-4 Weld lnservice Examination Results for Palisades PCS Piping, Line PCS-30-RCL-1 B (Page 2 of 3)

• TAR1/11/96 1.3-61

Weld ISi ISi ISi Weld# Description Performed Implemented Results (Interval-Outage) 11 LD Longitudinal Weld- --- Inaccessible ---

Downstream 12LU Longitudinal Weld- --- Inaccessible ---

Upstream 12 Pipe to Elbow --- Inaccessible ---

12LD-1 Longitudinal Weld- --- Inaccessible ---

Downstream 12LD-2 Longitudinal Weld- --- Inaccessible ---

Downstream 13LU-1 Longitudinal Weld- --- Inaccessible ---

Upstream 13LU-2 Longitudinal Weld- --- Inaccessible ---

• 13 13LD-1 Upstream Elbow to Transition Piece Longitudinal Weld-Downstream UT UT 2-9 2-9 No indications No indications 13LD-2 Longitudinal Weld- UT 2-9 No indications Downstream 14LU-1 Longitudinal Weld- UT 2-9 No indications Upstream 14LU-2 Longitudinal Weld- UT 2-9 No indications Upstream 14 Transition Piece to UT 2-9 No indications Nozzle Table 1.3.D.9.2-4 Weld lnservice Examination Results for Palisades PCS Piping, Line PCS-30-RCL-1 B (Page 3 of 3)

• TAR 1/11 /96 1.3-62

Weld ISi ISi ISi Weld# Description Performed Implemented Results (Interval-Outage) 1 Nozzle to Transition surface, UT new No indications Piece 2-401-358A (New weld in 1991) 2 Transition Piece to surface, UT new No indications Elbow (New weld in 1991 l 2LD-1 Longitudinal Weld- --- No Previous ---

Downstream Examinations 2LD-2 Longitudinal Weld- --- No Previous ---

Downstream Examinations 3LU-1 Longitudinal Weld- UT 1-4 No indications Upstream 3LU-2 Longitudinal Weld- UT 1-4 No indications Upstream 3 Elbow to Pipe surface, UT new UT indication (New weld in 1991 l only, acceptable to ASME Section XI 3LD Longitudinal Weld- UT 1-4 No indications Downstream 4LU Longitudinal Weld- VT 1-1 No indications Upstream ('71 Code) 4 Pipe to Elbow VT 1-1 No indications

('71 Code) 4LD-1 Longitudinal Weld- VT 1-1 No indications Downstream ('71 Code) 4LD-2 Longitudinal Weld- VT 1-1 No indications Downstream ('71 Code)

Table 1.3.D.9.2-5 Weld lnservice Examination Results for Palisades PCS Piping, Line PCS-30-RCL-2A (Page 1 of 3)

• TAR1/11/96 1 .3-63

Weld ISi ISi ISi Weld# Description Performed Implemented Results (Interval-Outage) 5LU-1 Longitudinal Weld- VT 1-1 No indications Upstream ('71 Code) 5LU-2 Longitudinal Weld- VT 1-1 No indications Upstream ('71 Code) 5 Elbow to Pipe VT 1-1 No indications

('71 Code) 5LD Longitudinal Weld- VT 1-1 No indications Upstream ('71 Code) 7LU Longitudinal Weld- VT 1-1 No indications Upstream ('71 Code) 7 Pipe to Transition VT 1-1 No indications Piece ('71 Code) 8 Transition Piece to UT 1-4, 2-4 No indications Elbow

• 9 10 Elbow to Pump P-50C Pump P-50C to Transition Piece UT surface, UT 1-4, 2-4 1-9, 2-9 No indications Surface indications only, all acceptable to ASME Section XI 11 Transition Piece to VT 1-1 No indications Pipe ('71 Code) 11 LD Longitudinal Weld- --- Inaccessible ---

Downstream 12LU Longitudinal Weld- --- Inaccessible ---

Upstream 12 Pipe to Pipe --- Inaccessible ---

12LD Longitudinal Weld- --- Inaccessible ---

Downstream Table 1.3.D.9.2-5 Weld lnservice Examination Results for Palisades PCS Piping, Line PCS-30-RCL-2A (Page 2 of 3)

• TAR 1/11 /96 1.3-64

Weld ISi ISi ISi Weld# Description Performed Implemented Results (Interval-Outage) 13LU Longitudinal Weld- --- Inaccessible ---

Upstream 13 Pipe to Elbow --- Inaccessible ---

13LD-l Longitudinal Weld- --- Inaccessible ---

Downstream 13LD-2 Longitudinal Weld- --- Inaccessible ---

Downstream 14LU-1 Longitudinal Weld- --- Inaccessible ---

Upstream 14LU-2 Longitudinal Weld- --- Inaccessible ---

Upstream 14 Elbow to Transition Piece UT 2-9 No indications 14LD-1 Longitudinal Weld- UT 2-9 No indications Downstream 14LD-2 Longitudinal Weld- UT 2-9 No indications Downstream 15LU-1 Longitudinal Weld- UT 2-9 No indications Upstream 15LU-2 Longitudinal Weld- UT 2-9 No indications Upstream 15 Transition Piece to UT 1-9, 2-9 No indications Nozzle Table 1.3.D.9.2-5 Weld lnservice Examination Results for Palisades PCS Piping, Line PCS-30-RCL-2A (Page 3 of 3)

• TAR 1/11 /96 1.3-65

Weld ISi ISi ISi Weld# Description Performed Implemented Results (Interval-Outage) 1 Nozzle to Transition surface, new No indications Piece 2-401-3580 UT (New weld in 1991) 2 Transition Piece to surface, new No indications Elbow UT (New weld in 1991) 2LD-1 Longitudinal Weld- --- No Previous ---

Downstream Examinations 2LD-2 Longitudinal Weld- --- No Previous ---

Downstream Examinations 3LU-1 Longitudinal Weld- --- No Previous ---

Upstream Examinations 3LU-2 Longitudinal Weld- --- No Previous ---

Upstream Examinations

• 3 3LD Elbow to Pipe (New weld in 1991)

Longitudinal Weld-surface, UT new No Previous UT indications only (embedded),

all acceptable per ASME Section XI Downstream Examinations 4LU Longitudinal Weld- UT 2-1 No indications Upstream 4 Pipe to Elbow UT 2-1 No indications 4LD-1 Longitudinal Weld- UT 2-1 No indications Downstream 4LD-2 Longitudinal Weld- UT 2-1 No indications Downstream Table 1.3.D.9.2-6 Weld lnservice Examination Results for Palisades PCS Piping, Line PCS-30-RCL-2B (Page 1 of 3)

• TAR 1/11/96 1.3-66

• Weld#

Weld Description ISi Performed ISi Implemented (Interval-Outage)

ISi Results 5LU-1 Longitudinal Weld- UT 1-4 No indications Upstream 5LU-2 Longitudinal Weld- UT 1-4 No indications Upstream 5 Elbow to Pipe UT 1-4 No indications 5LD Longitudinal Weld- UT 1-4 No indications Downstream 6LU Longitudinal Weld- UT 1-4 No indications Upstream 6 Pipe to Transition Piece UT 1-4 No indications 7 Transition Piece to surface, 1-6,2-4 Surface Elbow UT indications and UT indication (embedded), all acceptable per ASME Section XI 8 Elbow to Pump P-50D UT 2-4 No indications 9 Pump P-50D to surface, 1-9, 2-9 Surface Transition Piece UT indications only, all acceptable per ASME Section XI 10 Transition Piece to Pipe --- No Previous ---

Examinations 10LD Longitudinal Weld- --- No Previous ---

Downstream Examinations 11LU Longitudinal Weld- --- Inaccessible ---

Upstream 11 Pipe to Pipe --- Inaccessible ---

Table 1.3.D.9.2-6 Weld lnservice Examination Results for Palisades PCS Piping, Line PCS-30-RCL-28 (Page 2 of 3)

• TAR1/11/96 1.3-67

Weld ISi ISi ISi Weld# Description Performed Implemented Results (Interval-Outage) 11 LO Longitudinal Weld- --- Inaccessible ---

Downstream 12LU Longitudinal Weld- --- Inaccessible ---

Upstream 12 Pipe to Pipe --- Inaccessible ---

12LD Longitudinal Weld- --- Inaccessible ---

Downstream 13LU Longitudinal Weld- * --- Inaccessible ---

Upstream 13 Pipe to Elbow --- Inaccessible ---

13LD-1 Longitudinal Weld- --- lnaccessibie ---

Downstream 13LD-2 Longitudinal Weld- --- Inaccessible ---

Downstream 14LU-1 Longitudinal Weld- --- Inaccessible ---

Upstream 14LU-2 Longitudinal Weld- --- Inaccessible ---

Upstream 14 Elbow to Transition UT 2-9 No indications Piece 14LD-1 Longitudinal Weld- UT 2-9 No indications Downstream 14LD-2 Longitudinal Weld- UT 2-9 No indications Downstream 15LU-1 Longitudinal Weld- UT 2-9 No indications Upstream 15LU-2 Longitudinal Weld- UT 2-9 No indications Upstream 15 Transition Piece to UT 1-9, 2-9 No indications Nozzle Table 1.3.D.9.2-6 Weld lnservice Examination Results for Palisades PCS Piping, Line PCS-30-RCL-2B (Page 3 of 3)

• TAR 1/11 /96 1.3-68

~--* ....*

t. Figure 1.3.A.1-1 Piping Components of Palisades Primary Coolant System - Plan View TAR 1/11 /96 1.3-69

ct 0

~

"":z

~

li

""t; g

IO

'i~ ~~ ~

2 0

~

~

w L.i.I

~

I ::2

~

!w

• Figure 1.3.A. 1-2 Piping Components of Palisades Primary Coolant System -

Elevation View TAR1/11/96 1.3-70

El. 624'

• 6" REACTOR Region 3

~

T ....

a::

0 Region 2

<.:I

(.)

<I:

j_ El. 601'

• 6"'

~~

')Y Region 1

~I>

......___ _ _ _ _ _ _ _ _ Bloshield Wall Figure 1.3.C-1 Palisades Reactor Cavity Insulation Regions TAR 1/11 /96 1 .3-71

• • LINER AIR INSULATION REACTOR VESSEL 624. 31 IS~@

AIR 620 8 @)

AIR 0 620

~ 8

§ 8

(§

(§) @

62

• 10

(§) @ @ @ @

480 s 18 @)

490 @ @) ~ ~

601.50 Figure 1.3.C.9-1 Reactor Cavity Thermal Model TAR 1/11/96 1.3-72

1 /4" Carbon Steel Liner Reactor Vessel

RV

center

Line

<>Thermal Radiation Conductor Q Nodes D Conductance and Convection Conductors Figure 1.3.C.9-2 Typical Radial Cross Section of _Thermal Model - Region 2 TAR1/11/96 1.3-73

1 /4" Carbon Steel Liner Reactor Vessel

RV Insulated

Center Boundary

une 7.1667' Q Nodes D Conductance and Convection Conductors 0 Thermal Radiation Conductor Figure 1.3.C.9-3 Typical Radial Cross Section of Thermal Model - Region 3 TAR.1/11/96 1.3-74

___ 1 DISCHARGE I

, ~~\

LEG '

• MOTOR STAND PUMP CASING SUCTION LEG

• Figure 1.3.D.6-1 TAR 1/11/96 Primary Coolant Pump Support Arrangement 1.3-75

• DIRECTION OF THERMAL GROWTH 1 OPERATING GAP 2 PLACES B B t t OPERATING GAP GUIDE SHIM 2 PLACES STOP BLOCK VIEWA*A PUMP FOOT

~

SUPPORT - OPERATING GAP PAD 2 PLACES

. - - - - GUIDE SHIM CIRCULAR LUBRICATED PLATES LUBRICATED r--

SURFACES OF CIRCULAR

~~~~~~~:d---\)-PLATES

, I

~EEL).J FRAME HOLD*DOWN NUT VIEWB*B

• Figure 1.3.D.6-2 TAR 1/11/96 Primary Coolant Pump Support Details 1 .3-76

.).

• Snubber* Typical 8 places Keyway

• Typical 2 places

• • snubber support

• r** A---.-1~ 8 places l.. L ** J I  :

L ** .J View A*A f

A A

Key (keyway not shown)

I SNUBBER SUPPORTS I --

f -~---'

u I! i I

i .I I i

- TowardR.V.

OUTLET INLET NOZZLE NOZZLE SLIDING BASE Figure 1.3.D. 7-1 Palisades Steam Generator Support Arrangement TAR 1/11/96 1.3-77

SLOTTED PLATES TYPICAL EXPANSION PLATE TYPICAL EMBEDDED KEY

- - !HOT LEG STOP)

EMBEDDED KEY GAP TO BE SET AND VERIFIED EXPANSION DURING HOT FUNCTIONAL PLATE TESTING TYPICAL D D f t VIEWC*C CAST STEEL SLIDING BASE

~SLIDING SLIDING BEARINGS BEARING VIEWD*D EMBEDDED KEYS HOLD*DOWN BOLT VIEW E-E Figure 1.3.D. 7-2 Palisades Steam Generator Sliding Base TAR 1/11/96 1.3-78

• _ _,_1--.__ __,_ _....,_

OPERATING CLEARANCES BOTH SIDES SHIM 1

'-*c A I SHIM SECTION B*B RV PAD SOCKET MOUNTING BOLTS BEARING SOCKET SLIDING BEARING (CYLINDRICAL)

SHIM PLATE SECTION A*A

• Figure 1.3.D.8-1 TAR 1/11/96 Palisades Reactor Vessel Support Details 1.3-79

REACTOR VESSEL Figure 1.3.D.9.2-1 Palisades PCS Piping and Component Weld Identification TAR 1/11/96 1.3-80